iii

ABSTRACT

This report is an assessment of potential radiation doses associated with the current

exemptions from licensing for the majority of Part 30 byproduct and Parts 40 and 70 source

material in Title 10 of the Code of Federal Regulations (CFR). Doses were estimated for the

normal life cycle of a particular product or material, covering distribution and transport, intended

or expected routine use, and disposal using the ICRP 26 and 30 dose assessment

methodology, which was incorporated into the current requirements of 10 CFR Part 20 in

May 1991. In addition, assessments of potential doses due to accidents and misuse were

estimated. Also presented is an assessment of potential radiological impacts associated with

selected products containing byproduct material which currently may be used under a general

or specific license and may be candidates for exemption from licensing requirements.

v

CONTENTS

ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii

EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxi

FOREWORD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxxiii

1 INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1–1

1.1 Purpose of Present Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1–1

1.2 Basis for Existing Exemptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1–2

1.3 Requirements for Radiation Protection of the Public . . . . . . . . . . . . . . . . . . . . . . 1–4

1.4 Description of Present Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1–5

1.4.1 Dose Assessments for Normal Life Cycle . . . . . . . . . . . . . . . . . . . . . . 1–5

1.4.2 Dose Assessment for Accidents and Misuse . . . . . . . . . . . . . . . . . . . . 1–9

1.5 Organization and Content of Report ................................. 1–10

2 EXEMPTIONS FOR BYPRODUCT MATERIAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–1

2.1 Properties of Byproduct Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–1

2.1.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–1

2.1.2 Definition of Byproduct Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–1

2.1.3 Decay Data for Selected Byproduct Materials . . . . . . . . . . . . . . . . . . . 2–2

2.1.4 Dosimetry Data for Selected Byproduct Materials . . . . . . . . . . . . . . . . 2–2

2.2 Concentrations of Byproduct Material ............................... 2–15

2.2.1 Introduction ............................................ 2–15

2.2.2 Description of Products or Materials .......................... 2–16

2.2.3 Summary of Previous Analyses and Assessments ............... 2–17

2.2.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–20

2.2.5 Effect of Changes in Dose Limits and Internal Dosimetry Models . . . 2–32

2.2.6 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–33

2.3 Timepieces, Hands, and Dials ..................................... 2–49

2.3.1 Introduction ............................................ 2–49

2.3.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–49

2.3.3 Summary of Previous Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . 2–51

2.3.4 Current Assessment for Timepieces Containing 3

H .............. 2–53

2.3.5 Current Assessment for Timepieces Containing 147Pm . . . . . . . . . . . 2–60

2.3.6 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–65

2.4 Automobile Lock Illuminators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–73

2.4.1 Introduction ............................................ 2–73

2.4.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–73

2.4.3 Summary of Previous Analyses and Assessments ............... 2–73

2.4.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–74

2.4.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–78
vi

2.5 Balances of Precision . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–87

2.5.1 Introduction ............................................ 2–87

2.5.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–87

2.5.3 Summary of Previous Analyses and Assessments ............... 2–87

2.5.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–87

2.5.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–90

2.6 Automobile Shift Quadrants ....................................... 2–95

2.6.1 Introduction ............................................ 2–95

2.6.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–95

2.6.3 Summary of Previous Analyses and Assessments ............... 2–95

2.6.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–95

2.6.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–99

2.7 Marine Compasses and Navigational Instruments ..................... 2–105

2.7.1 Introduction ........................................... 2–105

2.7.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–105

2.7.3 Summary of Previous Assessments . . . . . . . . . . . . . . . . . . . . . . . . . 2–105

2.7.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–107

2.7.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–108

2.8 Thermostat Dials and Pointers Containing Tritium ..................... 2–111

2.8.1 Introduction ........................................... 2–111

2.8.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–111

2.8.3 Summary of Previous Analyses and Assessments .............. 2–111

2.8.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–111

2.8.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–115

2.9 Electron Tubes ................................................ 2–119

2.9.1 Introduction ........................................... 2–119

2.9.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–119

2.9.3 Summary of Previous Analyses and Assessments .............. 2–120

2.9.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–123

2.9.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–128

2.10 Ionizing Radiation Measurement Instruments Containing,

for Purposes of Internal Calibration or Standardization,

One or More Sources of Byproduct Material ....................... 2–139

2.10.1 Introduction .......................................... 2–139

2.10.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–140

2.10.3 Summary of Previous Analyses and Assessments ............. 2–141

2.10.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–141

2.10.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–145

2.11 Spark Gap Irradiators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–151

2.11.1 Introduction .......................................... 2–151

2.11.2 Description of Item . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–151

2.11.3 Summary of Previous Assessments . . . . . . . . . . . . . . . . . . . . . . . . 2–151

2.11.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–152

2.11.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–154

2.12 Resins Containing 46Sc for Sand Consolidation in Oil Wells ............. 2–161

2.12.1 Introduction .......................................... 2–161

2.12.2 Description of Item . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–161

2.12.3 Summary of Previous Analyses and Assessments ............. 2–162

2.12.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–162
vii

2.13 Quantities of Byproduct Material .................................. 2–163

2.13.1 Introduction .......................................... 2–163

2.13.2 Description of Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–164

2.13.3 Summary of Previous Analyses and Assessments ............. 2–165

2.13.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–166

2.13.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–179

2.14 Self-Luminous Products ........................................ 2–193

2.14.1 Introduction .......................................... 2–193

2.14.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–194

2.14.3 Summary of Previous Assessments . . . . . . . . . . . . . . . . . . . . . . . . 2–196

2.14.4 Current Assessment for Wristwatches . . . . . . . . . . . . . . . . . . . . . . 2–199

2.14.5 Current Assessment for Gun Sights . . . . . . . . . . . . . . . . . . . . . . . . 2–206

2.14.6 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–210

2.15 Gas and Aerosol Detectors ...................................... 2–215

2.15.1 Introduction .......................................... 2–215

2.15.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–216

2.15.3 Summary of Previous Analyses and Assessments ............. 2–219

2.15.4 Present Exemption Analysis for Smoke Detectors . . . . . . . . . . . . . 2–220

2.15.5 Present Exemption Analysis for Chemical Detectors . . . . . . . . . . . 2–223

2.15.6 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–227

3 EXEMPTIONS FOR SOURCE MATERIAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–1

3.1 Properties of Source Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–1

3.1.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–1

3.1.2 Definition of Source Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–1

3.1.3 Decay Data for Source Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–2

3.1.4 Dosimetry Data for Source Materials . . . . . . . . . . . . . . . . . . . . . . . . . . 3–4

3.1.5 Isotopic Composition of Chemically Separated Thorium . . . . . . . . . . . 3–5

3.2 Chemical Mixture, Compound, Solution, Alloy Containing

<0.05 Percent by Weight of Source Material . . . . . . . . . . . . . . . . . . . . . . . . 3–19

3.2.1 Introduction ............................................ 3–19

3.2.2 Description of Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–20

3.2.3 Summary of Previous Analyses and Assessments ............... 3–24

3.2.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–32

3.2.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–37

3.3 Unrefined and Unprocessed Ore Containing Source Material .............. 3–61

3.3.1 Introduction ............................................ 3–61

3.3.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–61

3.3.3 Summary of Previous Analyses and Assessments ............... 3–62

3.3.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–64

3.3.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–69

3.4 Incandescent Gas Mantles ........................................ 3–75

3.4.1 Introduction ............................................ 3–75

3.4.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–75

3.4.3 Summary of Previous Analyses and Assessments ............... 3–76

3.4.4 Present Exemption Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–80

3.4.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–87
viii

3.5 Thorium in Vacuum Tubes ........................................ 3–95

3.5.1 Introduction ............................................ 3–95

3.5.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–95

3.5.3 Summary of Previous Analyses and Assessments ............... 3–96

3.5.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–96

3.5.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–101

3.6 Welding Rods Containing Thorium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–103

3.6.1 Introduction ........................................... 3–103

3.6.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–103

3.6.3 Summary of Previous Analyses and Assessments .............. 3–104

3.6.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–111

3.6.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–119

3.7 Thorium in Electric Lamps for Illuminating Purposes . . . . . . . . . . . . . . . . . . . 3–123

3.7.1 Introduction ........................................... 3–123

3.7.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–123

3.7.3 Summary of Previous Analyses and Assessments .............. 3–123

3.7.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–125

3.7.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–131

3.8 Thorium in Germicidal Lamps, Sunlamps, and Lamps for 

Outdoor or Industrial Lighting ................................... 3–135

3.8.1 Introduction ........................................... 3–135

3.8.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–135

3.8.3 Summary of Previous Analyses and Assessments .............. 3–136

3.8.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–136

3.8.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–143

3.9 Rare Earth Products Containing Less Than 0.25% by Weight 

of Source Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–147

3.9.1 Introduction ........................................... 3–147

3.9.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–147

3.9.3 Summary of Previous Analyses and Assessments .............. 3–150

3.9.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–156

3.9.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–161

3.10 Thorium in Personnel Neutron Dosimeters .......................... 3–183

3.10.1 Introduction .......................................... 3–183

3.10.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–183

3.10.3 Summary of Previous Analyses and Assessments ............. 3–184

3.10.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–185

3.10.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–189

3.11 Glazed Ceramic Tableware . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–191

3.11.1 Introduction .......................................... 3–191

3.11.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–191

3.11.3 Summary of Previous Analyses and Assessments ............. 3–192

3.11.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–194

3.11.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–199
ix

3.12 Piezoelectric Ceramic Containing Not More Than 

2 Percent by Weight of Source Material . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–209

3.12.1 Introduction .......................................... 3–209

3.12.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–209

3.12.3 Summary of Previous Analyses and Assessments ............. 3–211

3.12.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–211

3.12.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–215

3.13 Glassware . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–217

3.13.1 Introduction .......................................... 3–217

3.13.2 Description of Exempt Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–217

3.13.3 Summary of Previous Analyses and Assessments ............. 3–218

3.13.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–218

3.13.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–223

3.14 Glass Enamel and Glass Enamel Frit Containing Source Material ........ 3–231

3.14.1 Introduction .......................................... 3–231

3.14.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–231

3.14.3 Summary of Previous Analyses and Assessments ............. 3–232

3.14.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–232

3.15 Photographic Film, Negatives, and Prints Containing Uranium or Thorium . . 3–233

3.15.1 Introduction .......................................... 3–233

3.15.2 Description of Item . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–233

3.15.3 Summary of Previous Analyses and Assessments ............. 3–234

3.15.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–234

3.15.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–236

3.16 Finished Tungsten- or Magnesium-Thorium Alloy Products or Parts ...... 3–239

3.16.1 Introduction .......................................... 3–239

3.16.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–239

3.16.3 Summary of Previous Analyses and Assessments ............. 3–240

3.16.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–240

3.16.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–245

3.17 Uranium in Counterweights ..................................... 3–249

3.17.1 Introduction .......................................... 3–249

3.17.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–249

3.17.3 Summary of Previous Analyses and Assessments ............. 3–251

3.17.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–254

3.17.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–259

3.18 Uranium Shielding in Shipping Containers . . . . . . . . . . . . . . . . . . . . . . . . . . 3–269

3.18.1 Introduction .......................................... 3–269

3.18.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–269

3.18.3 Summary of Previous Assessments . . . . . . . . . . . . . . . . . . . . . . . . 3–270

3.18.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–271

3.18.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–272

3.19 Thorium in Finished Optical Lenses ............................... 3–285

3.19.1 Introduction .......................................... 3–285

3.19.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–285

3.19.3 Summary of Previous Analyses and Assessments ............. 3–286

3.19.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–288

3.19.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–295
x

3.20 Aircraft Engine Parts Containing Nickel-Thoria Alloy . . . . . . . . . . . . . . . . . . 3–301

3.20.1 Introduction .......................................... 3–301

3.20.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–301

3.20.3 Summary of Previous Analyses and Assessments ............. 3–302

3.20.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–302

3.20.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–306

3.21 Uranium in Fire Detection Units . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–309

3.21.1 Introduction .......................................... 3–309

3.21.2 Description of Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–309

3.21.3 Summary of Previous Analyses and Assessments ............. 3–310

3.21.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–310

3.21.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–312

4 ITEMS CONSIDERED FOR EXEMPT DISTRIBUTION . . . . . . . . . . . . . . . . . . . . . . . . . . 4–1

4.1 Static Eliminators and Ion Generators Containing Polonium-210 . . . . . . . . . . . . 4–3

4.1.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–3

4.1.2 Description of Items Considered for Exempt Distribution . . . . . . . . . . . 4–4

4.1.3 Summary of Previous Analyses and Assessments . . . . . . . . . . . . . . . . 4–5

4.1.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–7

4.1.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–11

4.2 Beta Backscatter and Transmission Devices .......................... 4–15

4.2.1 Introduction ............................................ 4–15

4.2.2 Description of Items Considered for Exempt Distribution . . . . . . . . . . 4–16

4.2.3 Summary of Previous Analyses and Assessments ............... 4–17

4.2.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–17

4.2.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–25

4.3 Electron Capture Detectors for Gas Chromatographs ................... 4–29

4.3.1 Introduction ............................................ 4–29

4.3.2 Description of Items Considered for Exempt Distribution . . . . . . . . . . 4–30

4.3.3 Summary of Previous Analyses and Assessments ............... 4–30

4.3.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–30

4.3.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–38

4.4 X-ray Fluorescence Analyzers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–41

4.4.1 Introduction ............................................ 4–41

4.4.2 Description of Items Considered for Exempt Distribution . . . . . . . . . . 4–42

4.4.3 Summary of Previous Analyses and Assessments ............... 4–43

4.4.4 Present Exemption Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–43

4.4.5 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–50

4.5 Calibration and Reference Sources ................................. 4–53

4.5.1 Introduction ............................................ 4–53

4.5.2 Description of Items Considered for Exempt Distribution . . . . . . . . . . 4–55

4.5.3 Summary of Previous Analyses and Assessments ............... 4–57

4.5.4 Present Assessment for Calibration and Reference Sources Distributed

Internal to a Device . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–57

4.5.5 Present Analysis for Loose Calibration and Reference Sources .... 4–63

4.5.6 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–70

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . R–1
xi

APPENDIX A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A–0

A.1 GENERIC ACCIDENT METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–1

A.1.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–1

A.1.2 Airborne Concentrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–1

A.1.3 Resuspension Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–2

A.1.4 Release Fractions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–2

A.1.5 Inhalation Intakes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–2

A.1.6 Ingestion Intakes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–3

A.1.7 Radiation Dose Estimation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–3

A.1.8 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–9

A.2 GENERIC DISPOSAL METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–1

A.2.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–1

A.2.2 Description of Disposal Options . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–1

A.2.3 Dose Assessment for Disposal in Landfills . . . . . . . . . . . . . . . . . . . A.2–5

A.2.4 Dose Assessment for Disposal in Incinerators . . . . . . . . . . . . . . . . A.2–23

A.2.5 Dose Assessment for Metal Recycling . . . . . . . . . . . . . . . . . . . . . . A.2–27

A.2.6 Application of Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–28

A.3 GENERIC DISTRIBUTION METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . . . A.3–1

A.3.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–1

A.3.2 Derivation of the Dose Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–2

A.3.3 Application of the Dose Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–3

A.3.4 Commercial Truck . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–4

A.3.5 Warehousing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–5

A.3.6 Retailing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–6

A.3.7 Air Transport . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–6

A.4 OTHER MODELING METHODOLOGIES . . . . . . . . . . . . . . . . . . . . . . . . . . . A.4–1

A.4.1 Estimation of External Photon Dose Due to Bremsstrahlung Produced

by Low-Energy Electrons in Beta Decay . . . . . . . . . . . . . . . . . . . . . A.4–1

A.4.2 Generic Dose Modeling for Sources in Close Proximity to the Body

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.4–6

APPENDIX B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B–0

GLOSSARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B–1

SYMBOLS AND UNITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B–5

INDEX . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I–1

Figures

Figure 3.2.1 Heavy Mineral Concentrate Process Schematic . . . . . . . . . . . . . . . . . . . . . . . 3–39

Tables

Table 1 Highest Individual Annual Doses from Normal Use . . . . . . . . . . . . . . . . . . . . . . . . . xxx

Table 2 Collective Doses from Normal Use . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxxi

Table 2.1.1 Decay Data for Selected Byproduct Materials . . . . . . . . . . . . . . . . . . . . . . . . . . 2–5
xii

Table 2.1.2 Dosimetry Data for Selected Byproduct Materials . . . . . . . . . . . . . . . . . . . . . . . 2–7

Table 2.1.3 Gastrointestinal-Tract Absorption Fractions (f1) and Lung Clearance

Classes for Chemical Compounds of Selected Elements .................... 2–11

Table 2.2.1 Quantities of Byproduct Material Transferred in Irradiated Topaz Gemstones

by Single Materials Licensee . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–35

Table 2.2.2 Quantities of Byproduct Material Transferred in Irradiated Semiconductor

Materials by Single Materials Licensee . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–36

Table 2.2.3 Quantities of Byproduct Material Introduced Into High-Purity Silicon

Semiconductor Materials by Single Materials Licensee ...................... 2–37

Table 2.2.4 Quantities of Byproduct Material Introduced Into Engine Oil or Steel

by Materials Licensees . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–39

Table 2.2.5 Estimated Concentrations of Byproduct Material Introduced Into Baggage

Contents From Neutron Irradiation of 1-kg Masses of Various Elements

in Airport Explosive Detection System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–40

Table 2.2.6 Quantities of Radionuclides Distributed as Exempt Concentrations

of Byproduct Material During 1970 to 1989 ............................... 2–41

Table 2.2.7 Estimates of Dose From Processing and Assembly of Irradiated Silicon

Semiconductor Materials ............................................. 2–42

Table 2.2.8 Estimates of External Dose Rates Near Steel Slab Contaminated 

With 60Co . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–43

Table 2.2.9 Estimates of External Dose Rates From Exposure to Various Radionuclides

in Baggage Contents Irradiated by Neutrons in Airport Explosive 

Detection System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–44

Table 2.2.10 Estimates of Dose From Exposure to Baggage Contents Irradiated

by Neutrons in Airport Explosive Detection System . . . . . . . . . . . . . . . . . . . . . . . . . 2–45

Table 2.2.11 Estimates of External Dose While Wearing Irradiated Topaz Gemstones

Containing Exempt Concentrations of Byproduct Material .................... 2–46

Table 2.2.12 Comparison of Existing Exempt Concentrations Calculated to Result in

5 Rem Internal Annual Dose Based on Revised 10 CFR 20 .................. 2–47

Table 2.2.13 Summary of Potential External Radiation Doses From Use of Products or

Materials Containing Exempt Concentrations of Byproduct Material ............ 2–48

Table 2.3.1 Summary of Potential Radiation Doses From Timepieces Containing

3

H in Paint . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–66

Table 2.3.2 Summary of Model and Potential Individual and Collective Effective Dose

Equivalents for Distribution of Timepieces Containing 3

H in Paint . . . . . . . . . . . . . . 2–67

Table 2.3.3 Summary of Potential Radiation Doses From Timepieces Containing

147Pm in Painta ..................................................... 2–69

Table 2.3.4 Summary of Model and Potential Individual and Collective Effective Dose

Equivalents for Distribution of Timepieces Containing 147Pm in Paint . . . . . . . . . . . 2–70

Table 2.4.1 Hypothetical Doses From Distribution of Automobiles With Lock Illuminators

Containing 3

H or 147Pm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–79

Table 2.4.2 Assumptions and Estimated Doses From Use of Automobiles

With Lock Illuminators Containing 3

H.................................... 2–80

Table 2.4.3 Assumptions and Estimated Doses From Use of Automobiles

With Lock Illuminators Containing 147Pm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–81

Table 2.4.4 Assumptions and Doses From Exposure to Disposal of

Automobile Lock Illuminators Containing 3

H .............................. 2–82

Table 2.4.5 Assumptions and Doses From Exposure to Disposal of

Automobile Lock Illuminators Containing 147Pm . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–83
xiii

Table 2.4.6 Hypothetical Radiation Doses From Automobile Lock

Illuminators Containing 3H ............................................ 2–84

Table 2.4.7 Hypothetical Radiation Doses From Automobile Lock

Illuminators Containing 147Pm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–85

Table 2.5.1 Estimated Doses From Distribution and Use of Balances of Precision

Containing 3

H ..................................................... 2–91

Table 2.5.2 Estimated Doses From Disposal of Balances of Precision

Containing 3

H ..................................................... 2–92

Table 2.5.3 Summary of Radiation Doses From 10,000 Balances of Precision

Containing 3

H ..................................................... 2–93

Table 2.6.1 Estimated Doses From Distribution of 1 Million Automobiles With Shift Quadrants

Containing 3

H .................................................... 2–100

Table 2.6.2 Assumptions and Estimated Doses From Use of 1 Million Automobiles

With Shift Quadrants Containing 3

H ................................... 2–101

Table 2.6.3 Assumptions and Doses From Exposure to Disposal of 1 Million 

Automobile Shift Quadrants Containing 3

H .............................. 2–102

Table 2.6.4 Potential Radiation Doses From 1 Million Automobile Shift Quadrants

Containing 3

H .................................................... 2–103

Table 2.7.1 Potential Annual Radiation Doses From Marine Compasses and

Navigational Instruments ............................................ 2–110

Table 2.8.1 Estimated Doses From Distribution of 10,000 Thermostat

Dials and Pointers Containing 3

H...................................... 2–116

Table 2.8.2 Estimated Doses From Installation, Servicing, and Routine Use of 

10,000 Thermostat Dials and Pointers Containing 3

H ...................... 2–117

Table 2.8.3 Summary of Radiation Doses From 10,000 Thermostat Dials and Pointers

Containing 3

H .................................................... 2–118

Table 2.9.1 Average Annual Distribution of Electron Tubes Containing Byproduct

Material During the Period of 1970 to 1986 .............................. 2–129

Table 2.9.2 Potential Radiation Doses From Distribution and Transport of Electron

Tubes Containing Average Quantities of Byproduct Materials ................ 2–130

Table 2.9.3 Potential Radiation Doses From Routine Use of Electron Tubes

Containing Average Quantities of Byproduct Materials ..................... 2–131

Table 2.9.4 Potential Radiation Doses to Individuals From Routine Use of Electron 

Tubes Containing the Maximum Quantity of a Byproduct Material Allowed 

by Exemption . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–132

Table 2.9.5 Potential Individual Doses From Disposal of Electron Tubes Containing 

Average Quantities of Byproduct Material ............................... 2–133

Table 2.9.6 Potential Collective Doses From Disposal of Electron Tubes Containing 

Average Quantities of Byproduct Material ............................... 2–135

Table 2.9.7 Potential Radiation Doses From Electron Tubes Containing 

Byproduct Material ................................................. 2–137

Table 2.10.1 Reported Quantities of Byproduct Material and Numbers of Units

Distributed From 1985 to 1995 for Use as Internal Calibration or

Standardization Sources ............................................ 2–147

Table 2.10.2 Estimated Annual Quantity of Byproduct Material and Number of Sources

Distributed in Each Ionizing Radiation Monitoring Instrument Classification . . . . . 2–148
xiv

Table 2.10.3 Estimated Effective Dose Equivalents From Distribution, 

Use, and Disposal of Ionizing Radiation Measuring Instruments Containing 

One or More Sources of Byproduct Material for Purposes of Internal

Calibration or Standardization ........................................ 2–149

Table 2.11.1 Exposure Conditions for Distribution and Transport of 1,000 Irradiators

and Corresponding Doses ........................................... 2–156

Table 2.11.2 Exposure Conditions for Installation and Maintenance of 1,000 Irradiators

and Corresponding Doses ........................................... 2–157

Table 2.11.3 Exposure Conditions for Disposal of 1,000 Irradiators

and Corresponding Doses ........................................... 2–158

Table 2.11.4 Hypothetical Radiation Doses From Spark Gap Irradiators 

Containing 60Co . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–159

Table 2.13.1 Total Reported Quantities of Radionuclides in Byproduct Material

Distributed During 1989 to 1995 ...................................... 2–181

Table 2.13.2 Estimates of Potential Radiation Doses From External, Inhalation, and 

Ingestion Exposure to Quantities of Byproduct Material Authorized for 

Exempt Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–182

Table 2.13.3 Summary of Potential Radiation Doses to the Public From Use

of Quantities of Byproduct Material Authorized for Exempt Distribution ......... 2–190

Table 2.14.1 Potential Radiation Doses From Self-Luminous Wristwatches

Containing 3

H .................................................... 2–211

Table 2.14.2 Comparison of Estimated Maximum Individual Doses From Self-Luminous

Wristwatches Containing 3

H and Limiting Organ Doses From 10 CFR 32.23 .... 2–212

Table 2.14.3 Potential Radiation Doses From Self-Luminous Gun Sights 

Containing 3

H .................................................... 2–213

Table 2.14.4 Comparison of Estimated Maximum Individual Doses From 

Self-Luminous Gun Sights Containing 3

H and Limiting Organ Doses 

From 10 CFR 32.23 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–214

Table 2.15.1 Potential Radiation Doses From Smoke Detectors Containing 241Am . . . . . 2–229

Table 2.15.2 Exposure Conditions and Radiation Doses From Routine Home Use

of Two Smoke Detectors Containing 0.037 MBq (1 Ci) of 241Am Each . . . . . . . . 2–230

Table 2.15.3 Potential Radiation Doses From Routine Home Use of 10 Million 

Smoke Detectors Containing 0.037 MBq (1 Ci) Each of 241Am . . . . . . . . . . . . . . 2–231

Table 2.15.4 Comparison of Estimated Maximum Individual Doses From Smoke

Detectors Containing 37 kBq (1 Ci) of 241Am and Limiting Organ Doses 

From 10 CFR 32.27 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–232

Table 2.15.5 Potential Radiation Doses From Chemical Detectors

Containing 241Am . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–234

Table 2.15.6 Comparison of Estimated Maximum Individual Doses 

From Chemical Detectors Containing 5.9 MBq (0.16 mCi) of 241Am 

and Limiting Organ Doses From 10 CFR 32.27 ........................... 2–235

Table 2.15.7 Comparison of Estimated Maximum Individual Doses 

From Chemical Detectors Containing 0.37 GBq (10 mCi) of 63Ni 

and Limiting Organ Doses From 10 CFR 32.27 ........................... 2–237

Table 3.1.1 Mass and Activity Abundances of Naturally Occurring Isotopes in Chemically 

Separated Uranium and Thorium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–8

Table 3.1.2 Decay Data for Naturally Occurring Isotopes of Uranium 

and Their Radioactive Decay Products . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–9
xv

Table 3.1.3 Decay Data for Naturally Occurring Isotopes of Thorium 

and Their Radioactive Decay Products .................................. 3–11

Table 3.1.4 Activity of Radioactive Decay Products of Uranium vs. Time

After Chemical Separation of Uranium .................................. 3–12

Table 3.1.5 Activity of Radioactive Decay Products of Thorium vs. Time

After Chemical Separation of Thorium ................................... 3–13

Table 3.1.6 Dosimetry Data for Naturally Occurring Isotopes of Uranium

and Their Radioactive Decay Products .................................. 3–14

Table 3.1.7 Dosimetry Data for Naturally Occurring Isotopes of Thorium

and Their Radioactive Decay Products .................................. 3–17

Table 3.1.8 Gastrointestinal-Tract Absorption Fractions (f1) and Lung Clearance Classes

for Chemical Compounds of Uranium, Thorium, and Their Decay Products ...... 3–18

Table 3.2.1 Source Material Concentrations Associated With 0.05 Percent Weight 

of Source . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–40

Table 3.2.2 Naturally Occurring Radioactivity Related to Mineral Resources . . . . . . . . . . . 3–41

Table 3.2.3 Uranium and Thorium Concentrations in Common Building Materials ....... 3–43

Table 3.2.4 Examples of Mineral-Derived Products .............................. 3–44

Table 3.2.5 Mineral Extraction Processing Methods .............................. 3–45

Table 3.2.6 Uranium in Coal and Ash ......................................... 3–46

Table 3.2.7 Typical Radionuclide Concentrations in Phosphate Ore, 

Phosphate Slag, and Phosphogypsum .................................. 3–47

Table 3.2.8 Uranium Concentrations in Dental Products and Estimated Doses ......... 3–48

Table 3.2.9 Measured Air Concentrations at Plant B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–49

Table 3.2.10 Typical Activity Concentrations of Materials in Use . . . . . . . . . . . . . . . . . . . . 3–50

Table 3.2.11 Individual Exposures: External Irradiation and Dust Inhalation 

Exposure Routes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–51

Table 3.2.12 Radionuclide Concentrations Used in Storage and Disposal Assessment . . . 3–52

Table 3.2.13 Individual Effective Dose Equivalent to Workers From Storage

or Disposal of Wastes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–53

Table 3.2.14 Individual Effective Dose Equivalent to Members of the Public From 

Storage or Disposal of Wastes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–54

Table 3.2.15 Collective Effective Dose Equivalents From Storage or 

Disposal of Wastes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–55

Table 3.2.16 Estimated Doses From Exposures to a Coal Ash Pile and 

Coal Ash Product .................................................. 3–56

Table 3.2.17 Gamma Irradiation Dose of Individuals and Populations of Pocatello 

and Soda Springs, ID ............................................... 3–57

Table 3.2.18 Individual and Collective Effective Dose Equivalents Attributed

to the Disposal of Zircon Sands ........................................ 3–58

Table 3.2.19 Summary of Effective Dose Equivalents According to Percent Weight . . . . . 3–59

Table 3.3.1 Individual Doses to Workers and the Public From Unrefined Uranium 

and Thorium Ore ................................................... 3–70

Table 3.3.2 Individual Doses Attributed to Disposal of Uranium and 

Thorium Ore Samples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–71

Table 3.3.3 Collective Doses Attributed to Disposal of Uranium and 

Thorium Ore Samples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–72

Table 3.3.4 Summary of Doses Associated With Unrefined Uranium and 

Thorium Ore . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–73
xvi

Table 3.4.1 Radioactivity in a Gas Mantle Containing 250 mg of Natural Thorium or

Thorium That Has Been Chemically Extracted From Ore and Aged for 6 Months . . 3–89

Table 3.4.2 Potential Radiation Doses From Incandescent Gas Mantles

Containing Thorium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–90

Table 3.4.3 Summary of Potential Radiation Doses From Changing Two Mantles ....... 3–91

Table 3.4.4 Exposure Conditions for a 2-Night Camping Trip 

for an Avid Camper and Family ........................................ 3–92

Table 3.4.5 Estimates of Annual Effective Dose Equivalent 

for One-Time and Avid Campers ....................................... 3–94

Table 3.5.1 Potential Radiation Doses From Vacuum Tubes Containing Thorium ...... 3–102

Table 3.6.1 Individual and Collective Doses From Distribution and Transport 

of Thoriated-Tungsten Welding Rods Obtained in Previous Assessments ...... 3–120

Table 3.6.2 Summary of Potential Radiation Doses From Use

of Welding Rods Containing Thorium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–121

Table 3.7.1 Summary of Potential Radiation Doses From Use 

of Electric Lamps Containing Thorium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–133

Table 3.8.1 Summary of Potential Radiation Doses From Use of Thorium-Containing

Germicidal Lamps, Sunlamps, and Lamps for Outdoor or Industrial Lighting .... 3–145

Table 3.9.1 Products Derived From Rare Earth Elements ........................ 3–163

Table 3.9.2 Distribution of Rare Earth Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–168

Table 3.9.3 Cerium-Containing Materials and Their Applications ................... 3–169

Table 3.9.4 Annual Exposures to American Potash and Chemical Corporation

Personnel Working With Rare Earth Products ............................ 3–170

Table 3.9.5 Air Concentrations of Thorium in Rare Earth Processing Plant . . . . . . . . . . . 3–171

Table 3.9.6 Airborne Radioactivity Levels Across All Mineral Sands Industry .......... 3–172

Table 3.9.7 1992 Molycorp Personnel TLD Results ............................. 3–173

Table 3.9.8 Molycorp Plant Area TLD Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–174

Table 3.9.9 Environmental TLD Measurements Associated With Molycorp

Lead and Iron Filter Cake Project ..................................... 3–175

Table 3.9.10 Radium Concentrations in Slag, Metal Additives,

and Cerium Oxide Product .......................................... 3–176

Table 3.9.11 Individual Effective Dose Equivalents From Storage or Disposal

of Rare Earth Byproducts (Monazite Based) ............................. 3–177

Table 3.9.12 Annual Collective Effective Dose Equivalent From Storage 

or Disposal of Rare Earth Byproducts (Monazite Based) .................... 3–178

Table 3.9.13 Average Concentration of Radionuclides in Selected Materials .......... 3–179

Table 3.9.14 Summary of Annual Effective Dose Equivalent Due to Landfill Disposal

of Cerium Concentrate (0.25% by Weight of Source Material as Thorium) . . . . . . 3–180

Table 3.9.15 Summary of Doses to Industrial Workers and Other Members of the Public

From the Handling and Use of Rare Earth Products Containing Source Material . 3–181

Table 3.10.1 Potential Radiation Doses From Personnel Neutron Dosimeters

Containing Thorium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–190

Table 3.11.1 Source Representations of Ceramic Tableware . . . . . . . . . . . . . . . . . . . . . 3–201

Table 3.11.2 Absorbed Beta Dose Rates at 7 mg/cm2

 in Tissue at Selected Distances

From the Surface of a 25-cm- (10-in-) Diameter Disk of 0.025-cm-Thick Ceramic 

Glaze Containing 20% by Weight of 20-Year-Old Natural Uranium . . . . . . . . . . . . 3–202

Table 3.11.3 Effective Dose Equivalent Rates Due to Photon Irradiation at Selected

Distances From Pieces of Ceramic Tableware Coated With Glaze

Containing 20% by Weight of 20-Year-Old Natural Uranium . . . . . . . . . . . . . . . . . 3–203
xvii

Table 3.11.4 Effective Dose Equivalent Rates Due to Photon Irradiation at Selected

Distances From Arrays of Ceramic Tableware Coated With Glaze

Containing 20% by Weight of 20-Year-Old Natural Uranium . . . . . . . . . . . . . . . . . 3–204

Table 3.11.5 Summary of the Distribution Model and Potential Individual and

Collective Effective Dose Equivalents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–205

Table 3.11.6 Summary of Home Use Model and Potential Individual

Effective Dose Equivalents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–206

Table 3.11.7 Summary of Potential Radiation Doses From Uranium-Containing,

Glazed Ceramic Tableware . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–207

Table 3.12.1 Potential Radiation Doses From Piezoelectric Devices 

Containing Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–216

Table 3.13.1 Absorbed Dose Rates at 7 mg/cm2

 in Tissue at Selected Distances From the

Surface of a 0.64-cm-Thick, 25-cm- (10-in-) Diameter Glass Disk Containing

10% by Weight of 20-Year-Old Natural Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . 3–224

Table 3.13.2 Effective Dose Equivalent Rates Due to Photon Irradiation at Selected

Distances From Pieces of Glassware Containing 10% by Weight

of 20-Year-Old Natural Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–225

Table 3.13.3 Distribution Model and Potential Individual and Collective Effective Dose

Equivalents (EDEs) for Drinking Glasses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–226

Table 3.13.4 Distribution Model and Potential Individual and Collective Effective Dose

Equivalents (EDEs) for Decorative Objects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–227

Table 3.13.5 Summary of Home Use Model and Potential Individual Effective

Dose Equivalents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–228

Table 3.13.6 Summary of Potential Radiation Doses From Glassware 

Containing Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–229

Table 3.15.1 Potential Radiation Doses From Photographs Containing 

Natural Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–237

Table 3.16.1 Comparison of Materials, Ages, and Dose Rates With Distances ........ 3–247

Table 3.16.2 Summary of Potential Radiation Doses From Use of Finished 

Tungsten- or Magnesium-Thorium Alloy Products or Parts .................. 3–248

Table 3.17.1 Use of DU Counterweights in Domestic Aircraft ..................... 3–260

Table 3.17.2 Exposure Rates for DU Counterweights ........................... 3–261

Table 3.17.3 Installation of DU Counterweights in Aircraft—

Occupational Dosimetry Study ....................................... 3–262

Table 3.17.4 Estimated Annual Effective Dose Equivalents—Wing and Tail Assembly . . 3–263

Table 3.17.5 Collective Effective Dose Equivalents for Aircraft Crew and 

Flight Attendants .................................................. 3–264

Table 3.17.6 Collective Effective Dose Equivalent Estimates for Aircraft Passengers . . . 3–265

Table 3.17.7 Summary of Exposure Rates Attributed to 33-lb Counterweights ........ 3–266

Table 3.17.8 Individual and Collective Effective Dose Equivalent Summary .......... 3–267

Table 3.18.1 Shipping Cask Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–274

Table 3.18.2 Radiographic Exposure Devices ................................. 3–275

Table 3.18.3 Source Changers ............................................ 3–276

Table 3.18.4 Exposure Rate Measurement of Depleted Uranium Sheet with Shielding

Exposure Rate Measurements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–277

Table 3.18.5 Dosimetry Data From Assembly of Radiographic Device Container ...... 3–278

Table 3.18.6 Annual Effective Dose Equivalents From the Distribution

and Transport of Shipping Packages Containing Depleted Uranium ........... 3–279
xviii

Table 3.18.7 Effective Dose Equivalent Estimates for the Disposal of Shipping

Packages Containing Depleted Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–281

Table 3.18.8 Individual and Collective Effective Dose Equivalents Summary 

for Shipping Packages Containing Depleted Uranium . . . . . . . . . . . . . . . . . . . . . . 3–282

Table 3.19.1 Summary of Measured Photon Dose Rates From Cameras Containing

Thoriated Optical Glass . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–297

Table 3.19.2 Summary of Potential Radiation Doses From Use of Finished Optical

Lenses Containing Thorium in Television Cameras ........................ 3–298

Table 3.19.3 Summary of Potential Radiation Doses From Use of Finished Optical

Lenses Containing Thorium in 35-mm Photographic Cameras ............... 3–299

Table 3.20.1 Summary of Potential Radiation Doses From Use of Finished

Aircraft Engine Parts Containing Nickel-Thoria Alloy . . . . . . . . . . . . . . . . . . . . . . . 3–308

Table 3.21.1 Dose Estimates for Routine Exposure From Distribution and Transport . . . 3–313

Table 3.21.2 Dose Estimates for Routine Exposure for Household Members ......... 3–314

Table 3.21.3 Summary of Effective Dose Equivalents for Uranium in Fire Detectors . . . . 3–315

Table 4.1.1 Potential Radiation Doses From Static Eliminators Containing 210Po . . . . . . . . 4–12

Table 4.2.1 Summary of Radionuclides, Source Activities, and Dose Rates Near

Beam Port of Example Existing Backscatter and Transmission Devices ......... 4–26

Table 4.2.2 Potential Radiation Doses From Beta Backscatter and Transmission

Devices Using 85Kr, 106Ru, and 204Tl Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–27

Table 4.3.1 Potential Radiation Doses From Electron Capture Detectors 

Using 63Ni and 3

H Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–39

Table 4.4.1 Potential Radiation Doses From X-Ray Fluorescence Analyzers

Using 55Fe and 109Cd Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–51

Table 4.5.1 Estimated Annual Distribution of Calibration and Reference Sources

Contained Internal to Devices ......................................... 4–73

Table 4.5.2 Estimated Annual Distribution of Loose Calibration and Reference Sources . . 4–73

Table 4.5.3 Potential Radiation Doses From Calibration and Reference Sources

Distributed Internal to a Device . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–74

Table 4.5.4 Potential Radiation Doses From Loose Calibration and Reference Sources . . 4–75

Table A.1.1 Factors for Estimating Release of Respirable Size Particles 

During a Spill or Fire . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–11

Table A.1.2 Enclosure Volumes and Ventilation Rates . . . . . . . . . . . . . . . . . . . . . . . . . A.1–12

Table A.1.3 Average Concentrations in Air and Inhalation Intakes of Radioactive

Material During the 30-Minute Period Following a Fire . . . . . . . . . . . . . . . . . . . . A.1–13

Table A.1.4 Radiation Dose-to-Source Ratios for a Transportation

Accident Involving Fire . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–14

Table A.1.5 Radiation Dose-to-Source Ratios for a Warehouse Fire . . . . . . . . . . . . . . . A.1–16

Table A.1.6 Radiation Dose-to-Source Ratios for a Residential Fire . . . . . . . . . . . . . . . A.1–18

Table A.1.7 Average Concentrations in Air and Individual Intakes From Inhalation 

During Various Exposure Times Following a Spill in a Laboratory-Type Room . . A.1–20

Table A.1.8 Radiation Dose-to-Source Ratios for a Spill of a Liquid 

or Powder in a Laboratory-Type Room . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1–21

Table A.1.9 Radiation Dose-to-Source Ratios for Crushing of Glass Tubes . . . . . . . . . . A.1–23

Table A.2.1 Individual Dose-to-Source Ratios for Exposure to Waste

Collectors at Municipal Landfills . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–32

Table A.2.2 Collective Dose-to-Source Ratios for Exposure to Waste

Collectors at Municipal Landfills . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–34
xix

Table A.2.3 Individual Dose-to-Source Ratios for Exposure to Workers 

at Municipal Landfills . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–36

Table A.2.4 Collective Dose-to-Source Ratios for Exposure to Workers

at Municipal Landfills . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–38

Table A.2.5 Individual and Collective Dose-to-Source Ratios for Exposure to Off-Site

Residents at Municipal Landfills Due to Airborne Releases During

Landfill Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–40

Table A.2.6 Solid and Solution Distribution Coefficients (Kd) for Elements . . . . . . . . . . . A.2–42

Table A.2.7 Individual and Collective Dose-to-Source Ratios for Exposure to Off-Site

Residents at Municipal Landfills Due to Releases to Groundwater . . . . . . . . . . . A.2–43

Table A.2.8 Individual Dose-to-Source Ratios for Exposure to Future On-Site

Residents at Municipal Landfills . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–45

Table A.2.9 Collective Dose-to-Source Ratios for Exposure to Future On-Site

Residents at Municipal Landfills . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–47

Table A.2.10 Individual Dose-to-Source Ratios for Exposure to Waste

Collectors at Municipal Incinerators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–49

Table A.2.11 Collective Dose-to-Source Ratios for Exposure to Waste

Collectors at Municipal Incinerators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–51

Table A.2.12 Individual Dose-to-Source Ratios for Exposure to Workers

at Municipal Incinerators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–53

Table A.2.13 Collective Dose-to-Source Ratios for Exposure to Workers

at Municipal Incinerators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–55

Table A.2.14 Individual and Collective Dose-to-Source Ratios for Exposure to Off-Site

Residents at Municipal Incinerators Due to Airborne Releases During

Incinerator Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2–57

Table A.2.15 Individual Dose-to-Source Ratios for Exposure to Slag Workers

at Metal Smelter or User of Automobile Containing Recycled Metal . . . . . . . . . . A.2–59

Table A.2.16 Individual Dose-to-Source Ratios for Exposure to Off-Site Residents

at Metal Smelter Due to Airborne Releases During Operations . . . . . . . . . . . . . . A.2–61

Table A.3.1 Highly Exposed (Package Near Driver) Individual DFs

for Commercial Truck Transport of Byproduct Material . . . . . . . . . . . . . . . . . . . . . A.3–7

Table A.3.2 Average (Package in Center of Cargo Area) Individual DFs 

for Commercial Truck Transport of Byproduct Material . . . . . . . . . . . . . . . . . . . . . A.3–8

Table A.3.3 Collective DFs for Commercial Truck Transport 

of Byproduct Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–9

Table A.3.4 Highly Exposed (Package Near Driver) Individual DFs for Commercial Truck

Transport of Source Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–10

Table A.3.5 Average (Package in Center of Cargo Area) Individual DFs 

for Commercial Truck Transport of Source Material . . . . . . . . . . . . . . . . . . . . . . A.3–11

Table A.3.6 Collective DFs for Commercial Truck Transport of Source Material . . . . . . A.3–12

Table A.3.7 Highly Exposed Individual DFs for Air Transport, 

Warehousing, and Retailing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–13

Table A.3.8 Collective DFs for Air Transport, Warehousing, and Retailing . . . . . . . . . . A.3–14

Table A.3.9 Example Estimation of Individual and Collective Effective 

Dose Equivalents (EDEs) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–15

Table A.3.10 Steps, Types and Numbers of Persons Exposed, and Exposure Conditions

Involved in Shipment of Goods Via Express Truck Delivery . . . . . . . . . . . . . . . . A.3–16

Table A.3.11 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Local Truck Delivery . . . . . . . . . . A.3–19
xx

Table A.3.12 Steps, Types and Numbers of Persons Exposed, and Exposure

Conditions Involved in Shipment of Goods Via Regional Truck Delivery . . . . . . . A.3–21

Table A.3.13 Exposure Conditions for a Large Warehouse . . . . . . . . . . . . . . . . . . . . . . A.3–24

Table A.3.14 Exposure Conditions for a Medium-Sized Warehouse . . . . . . . . . . . . . . . A.3–25

Table A.3.15 Exposure Conditions for a Large Retail Store . . . . . . . . . . . . . . . . . . . . . . A.3–26

Table A.3.16 Exposure Conditions for a Small Retail Store . . . . . . . . . . . . . . . . . . . . . . A.3–27

Table A.3.17 Exposure Conditions for an Air-Freight Terminal . . . . . . . . . . . . . . . . . . . A.3–28

Table A.3.18 Exposure Conditions in an Airplane . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3–30

Table A.4.1 Comparison of External Doses Due to Bremsstrahlung for Exposure

to Contaminated Ground Surface Calculated Using the CONDOS II Code 

With External Dose Coefficients Given in Federal Guidance Report No. 12 . . . . . A.4–9

Table A.4.2 Dose Reduction Factors Applied to Estimates

of External Dose Due to Bremsstrahlung Obtained

Using the CONDOS II Code . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.4–10
xxi

EXECUTIVE SUMMARY

1 Purpose

The primary purpose of this report is to present a systematic assessment of potential individual

and collective (population) radiation doses associated with the current exemptions from

licensing. The results of this study are intended to provide an assessment upon which the NRC

can review and examine the radiological impact of current exemptions and determine if

regulatory actions may be needed for ensuring public health and safety.

This report also presents an assessment of potential radiological impacts associated with

selected products containing byproduct material which currently may be used under a general

or a specific license and may be potential candidates for exemption from licensing

requirements.

2 General Approach to Dose Assessments

The dose assessments were, in general, based on reasonable assumptions taking into

consideration the provisions of the exemptions. Establishing exposure assumptions for some of

the exemptions was difficult, mainly because of the absence of reliable data on actual use of

the exemptions by individuals either in the workplace or the general environment. Actual data

would be needed to establish realistic dose estimates.

The doses reported are in effective dose equivalent (EDE)/yr to the average member of the

critical group. These EDE/yr estimates include the contribution from exposures to the skin

expressed in shallow-dose equivalent (SDE)/yr. Rarely is the SDE/yr a significant contributor to

risk, expressed as EDE/yr. The SDE/yr is only reported in cases where almost all of the

exposure is to the skin because the radiation emission is too weak to penetrate the skin, or

cases in which the source is in direct contact with the skin, or both. Dose calculation

methodology of International Commission on Radiological Protection (ICRP) 26 and 30 have

been used in order to be consistent with 10 CFR Part 20 methodology.

The final dose estimates (individual and collective) have been rounded to one (1) significant

figure, to reflect the accuracy of the modeling. Also, when the calculated individual dose is less

than 1×105

 millisievert (mSv) (<0.001 mrem), the dose has been presented as a less than

value (i.e., less than 1×105

 mSv (<0.001 mrem)). However, for purposes of estimating the

collective dose, the actual calculated value has been used.

2.1 Dose Assessments for Normal Life Cycle

In this study, individual and collective doses were estimated for the normal life cycle of a

particular product or material, covering distribution and transport, intended or expected routine

use, and disposal occurring over a 1 year time period. Distribution and transport could involve,

for example, exposure to individuals during shipments from licensed manufacturers through

distribution networks to retail stores. Routine use involves exposure to individuals during the

process, handling and day-to-day use of the applicable products. The different methods of

disposal considered in this assessment include evaluating exposure to individuals due to
xxii

placement in municipal landfills, incineration, and, to a limited extent, recycling. Actual, or

expected, quantities of radioactive material in products and materials were used for the dose

estimates, when known; otherwise, the maximum allowed under the exemption was used.

2.2 Dose Assessments for Accidents or Misuse

In this study, individual doses were estimated for accidents involving fires, spills, and accidental

dispersion of products and materials. Scenarios for misuse of products or materials generally

were defined on a case-by-case basis and, in most situations, involve the direct handling of a

product over an extended time period. Collective doses were not estimated in the assessments

of accidents or misuse.

In developing and implementing scenarios for accidents and misuse, the intent is to use

scenarios that reasonably could occur, albeit with a substantially lower probability than

scenarios describing the normal life cycle of a product or material, and based on reasonable but

somewhat conservative parameter values. In some assessments, especially those involving

potential misuse of products or materials, unlikely scenarios may have been assumed in order

to obtain bounding estimates of dose.

3 Summary of Results

3.1 Assessments of Current Exemptions for Byproduct Material

3.1.1 Individual Doses During Normal Life Cycle

The estimates of individual dose during the normal life cycle of a product or material associated

with the current exemptions for byproduct material range from less than 1×105

 mSv/yr

(<0.001 mrem/yr) to 0.2 mSv/yr (20 mrem/yr).

The estimated individual doses equal or exceed 0.1 mSv/yr (10 mrem/yr) for two (2)

exemptions:

• 10 CFR 30.15(a)(9): Ionizing radiation measuring instruments containing byproduct

material, and

• 10 CFR 30.15(a)(10): Spark gap irradiators containing 60Co.

In the case of the ionizing radiation measuring instruments, an estimated dose of 0.2 mSv/yr

(20 mrem/yr) would be received by a laboratory technician working with a bench-top instrument. 

The dose a maintenance worker installing and maintaining spark gap irradiators would receive

is estimated to be 0.1 mSv/yr (10 mrem/yr).

The estimated individual doses equal or exceed 0.01 mSv/yr (1 mrem/yr) but are less than

0.1 mSv/yr (<10 mrem/yr) for the following:

• 10 CFR 30.15(a)(1): Timepieces, hand, or dials containing 3

H or 147Pm,

• 10 CFR 30.15(a)(8): Electron tubes containing byproduct material,
xxiii

• 10 CFR 30.18: Exempt quantities of byproduct material, and

• 10 CFR 30.20: Gas and aerosol detectors containing byproduct material.

In the case of timepieces, the dose of 0.09 mSv/yr (9 mrem/yr) was estimated for a driver of a

large regional delivery truck that delivers 3

H timepieces, and the dose from timepieces

containing 147Pm is considerably less. In the case of electron tubes containing byproduct

material, the dose to a worker of 0.05 mSv/yr (5 mrem/yr), would be slightly greater than the

dose of 0.02 mSv/yr (2 mrem/yr) to a user in a home. For quantities of byproduct material

authorized for exempt distribution, the estimated dose of 0.02 mSv/yr (2 mrem/yr) is for

exposure during transport and during laboratory use of calibration sources. For gas and

aerosol detectors, the dose of 0.02 mSv/yr (2 mrem/yr) would apply to the operator of a

portable chemical detector containing 241Am, while the dose from smoke detectors containing

241Am would be 0.01 mSv/yr (1 mrem/yr) for disposal by incineration. The dose to a user in the

home for smoke detectors containing 241Am would be 2×105

 mSv/yr (0.002 mrem/yr).

The estimated individual doses for all the remaining exemptions for byproduct material, for

which a dose assessment was made, are less than 0.01 mSv/yr (<1 mrem/yr).

In addition, certain products distributed for use under the exemptions of 10 CFR 30.19 for

self-luminous products containing 3

H, 85Kr, or 147Pm and 10 CFR 30.20 for gas and aerosol

detectors containing byproduct material must meet dose limits to various parts of the body. In

the case of wristwatches containing 3

H gas, the estimated dose to a small area of skin due to

absorption of 3

H is estimated to be 0.4 mSv/yr (40 mrem/yr), which exceeds the specified safety

criterion of 0.15 mSv/yr (15 mrem/yr) for normal use of these products. This estimate has a

large degree of uncertainty, due to limited data on dosimetric modeling for localized skin doses

from 3

H uptake. Comparable doses would be received in the case of exempt concentrations of

byproduct material (10 CFR 30.14), where the dose to a small area of the skin while wearing

irradiated topaz gemstones is 0.3 mSv/yr (30 mrem/yr), but there is no corresponding safety

criterion for this exemption. The estimated doses in all cases are well below levels for induction

of deterministic effects, and the contribution to the effective dose equivalent is negligible.

Finally, the exemption in 10 CFR 30.21, Radioactive drug: Capsules containing carbon-14 urea

for “in vivo” diagnostic use for humans, was not re-evaluated because this item was recently

added to 10 CFR Part 30, (62 FR 63640, Dec. 2, 1997), and a dose assessment was

performed at that time.

3.1.2 Collective Doses During Normal Life Cycle

The estimates of collective dose during the normal life cycle of a product or material associated

with the current exemptions for byproduct material given in Table 2 range from 0.1 person-Sv

(10 person-rem) to 40 person-Sv (4000 person-rem) for 1 year’s distribution. For two

exemptions, the estimated collective doses equaled or exceeded 10 person-Sv

(1000 person-rem). The collective dose for 10 CFR 30.15(a)(1), timepieces, hand, or dials

containing 3

H or 147Pm, was estimated to be 40 person-Sv (4000 person-rem), predominantly

received by a large number of individuals who wear timepieces (wristwatches). Collective

doses were not estimated for certain exemptions because the product is no longer in

production, was never produced or produced in limited quantity, or is not currently in wide-scale

use.
xxiv

The collective dose for 10 CFR 30.15(a)(8), electron tubes containing byproduct material, is

estimated to be 10 person-Sv (1000 person-rem) over the useful lifetime of 10 years. In this

case, most of the collective dose would be the result of a large number of people exposed to

electron tubes in the home and workplace.

3.1.3 Individual Doses Due to Accidents or Misuse

The estimates of individual dose due to accidents or misuse of a product or material associated

with the current exemptions for byproduct material range from 0.001 mSv (0.1 mrem) to 10 mSv

(1000 mrem). In the cases of microwave receiver protector tubes containing 3H, spark gap

irradiators containing 60Co, and smoke detectors containing 241Am, the estimated doses would

exceed 1 mSv (100 mrem).

When the estimated dose is 0.1 mSv (10 mrem) or greater, it is the result of an accident

causing the release of all or part of the radioactive material (e.g., the crushing of a glass tube or

an abnormal leak) or a scenario for misuse involving ingestion of radioactive material or

carrying of a source in a pocket. The one exception is the dose to a cleanup worker after a

transportation fire involving chemical detectors containing 241Am, where the dose is 0.3 mSv

(30 mrem).

In some of the exemptions for byproduct material, irradiation of localized parts of the body,

including the hands and a small area of the skin, due to misuse also was considered. In only

one case did the estimated dose approach a level for induction of deterministic effects. For the

spark gap irradiator (10 CFR 30.15(a)(10)), the dose to a small area of the skin could approach

1 gray (100 rads) for a serviceman who ignores the caution statement on the package insert

and carries an irradiator for 2000 hours in a pocket during any 1 year.

3.2 Assessments of Current Exemptions for Source Material

3.2.1 Individual Doses During Normal Life Cycle

The estimates of individual dose during the normal life cycle of a product or material associated

with the current exemptions for source material range from less than 1×105

 mSv/yr

(<0.001 mrem/yr) to 40 mSv/yr (4000 mrem/yr). Table 1 shows the ranges for various

exemptions.

The estimated individual doses exceed 10 mSv/yr (1000 mrem/yr) for the following two (2)

exemptions:

10 CFR 40.13(a): Chemical mixture, compound, solution, or alloy containing less than

0.05% by weight source material, and

10 CFR 40.13(c)(1)(vi): Rare earth metals and compounds, mixtures, and products

containing not more than 0.25% by weight source material.

The high estimates in these cases result from the large volumes of exempted material present

in workplaces and the high concentrations of uranium and thorium in these materials. These

doses would be reduced substantially if the affected workers used respiratory protection.
xxv

The estimated individual doses are greater than or equal to 1 mSv/yr (100 mrem/yr) but less

than 10 mSv/yr (<1000 mrem/yr) for the following three (3) exemptions:

• 10 CFR 40.13(b): Unrefined and unprocessed ore containing source material,

• 10 CFR 40.13(c)(1)(i): Incandescent gas mantles containing thorium, and

• 10 CFR 40.13(c)(1)(iii): Welding rods containing thorium.

In the case of unrefined and unprocessed ore, the estimated dose of 3 mSv/yr (300 mrem/yr) to

the truck driver results from the large volume of exempted material that is handled and the

relatively high concentration of uranium in the material. For incandescent gas mantles, the

estimated dose to a person using only gas lanterns for light would be 2 mSv/yr (200 mrem/yr);

the dose to an individual who uses portable camping lanterns would be 0.1 mSv/yr

(10 mrem/yr). For welding rods, the estimated dose of 8 mSv/yr (800 mrem/yr) to a dedicated

grinder of welding rods probably represents an unusual situation that would occur only at

construction sites where many welders are employed.

The estimated individual doses are greater than or equal to 0.1 mSv/yr (10 mrem/yr) but less

than 1 mSv/yr (<100 mrem/yr) for the following five (5) exemptions:

• 10 CFR 40.13(c)(2)(i): Glazed ceramic tableware containing source material,

• 10 CFR 40.13(c)(4): Finished product or part containing tungsten- or

magnesium-thorium alloys with the thorium content of the alloy not exceeding 4% by

weight,

• 10 CFR 40.13(c)(5): Uranium contained in counterweights used in aircraft, rockets,

projectiles and missiles,

• 10 CFR 40.13(c)(7): Finished optical lenses containing thorium,

• 10 CFR 40.13(c)(8): Any finished aircraft engine part containing nickel-thoria alloy with

the thorium content of the alloy not exceeding 4% by weight, and

In the case of glazed ceramic tableware, the estimated dose of 0.5 mSv/yr (50 mrem/yr) would

be to a user of tableware in a home. For the two exemptions for finished products or parts

containing different thorium alloys, the estimated doses would be to individuals who perform

maintenance activities on aircraft engines. For uranium in counterweights, the estimated dose

to the maintenance worker involved in the installation and removal of counterweights from

aircraft would be 0.9 mSv/yr (90 mrem/yr). For finished optical lenses, the dose to an operator

of a television camera is estimated to be 0.2 mSv/yr (20 mrem/yr), which is a factor of 10

greater than the dose of 0.02 mSv/yr (2 mrem/yr) to an avid photographer who uses a 35-mm

photographic camera.

The estimated individual doses are greater than or equal to 0.01 mSv/yr (1 mrem/yr) but less

than 0.1 mSv/yr (<10 mrem/yr) for the following two (2) exemptions:
xxvi

• 10 CFR 40.13(c)(2)(iii): Glassware containing source material, and

• 10 CFR 40.13(c)(6): Natural or depleted uranium used as shielding in shipping

containers.

For glassware, the estimated dose of 0.04 mSv/yr (4 mrem/yr) would be for the driver of a

delivery truck from the manufacturing facility; the estimated dose to a user of these items in a

home would be 0.02 mSv/yr (2 mrem/yr). In the case of shipping containers, the estimated

dose of 0.05 mSv/yr (5 mrem/yr) would be to the individual handling the shipping containers

during loading for air transport. Only the dose from the container, not an enclosed source, was

included in this estimate.

The individual doses associated with all or the remaining exemptions for source material are

less than 0.01 mSv/yr (<1 mrem/yr).

In the exemption for any chemical mixture, compound, solution, or alloy containing less than

0.05% by weight of source material (10 CFR 40.13(a)), irradiations of localized parts of the

body, including the cornea of the eye and the basal mucosa of the mouth, by alpha or beta

particles during normal use could occur. The dose to the cornea of 0.04 Sv/yr (4 rem/yr) from

ophthalmic glass lenses and to the basal mucosa of the mouth of 5 mSv/yr (0.5 rem/yr) from

dental products are below levels for induction of deterministic effects. The contribution to the

effective dose equivalent is negligible.

3.2.2 Collective Doses During Normal Life Cycle

The estimates of collective dose during the normal life cycle of a product or material associated

with the current exemptions for source material given in Table 2 range from 0.001 person-Sv

(0.1 person-rem) to 700 person-Sv (70,000 person-rem) for 1 year’s distribution. Collective

doses were not estimated in cases where the product is no longer in production, was never

produced or produced in limited quantity, or is not currently in wide-scale use. 

The following exemptions have collective dose estimates equal to or greater than

100 person-Sv (10,000 person-rem):

• 10 CFR 40.13(a): Chemical mixture, compound, solution, or alloy containing less than

0.05% by weight of source material,

• 10 CFR 40.13(c)(1)(i): Incandescent gas mantles containing thorium, 

• 10 CFR 40.13(c)(1)(iii): Welding rods containing thorium, and

• 10 CFR 40.13(c)(2)(ii): Glassware containing source material.

For the chemical mixture, compound, solution, or alloy containing less than 0.05% by weight

source material, the collective dose is a combination of estimated doses from the ophthalmic

glass, phosphate slag for building construction, and future on-site residents from disposal. For

the incandescent gas mantles, the users of portable camping lanterns contributes most to the

collective dose. The current trend toward use of gas mantles not containing thorium and other

lighting devices should significantly reduce this collective dose estimate. In the case of welding
xxvii

rods, the collective dose estimate of 300 person-Sv (30,000 person-rem) is predominantly to

professional welders over a 1 year time period. For glassware, the dose due to display of large

numbers of items (in homes and museums) contributed to the collective dose.

There are three (3) exemptions where the collective doses are greater than or equal to

10 person-Sv (1,000 person-rem) but less than 100 person-Sv (<10,000 person-rem):

• 10 CFR 40.13(c)(1)(vi): Rare earth metals and compounds, mixtures, and products

containing not more than 0.25% by weight source material,

• 10 CFR 40.13(c)(2)(i): Glazed ceramic tableware containing source material, and

• 10 CFR 40.13(c)(7): Finished optical lenses containing thorium.

For rare earth metals and compounds, mixtures, and products, the collective dose contributions

come from bastnasite and cerium concentrates (industrial workers), television faceplates, and

waste disposal (future on-site residents at landfills). For glazed ceramic tableware, the

estimated doses are due to display of large numbers of items (in homes and museums). In the

case of thorium in finished optical lenses, the estimated doses to users of 35-mm photographic

cameras contributes most of the collective dose. 

The collective doses for all the remaining 10 CFR Part 40 exemptions, for which a dose

assessment was made, are less than 10 person-Sv (<1000 person-rem).

3.2.3 Individual Doses Due to Accidents or Misuse

The estimates of individual dose due to accidents or misuse of a product or material associated

with the current exemptions for source material range from 4×105

 mSv (0.004 mrem) to

0.7 mSv (70 mrem). There appear to be no credible scenarios for accidents or misuse of

exempted products or materials containing source material that could result in doses in excess

of the current 10 CFR Part 20 radiation dose limit of 1 mSv/yr (100 mrem/yr) to individual

members of the public. The low doses due to accidents or misuse are a result of the low

specific activity of uranium and thorium, used under the exemptions.

In some of the exemptions for source material, irradiations of localized parts of the body,

including the cornea of the eye, skin of the hand, and a small area of the skin, by alpha or beta

particles due to misuse also were considered. For example, the estimated absorbed dose to

the cornea is 0.4 gray (Gy)/yr (40 rad/yr) due to irradiation by alpha particles during

unauthorized use of thoriated glass eyepieces in optical instruments. However, the estimated

dose is below the level for induction of deterministic effects. The contribution to the effective

dose equivalent is negligible.
xxviii

3.3 Assessments of Some Generally Licensed Products, Containing Byproduct Material,

That are Candidates for Exemption

3.3.1 Individual Doses During Normal Life Cycle

The estimates of the maximum dose to an individual during the normal life cycle of some

generally licensed products containing byproduct material range from 6×104

 mSv/yr

(0.06 mrem/yr) for static eliminators containing 210Po used in consumer products to 0.3 mSv/yr

(30 mrem/yr) for loose calibration and reference sources containing up to 10 times an exempt

quantity of byproduct material as listed in 10 CFR 30.71 Schedule B.

Except for static eliminators containing 210Po used in consumer products, all of the potential

candidates for exemption must meet safety criteria for normal use specified in 10 CFR 32.51, in

the form of dose limits to the whole body, body organs, and various other parts of the body. All

of the products considered in this assessment comply with the safety criteria in this section of

the regulations.

For many of the potential candidates for exemption containing byproduct material, irradiation of

the hands during normal use may be of concern. In all cases, however, the estimated doses

are well below levels for induction of deterministic effects, with the highest being 0.04 Sv/yr

(4 rem/yr) to the fingers from a 204Tl beta backscatter device used for measuring thicknesses of

various coatings in an industrial environment.

3.3.2 Collective Doses During Normal Life Cycle

The estimates of collective dose during the normal life cycle of a product associated with these

candidates for exemption are all less than 1 person-Sv (<100 person-rem) for 1 year’s

distribution. The highest is 0.6 person-Sv/yr (60 person-rem/yr) for X-ray fluorescence

analyzers containing 55Fe or 109Cd.

3.3.3 Individual Doses Due to Accidents or Misuse

The estimates of individual dose due to accidents or misuse of a product associated with these

potential candidates for exemption ranged from 0.005 mSv (0.05 mrem) to 2 mSv (200 mrem),

with the highest value associated with both a leaking 63Ni or 3

H source for a gas chromatograph

and with a cleanup after a transportation fire involving commercial static eliminators containing

210Po.

Except for static eliminators containing 210Po used in consumer products, all of the potential

candidates for exemption must currently meet safety criteria for accidents, as specified in

10 CFR 32.51, which are in the form of dose limits to the whole body, body organs, and various

other parts of the body. Except as described below, all of the products considered in this

assessment comply with these safety criteria.

In the cases of beta backscatter or transmission devices, X-ray fluorescence analyzers, and

calibration and reference sources, irradiation of localized parts of the body, including the hands

and a small area of the skin, due to misuse also was considered. In three cases, including

carrying of a discarded 85Kr source from a beta transmission device or a discarded 55Fe source
xxix

from an X-ray fluorescence analyzer in a pocket for a short time and misplacement of a loose

calibration or reference source in the folds of a desk chair for a longer time, the estimated

doses to a small area of the skin of 3 to 5 Sv (300 to 500 rem) exceed the specified safety

criterion of 2 Sv (200 rem) for accidents involving these products. In a scenario involving

handling of a loose calibration or reference source containing 204Tl, the estimated dose to the

hands of 1 Sv (100 rem) is high, but the specified safety criterion, which is the same as the

value for irradiation of a small area of the skin given above, would not be exceeded.
xxx

Table 1 Highest Individual Annual Doses from Normal Use a

Report

Section Exemption

Effective

Dose

Equivalent

(mrem)b

Byproduct Material

2.2

2.3

2.4

2.5

2.6

2.7

2.8

2.9

2.10

2.11

2.13

2.14

2.15

Exempt Concentrations

Timepieces, Hands & Dials

Automobile Lock Illuminators

Balances of Precision

Automobile Shift Quadrants

Marine Compasses & Navigational Instruments

Thermostat Dials & Pointers

Electron Tubes

Ionizing Radiation Measurement Instruments

Spark Gap Irradiators

Exempt Quantities

Self-Luminous Products

Gas & Aerosol Detectors

<1

9

0.2

0.001

0.3

0.03

0.08

5

20

10

2

0.3

2

Source Material

3.2

3.3

3.4

3.5

3.6

3.7

3.8

3.9

3.10

3.11

3.12

3.13

3.15

3.16

3.17

3.18

3.19

3.20

3.21

Chemical Mixture, Compound, Solution, or Alloy

Unrefined & Unprocessed Ore

Incandescent Gas Mantles

Vacuum Tubes

Welding Rods

Electric Lamps for Illuminating Purposes

Germicidal Lamps, Sunlamp, & Lamps for Outdoor or Industrial Lighting

Rare Earth Metals and Compounds, Mixtures and Products

Personnel Neutron Dosimeters

Glazed Ceramic Tableware

Piezoelectric Ceramic

Glassware

Photographic Film, Negatives & Prints

Finished Tungsten- or Magnesium-Thorium Alloy Products or Parts

Uranium in Counterweights

Uranium Shielding in Shipping Containers

Thorium in Finished Optical Lenses

Aircraft Engine Parts Containing Nickel-Thoria Alloy

Uranium in Fire Detection Units

4,000

300

200

0.2

800

<0.001

0.01

3,000

0.9

50

0.2

4

0.03

50

90

5

20

10

0.06

a

 Normal use encompasses the highest dose value from distribution and transport, routine use,

and disposal for the amount assumed to be distributed in 1 year.

b

 Since the summary table of radiation doses for each exemption in this report show dose in the

unit of mrem” (with a footnoted conversion factor to mSv), this table is also presented in that

format; 1 mrem = 0.01 mSv.
xxxi

Table 2 Collective Doses from Normal Use a

Report

Section Exemption

Collective

Effective Dose

Equivalent

(person-rem)b

Byproduct Material

2.2

2.3

2.4

2.5

2.6

2.7

2.8

2.9

2.10

2.11

2.13

2.14

2.15

Exempt Concentrations

Timepieces, Hands & Dials

Automobile Lock Illuminators c

Balances of Precision c

Automobile Shift Quadrants c

Marine Compasses & Navigational Instruments c

Thermostat Dials & Pointers c

Electron Tubes

Ionizing Radiation Measurement Instruments

Spark Gap Irradiators c

Exempt Quantities

Self-Luminous Products

Gas & Aerosol Detectors

90

4,000

—

—

—

—

—

1,000

10

—

60

20

200

Source Material

3.2

3.3

3.4

3.5

3.6

3.7

3.8

3.9

3.10

3.11

3.12

3.13

3.15

3.16

3.17

3.18

3.19

3.20

3.21

Chemical Mixture, Compound, Solution, or Alloy

Unrefined & Unprocessed Ore

Incandescent Gas Mantles

Vacuum Tubes

Welding Rods

Electric Lamps for Illuminating Purposes

Germicidal Lamps, Sunlamps, & Lamps for Outdoor or Industrial Lighting

Rare Earth Metals and Compounds, Mixtures and Products

Personnel Neutron Dosimeters c

Glazed Ceramic Tableware

Piezoelectric Ceramic

Glassware

Photographic Film, Negatives & Prints

Finished Tungsten- or Magnesium-Thorium Alloy Products or Parts

Uranium in Counterweights

Uranium Shielding in Shipping Containers

Thorium in Finished Optical Lenses

Aircraft Engine Parts Containing Nickel-Thoria Alloy

Uranium in Fire Detection Units c

50,000

10

70,000

300

30,000

0.1

2

1,000

—

4,000

 0.1 

10,000

30

100

300

200

10,000

5

 —

a

 Collective dose from normal use for each exemption is the sum of the collective doses from all

exposure scenarios under distribution and transport, routine use, and disposal.

b

 Refer to the text discussion of each section for the time period of the collective dose

calculations. Since the summary table of radiation doses for each applicable exemption in this

report show collective dose in the unit of person-rem” (with a footnoted conversion factor to

person-Sv), this table is also presented in that format; 1 person-rem = 0.01 person-Sv.

c

 Collective doses were not estimated for these exemptions because the product is no longer in

production, was never produced or produced in limited quantity, or is not currently in wide-scale

use.
xxxii
xxxiii

FOREWORD

The primary purpose of this report is to present an assessment of potential individual and

collective (population) radiation doses associated with the current exemptions for byproduct and

source material in Title 10, of the Code of Federal Regulations (CFR). The assessment for

each exemption considers potential impacts during the normal life cycle of a product or material

and from accidents or misuse. Also presented is an assessment of potential radiological

impacts associated with selected products containing byproduct material which currently may

be used under a general or a specific license and may be candidates for exemption from

licensing requirements. This report may be useful in confirming the acceptability of the current

exemptions and assessing the acceptability of future exemptions using new dose assessment

information.

This study was initiated in 1990. For some of the exemptions, it is known that present day use

may have changed from that identified and used in this study. An effort was made to up-date

references when possible. For some of the exemptions, there is no evidence that the analyzed

products were ever made, or if made at one time, there is no evidence that they are still

manufactured. Therefore, it was not feasible to re-establish present day use for all of the

exemptions.

The majority of this report was developed prior to implementation of the NRC’s metrication

policy for dual units which requires that the newer International System of Units (SI units) (i.e.,

becquerel, gray, and sievert) precede the older Special Units (i.e., curie, rad, and rem). All

activity, absorbed dose, and dose equivalent values presented in the text of this report are in

the dual units format. However, the arduous task of revising the tables to include dual units or

principally SI units was not done due to the large number of tables, complexity of many tables,

and time limitations to complete this report. Consequently, tables are presented in Special

Units and include a footnote providing the conversion factor(s) to SI units.

NUREG–1717 is not a substitute for NRC position papers or regulations, and compliance is not

required. The results, approaches, and methods described in this NUREG are provided for

information only. 

Thomas L. King, Director

Division of Risk Analysis and Applications

Office of Nuclear Regulatory Research
1–1

1 INTRODUCTION

1.1 Purpose of Present Study

In Title 10, of the Code of Federal Regulations (CFR), the U.S. Nuclear Regulatory Commission

(NRC) and its predecessor agency, the U.S. Atomic Energy Commission (AEC), established

regulations specifying products or materials containing source and byproduct material for which

the possession, use, and transfer are exempted from requirements for domestic licensing. 

Many of the current exemptions apply to consumer products containing radioactive material. 

However, other exemptions apply to any uses of radioactive material or, conversely, only to

highly specialized uses of radioactive material not involving consumer products (e.g., uses in

particular industries).

The primary purpose of this report is to present an assessment of potential radiological impacts

on the public associated with the present regulatory exemptions for source and byproduct

materials. As described in Section 1.2, these exemptions generally have been based on a

determination by the AEC or NRC that the possession, use, and transfer of the exempted

materials would not constitute an unreasonable risk to public health and safety. However, the

exemptions were established over many years, some as early as the 1940s when radiation

protection standards for the public were not yet included in AEC regulations (such standards

were first established in 10 CFR Part 20 in 1957) and methods for quantitative assessment of

dose to the public had not yet been developed. Therefore, approaches used by the AEC and

NRC in assessing radiological impacts on the public in support of establishing the exemptions

have varied widely. In addition, for exposures involving ingestion or inhalation of radionuclides,

assessments often were based on internal dosimetry models and databases that, although

representing the state of the art at the time, have since been superseded and are no longer

used by Federal agencies. In particular, most assessments were based on models for

estimating dose to the whole body or the so-called critical organ (usually the organ receiving the

highest dose) as presented in Publication 2 of the International Commission on Radiological

Protection (ICRP 2). Federal agencies now use internal dosimetry data in the form of effective

dose equivalents (EDEs) that are based on the recommendations in ICRP Publication 26 and

internal dosimetry models presented in ICRP Publication 30. Therefore, there was a need to

reevaluate the current exemptions for source and byproduct materials to determine the potential

radiological impacts on the public.

Regulations in 10 CFR Parts 30 and 40 also have provisions for general licenses that permit the

possession and use of specified quantities of certain radionuclides without the need for specific

application or issuance of licensing documents to the persons using the radioactive materials. 

Generally licensed radioactive materials usually are incorporated into products, devices, or

equipment manufactured under a specific license issued by the NRC or an Agreement State. 

Some generally licensed items contain only small quantities of byproduct material, and these

items are potential candidates for exemption from licensing requirements. As part of this study,

assessments of the potential radiological impacts on the public associated with five generally

licensed items containing byproduct material were performed. The results of these

assessments could be used to support establishing exemptions for these items.
1–2

1.2 Basis for Existing Exemptions

As indicated in the previous section, the existing exemptions for source and byproduct materials

generally were based on a determination by the AEC or NRC that the possession, use, and

transfer of the exempted products or materials would not constitute an unreasonable risk to

public health and safety. On March 16, 1965 (30 FR 3462), the AEC issued a policy statement

that discussed the criteria that were applied in exempting the use of consumer products

containing source or byproduct material. Although each exemption was considered individually,

these criteria were used in establishing many of the existing exemptions, with the exception of

the few exemptions that are not specifically for consumer products.

The criteria developed by the AEC include some general considerations that are used in

establishing exemptions and the principal factors that are evaluated for each consumer product. 

The general considerations in the 1965 policy statement are described below.

1. At the time of issuance of an exemption, it should appear unlikely that the total radiation

exposure to the general public from use of consumer products containing radioactive

material would exceed small fractions of recommended limits for exposure to radiation

from all sources. Information on the total quantities of radioactive materials being used

in such products and the number of items being distributed will be obtained through

keeping of records and reporting requirements applicable to the manufacture and

distribution of such products. If, at any time, radioactive materials are used in sufficient

quantities in products reaching the public that population doses could become a

significant fraction of the permissible dose to the gonads, then the policy on use of

radioactive materials in products will be reconsidered.

2. Exemption of a product intended for use by the general public will depend on both the

associated radiation exposures and the apparent usefulness of the product. In general,

risks from radiation exposure will be considered acceptable if (a) handling, use, and

disposal of the product are unlikely to result in doses to individuals in the population

exceeding a few percent of dose limits for individual members of the public

recommended by such groups as the ICRP, the National Council on Radiation

Protection and Measurements, and the Federal Radiation Council (FRC) and (b) the

probability of individual doses approaching any of the specified limits is negligibly small. 

If these conditions are not met, a more careful weighing of all factors will be required.

3. As a general rule, exempted products will be considered useful to some degree. When

tangible benefits to the public are questionable and approval of a product may result in

widespread use of radioactive material, the degree of usefulness and benefit to the

public may be a deciding factor in granting an exemption. In particular, the use of

radioactive material in toys, novelties, and adornments may be of marginal benefit.

4. Exemptions for off-the-shelf” items that are subject to mishandling, especially by

children, will be granted only if such items are found to combine an unusual degree of

utility and safety.

5. Certain longstanding and widespread uses of source material are exempted primarily

because they antedate the atomic energy program. These include (a) the use of
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uranium to color glass and glazes for certain decorative purposes, (b) the use of thorium

in various alloys and products to impart desirable physical characteristics, and (c) the

use of uranium and thorium in photographic film and prints.

6. The use of tritium as a luminous material on watch and clock dials and hands has been

exempted to provide a substitute for the longstanding use of radium for this purpose.

7. In exempting uses of source and byproduct materials in consumer products, limits on

quantities or concentrations of radioactive materials and, if appropriate, levels of

radiation emitted may be established. In some cases, requirements on quality control

and testing also are specified if they are considered important to health and safety.

The 1965 policy then states that the principal factors to be evaluated for each consumer

product in deciding whether to grant an exemption include the following considerations.

1. In evaluating proposals for exempting the use of radioactive materials in consumer

products, the principal considerations are (a) the potential external and internal

exposure of individuals in the population from handling, use, and disposal of individual

products, (b) the potential total dose to individuals in the population who may be

exposed to a number of products, (c) the potential long-term exposure of the general

population from uncontrolled disposal and dispersal of radioactive materials in the

environment, and (d) the benefit that will accrue to or be denied the public because of

the utility of the product by approval or disapproval of an exemption for a specific

product.

2. Detailed evaluations of potential exposures to radioactive materials in a consumer

product would consider (a) the external radiation levels from the product, (b) the

proximity of the product to human tissue during use, (c) the area of tissue exposed,

particularly for exposure of the skin, (d) the radiotoxicity of the radionuclides, with less

toxic materials considered more favorably than materials with a high radiotoxicity, (e) the

quantity of radioactive material per individual product, with relatively small quantities

considered more favorably, (f) the form of the material, with materials with low solubility

in body fluids considered more favorably than those with high solubility, (g) containment

of the material provided by the product, particularly under very severe environmental

conditions, and (h) the degree of access to the product during normal handling and use,

with inaccessible products considered more favorably.

Thus, the current policy for exempting consumer products containing source or byproduct

material calls for considerations of (1) the benefits from use of the products, (2) radiation doses

to individuals and populations from normal handling, use, and disposal of the products, and

(3) risks to the public from accidents and misuse of the products.

The existing exemptions for self-luminous products (10 CFR 30.19) and gas and aerosol

detectors (10 CFR 30.20) are considered class” exemptions. For these two exemptions, new

products within a class can be approved through licensing, rather than by establishing a new,

separate exemption through rulemaking. The conditions for the class exemptions include dose

criteria that are applicable to scenarios for accidents and misuse, and an applicant for a license

to distribute a product for use under either of these exemptions must demonstrate that the

proposed product meets the criteria. Because these criteria are more specific with respect to
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acceptable risks from accidents and misuse involving the product than is the general policy on

consumer products described above, the requirements for compliance with the dose criteria can

be considered an extension of the policy in this area.

The dose criteria applicable to scenarios for accidents and misuse for the class exemptions are

specified in 10 CFR 32.23 and 32.24 and in 10 CFR 32.27 and 32.28, and are stated below.

In use and disposal of a single exempt unit and in handling and storage of the quantities

of exempt units that are likely to accumulate in one location during marketing,

distribution, installation, and servicing of the product, the probability is low that the

containment, shielding, or other safety features of the product would fail under such

circumstances that an individual would receive an external dose or dose commitment in

excess of 5 millisieverts (mSv) (0.5 rem) to the whole body, head and trunk, active

blood-forming organs, gonads, or lens of the eye; 75 mSv (7.5 rem) to the hands and

forearms, feet and ankles, or localized areas of the skin averaged over areas no larger

than 1 square centimeter; and 15 mSv (1.5 rem) to any other organs; and the probability

is negligible that an individual would receive an external dose or dose commitment in

excess of 150 mSv (15 rem) to the whole body, head and trunk, active blood-forming

organs, gonads, or lens of the eye; 2 Sv (200 rem) to the hands and forearms, feet and

ankles, or localized areas of the skin averaged over areas no larger than 1 square

centimeter; and 0.5 Sv (50 rem) to any other organs.

The term dose commitment” refers to the 50-year committed dose from internal exposure.

A footnote to the dose criteria for the class exemptions states that the probabilities are

expressed in general terms, rather than quantitatively, to emphasize the approximate nature of

the estimates to be made. However, the following guidance is provided for quantifying low and

negligible probabilities of failure of safety features for purposes of demonstrating compliance

with the dose criteria: a probability of failure is low” if there is not more than one failure per

year for each 10,000 exempt units distributed, and a probability of failure is negligible” if there

is not more than one failure per year for each 1 million exempt units distributed.

1.3 Requirements for Radiation Protection of the Public

Radiation doses to individuals and populations are a particular concern in approving

exemptions for products and materials containing source or byproduct material. However, the

policy for consumer products described in Section 1.2 does not include quantitative dose

criteria for limiting exposure to the public for normal (i.e., routine, expected) exposure

situations. Rather, the policy states only that doses to individuals and populations should be a

small fraction of applicable limits in radiation protection standards for the public. This section

discusses requirements for radiation protection of the public and the implications of these

requirements with regard to the current policy for approving exemptions for source or byproduct

material.

When most of the current exemptions were established, the recommended dose limit for

individual members of the public from all manmade sources of exposure was 5 mSv/yr

(500 millirem (mrem)/yr) to the whole body (FRC, 25 FR 4402). In addition, the FRC had
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issued guidance that the dose to the gonads of average individuals in the population should not

exceed 50 mSv (5 rem) in 30 years, or an average of 1.7 mSv/yr (170 mrem/yr)

(FRC, 25 FR 4402). Thus, a small fraction of the applicable limits on radiation exposure to the

public included doses as high as a few tenths of a mSv/yr (few tens of an mrem/yr). Finally,

radiation protection standards for the public included a provision that doses should be reduced

as low as reasonably achievable (ALARA). This involves taking into account economic factors

(i.e., cost-benefit for dose reduction) and other societal concerns (FRC, 25 FR 4402;

AEC, 25 FR 10914).

However, since most of the current exemptions were established, two particularly important

developments in radiation protection of the public occurred. First, the dose limit for individual

members of the public from all manmade sources of exposure was lowered to 1 mSv/yr

(100 mrem/yr) (NRC, 56 FR 23360). Second, use of the ALARA principle is now a requirement

(NRC, 56 FR 23360), and this requirement applies to reduction of doses below any authorized

limits for specific practices or sources.

Thus, within the current framework for radiation protection of the public, it may no longer be the

case that doses to individual members of the public as high as a few tenths of a mSv (few tens

of a mrem) from the exempt use of products and materials containing source or byproduct

material would be considered acceptable. Based on this consideration, the NRC decided to

reevaluate the radiological impacts associated with current exemptions to determine whether

potential doses to individuals are consistent with current authorized limits and doses for

regulated practices or sources.

1.4 Description of Present Study

As indicated in Section 1.1, the primary purpose of this study is to provide an assessment of

potential radiological impacts on the public associated with all of the present exemptions for

source and byproduct materials. In addition, this study provides an assessment of potential

radiological impacts associated with certain generally licensed items containing byproduct

material that are potential candidates for exemption.

The impetus for this study is the need for a systematic assessment of all exemptions and

potential candidates for exemption using a reasonably consistent dose assessment

methodology based on current modeling approaches and updated internal dosimetry data. 

Such a systematic assessment would allow comparisons of radiological impacts associated with

different exemptions or potential candidates for exemption, as well as evaluations of the total

radiological impacts associated with all exemptions combined. As noted in Section 1.1,

previous assessments used a wide variety of approaches to evaluate radiological impacts,

ranging from mostly qualitative considerations to quantitative modeling studies, as well as

internal dosimetry data that are now outdated. Therefore, previous assessments cannot readily

be used to compare impacts associated with different exemptions or to evaluate impacts

associated with all exemptions combined.

1.4.1 Dose Assessments for Normal Life Cycle

This section provides a general introduction to the approaches used in this study in assessing

radiological impacts on the public from normal life cycle of products or materials containing
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source or byproduct material and the types of results presented in this report. Both individual

and collective dose assessments were performed. However, if an exempt product is not

currently being produced and is not in wide-scale use, collective doses were not estimated

since there does not exist a basis for such an assessment.

The final dose estimates (individual and collective) have been rounded to one (1) significant

figure, based on the overall accuracy of the modeling. Also, if the calculated individual dose

was less than 1×105

 mSv (<0.001 mrem), the dose was presented as a less than value” (i.e.,

<1×105

 mSv (<0.001 mrem)). However, for purposes of estimating collective dose, the

calculated value, not a less than value, was used. Inconsistencies between individual doses

and a resulting extrapolated collective dose is a result of the rounding and the use of less than

values.

1.4.1.1 Stages of Normal Use

For all current exemptions and potential candidates for exemption, doses to members of the

public are assessed for all stages of normal (expected), unregulated use throughout the life

cycle of the product or material. The particular stages of the normal life cycle for which

radiological impacts were evaluated include the following:

The distribution and transport, e.g., from a licensed manufacturer of an exempted

product or material to members of the public who are users of the product or material.

The intended or expected routine use of the product or material.

The disposal of the product or material.

For each of these stages, individual and collective (population) doses were estimated based on

an assumed amount of radioactive material per item and an assumed annual distribution of

radioactive material in all items. These assumptions may be based on such information as

specifications in the exemption and data from materials licensee reports on the total number of

items distributed annually and the amount of radioactive material per item.

In all assessments, individual and collective doses for the different stages of the normal life

cycle estimated in this study are in the form of EDEs (ICRP 26; NRC, 56 FR 23360). In a few

cases, including the assessments for the two class exemptions for self-luminous products and

gas and aerosol detectors, individual doses also are given in the form of dose equivalents to the

whole body or particular organs or tissues, when the regulations establishing these exemptions

specify limits on dose equivalent during normal use in this form. 

1.4.1.2 Individual Doses Estimated in Assessments

The individual doses during the normal life cycle presented in this study are in the form of

annual (yearly) doses for those groups of individuals expected to receive the highest doses for

each of the three life cycle stages listed above. The use of annual doses for individuals

conforms to the conventional approach in radiation protection of the public (NRC, 56 FR

23360). The estimates of annual individual dose for routine use and disposal take into account

radioactive decay whenever it is significant, but the estimates of annual individual dose for

distribution and transport generally ignore decay, because distribution and transport is assumed
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to occur only over a relatively short period of time compared with the half-lives of the

radionuclides of concern.

In estimating individual doses for routine use accidents and misuse, the primary emphasis is on

doses associated with the maximum allowable amounts of radionuclides in products or

materials, as specified in the applicable regulations. Such estimates provide a measure of the

highest allowable impacts on individuals. In many cases, however, the actual amounts of

radionuclides present in the products or materials are known to be considerably less than the

maximum allowable amounts, and the difference between the individual doses for the maximum

allowable and actual amounts is noted. In addition, some regulations do not specify the

maximum allowable amounts of radionuclides, and the assessment of individual doses in these

cases is based on information or assumptions about the actual amounts present. For

distribution, transport, and disposal actual amounts of radionuclides present in the products or

materials have been used, where available.

1.4.1.3 Collective Doses Estimated in Assessments

The collective doses presented in this study are in the form of total doses over time for an

assumed annual distribution of radioactive material. The approach to calculating the total

collective dose from 1 year’s distribution of radioactive material for the different life cycle stages

of normal use is described as in the following paragraphs.

For distribution and transport, the collective dose is assumed to be experienced only during the

same year as the initial distribution. As in estimating individual dose, radioactive decay

generally is ignored in estimating collective dose during distribution and transport.

For routine use, the collective dose from 1 year’s distribution of radioactive material is

calculated over the useful lifetime of the product or material. For example, if a product has an

expected lifetime of 10 years, the collective dose is the total dose over 10 years from the

assumed amount of radioactive material distributed in 1 year. At steady-state, this is the same

as the collective dose in 1 year from the total quantity of radioactive material in use, taking into

account the annual distribution and radioactive decay over the lifetime of the product.

For disposal, the collective dose from 1 year’s distribution of radioactive material is the dose

during the same year that disposals occur for some population groups, but is the total dose

over time after disposal for other population groups. Specifically, the collective dose to various

population groups during waste operations at disposal facilities is the dose received only during

the year that disposals occur. However, the collective dose to various population groups

following closure of disposal facilities is the total dose over 1000 years from 1 year’s distribution

of radioactive material. In all cases, the collective dose from disposal takes into account

radioactive decay between the time of distribution and the time exposures are assumed to

occur.

In estimating collective doses for distribution and transport, routine use, and disposal, the

primary emphasis is on doses associated with the actual amounts of radionuclides distributed,

particularly when these amounts can be accurately estimated. When the actual amount of a

radionuclide distributed in a product is substantially less than the maximum allowable amount

for that product, estimates of collective dose based on the actual amount will provide a

reasonable measure of population impacts. In some cases, however, the maximum allowable
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amounts of radionuclides were used in estimating collective dose, particularly when they did not

differ greatly from the actual amounts.

1.4.1.4 Development of Exposure Scenarios

The estimates of individual and collective dose for distribution and transport, routine use, and

disposal obtained in this study generally are based on assumptions about exposure scenarios. 

For example, in estimating external dose, assumptions generally are made about the distance

between a source and exposed individuals, the amount of shielding between the source and

receptor locations, and the amount of time spent near the source. Similarly, in estimating

inhalation dose, assumptions are made about the amount of radioactive material released from

a particular product or material into the air, the size and ventilation rate of the air space into

which the material is released, the breathing rate of individuals, and the exposure time. For

estimates of ingestion dose, assumptions were also made about the amount of radioactive

material released and the fraction of the released material that would be ingested.

Because the purpose of this study is to provide a systematic assessment of potential

radiological impacts on the public associated with a wide variety of products or materials and

practices, standard assumptions were used in defining and evaluating exposure scenarios for

all assessments to the extent practicable and reasonable. For example, generic methodologies

were developed to provide standard, default estimates of individual and collective doses for

distribution, transport and disposal, and the results of these methodologies were used in all

assessments for particular products or materials when they are appropriate. In addition, a

standard set of dose coefficients for inhalation and ingestion of radionuclides is used in all

assessments; a minimal set of standard, default assumptions about room sizes and ventilation

rates and breathing rates is used in assessments of inhalation dose; a minimal set of computer

codes and databases was used in estimating doses from external exposure to photons or

electrons; and reasonably uniform assumptions about exposure times and source-to-receptor

distances were used in many cases in estimating external dose from photon exposure.

However, in spite of the desire for a uniform, standardized approach to dose assessments for

the wide variety of products or materials and practices of concern, the approach to estimating

individual and collective doses was performed on a case-by-case basis throughout this study,

particularly in developing exposure scenarios for routine use, because reasonable scenarios

may vary considerably depending on the particular product or material. Another important

factor in estimating doses in some cases is the availability of relevant measurements, e.g., on

external dose rates near sources or airborne concentrations of radionuclides during use of

products or materials. In such cases, the measurements normally are used in estimating dose,

rather than results based on standard models or assumptions. Even in applying the generic

methodologies for distribution and transport and disposal, judgments are required in defining

exposure scenarios for particular products or materials, and the appropriate scenarios can differ

from one case to another.

The development and application of exposure scenarios clearly involves considerable

uncertainty. However, explicit consideration of uncertainties in the estimated individual and

collective doses was beyond the scope of this study. Rather, the intent was to develop credible

exposure scenarios for distribution and transport, routine use, and disposal for each

assessment, then to implement the scenarios using parameter values that are reasonable

considering the range of possible exposure scenarios and associated parameter values. 
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However, use of this approach will produce estimates of individual and collective dose that are

more likely than not to overestimate actual impacts, but the intent was not to calculate doses

that are so conservative that they would overestimate any impacts that might occur. 

1.4.2 Dose Assessment for Accidents and Misuse

All assessments performed in this study also consider doses from accidents and misuse of

products or materials containing source or byproduct material. However, in contrast to the dose

assessments for the life cycle only individual doses were calculated for accidents and misuse.

As in the dose assessments for life cycle the estimates of individual dose from accidents and

misuse obtained in this study generally are based on assumptions about exposure scenarios. 

Scenarios for accidents and misuse generally may involve external or internal exposure, and

many of the considerations about scenario development and application discussed in

Section 1.4.1.4 for normal use also apply here.

In estimating doses from accidents and misuse, the intent was to develop exposure scenarios

that, although unlikely to occur, are nonetheless plausible for the particular product or material

of concern, rather than to develop extreme but highly unlikely scenarios. Thus, the intent

usually was to provide reasonable upper bounds on doses. In some cases, however, an

extreme but highly unlikely scenario (e.g., ingestion of an entire exempt item) was used in order

to clearly bound any possible doses from accidents and misuse, particularly when the amount

of radioactive material that could be involved was relatively small and the results of the

assessment could be used to demonstrate that doses resulting from any scenario would not be

high.

In many cases, a generic accident methodology developed in this study was used in estimating

doses from accidents. The generic methodology provides standard, default dose estimates for

fires involving radioactive material, spills of radioactive materials in liquid or powder form, and

crushing of glass tubes containing radioactive gases. The results of this methodology were

used whenever any of these accident scenarios were considered appropriate for a particular

product or material. However, a variety of other scenarios involving inadvertent external,

inhalation, or ingestion exposure were considered in many of the assessments, and these

scenarios generally were developed and implemented on a case-by-case basis.

In all assessments, individual doses from accidents and misuse were calculated in the form of

EDEs, to be consistent with the assessments for normal use. For the two class exemptions for

self-luminous products and gas and aerosol detectors, doses from accidents and misuse were

also calculated in the form of dose equivalents to the whole body or particular organs or tissues

because, as discussed in Section 1.2, the safety criteria for accidents and misuse that apply to

these exemptions are expressed in this form.

Except for the two class exemptions for self-luminous products and gas and aerosol detectors,

individual doses from accidents and misuse are estimated without consideration of the

probability of occurrence of the assumed exposure scenarios. Rather, as discussed above, the

intent usually was to develop plausible scenarios for accidents and misuse that would represent

a variety of exposure situations that could occur. For the two class exemptions, however, the

safety criteria for accidents and misuse discussed in Section 1.2 require consideration of the
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probabilities of occurrence of exposure scenarios, because the allowable doses are higher for

scenarios with a negligible” probability than for scenarios with a low” probability. Even though

the regulations establishing these exemptions provide quantitative guidance on the meaning of

negligible” and low” probabilities, the assessments of scenario probabilities in this study are

more a matter of subjective, qualitative judgment than a rigorous quantitative analysis. 

1.5 Organization and Content of Report

The report consists of three major sections. Section 2 presents the assessments for exempted

products or materials containing byproduct material, Section 3 presents the assessments for

exempted products or materials containing source material, and Section 4 presents the

assessments for certain generally licensed items containing byproduct material that are

potential candidates for exemption. 

Each assessment for a particular product or material is presented in a separate section in one

of the three major sections of the report described above. A complete listing of the different

assessments performed in this study and presented in Sections 2 to 4 is given in the Table of

Contents.

The presentation of each assessment is generally organized as follows: The first part is an

introduction that describes the existing regulations, specifying the conditions that apply to the

particular exempted or generally licensed product or material. The introductory part also

discusses the bases for the existing exemption or general license, as presented by the AEC or

NRC in proposed and final rulemakings, and it may include a brief introduction to other studies

of radiological impacts on the public.

The second part of each assessment presents a description of the products or materials of

concern. The information presented depends on the particular products or materials but

generally includes (1) the physical and chemical form, and the size and construction of the

products or materials; (2) the intended or known uses of the products or materials and the

benefits provided by the incorporation of source or byproduct material; (3) the amounts of

source or byproduct material normally contained in the products or materials; and (4) the annual

production or distribution of all products or materials of concern.

The third part of each assessment presents a summary of previous analyses and assessments

of radiological impacts on the public associated with the exempted or generally licensed

products or materials for normal use, accidents, and misuse. These summaries also present

relevant information on the assumptions used in the previous studies and the individual and

collective doses that were obtained. 

The fourth part of each assessment presents the dose analysis for the normal life cycle,

accidents, and misuse. This part of the assessment documents all assumptions, models, and

methods used in calculating individual or collective doses, and it presents the resulting

estimates of dose.

The final part of each assessment presents a summary of the results obtained in the present

study for the particular products or materials. The estimated individual and collective doses for
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distribution and transport, routine use, and disposal and the estimated individual doses for

accidents and misuse are presented in a summary table.

Following Sections 2 to 4 of the main report, are three appendixes that present the generic

methodologies used in this study. A fourth appendix addresses a correction factor for the

CONDOS code (Computer Codes, O’Donnell, 1975) where the dose from bremsstrahlung

radiation for low energy electrons is overestimated and provides generic modeling for sources

in close proximity to the body. As noted in Sections 1.4.1.4 and 1.4.2, results obtained from the

four generic methodologies were used in most of the assessments for particular products or

materials. The appendixes provide a complete documentation of the methodologies, including

information on the assumed exposure scenarios, the models and databases used in estimating

doses for each scenario, tabulations of results that can be used in estimating dose for particular

radioactive materials, and a discussion of judgments or assumptions that must be used in

applying the results of the generic methodology to a particular assessment.
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2 EXEMPTIONS FOR BYPRODUCT MATERIAL

2.1 Properties of Byproduct Material

2.1.1 Introduction

This section provides an introduction to the following sections of Section 2, which present the

results of assessments of radiological impacts on members of the public from products or

materials containing exempted amounts of byproduct materials. The information presented in

this section includes the definition of byproduct material (Section 2.1.2), radioactive decay data

for selected byproduct materials (Section 2.1.3), and dosimetry data used in estimating dose

from external and internal exposure for the selected byproduct materials (Section 2.1.4). The

radioactive decay and dosimetry data presented in this section also are used in Section 4 in the

assessments of radiological impacts for certain generally licensed items containing byproduct

materials that are potential candidates for exemption.

2.1.2 Definition of Byproduct Material

As defined in 10 CFR 30.4 and used in this report, the term byproduct material” means:

....any radioactive material (except special nuclear material) yielded in or made

radioactive by exposure to the radiation incident to the process of producing or utilizing

special nuclear material.”

The term special nuclear material” in this definition is defined in 10 CFR 70.4 as:

....(1) plutonium, uranium 233, uranium enriched in the isotope 233 or in the isotope

235, and any other material which the Commission, pursuant to the provisions of section

51 of the act, determines to be special nuclear material, but does not include source

material; or (2) any material artificially enriched by any of the foregoing but does not

include source material.”

In the latter definition, the Commission” refers to the Nuclear Regulatory Commission (NRC),

the act” refers to the Atomic Energy Act of 1954, as amended, and the term source material”

is defined in Section 3.1.2 of this report.

Based on the definition in 10 CFR 30.4 given above, byproduct material includes any

radioactive material associated with operations of nuclear reactors, except for the source

material from which nuclear fuel is made and the special nuclear material which constitutes the

fuel in a reactor. Section 11(e)(2) of the Atomic Energy Act and 10 CFR 40.4 also defines

byproduct material to include the tailings or wastes produced by the extraction or concentration

of uranium or thorium from any ore processed primarily for its source material content. 

However, this report is not concerned with any such byproduct materials.

As indicated in 10 CFR 30.70, Schedule A, which lists exempt concentrations of byproduct

material (see Section 2.2), and 10 CFR 30.71, Schedule B, which lists exempt quantities of

byproduct material (see Section 2.13), a large number of byproduct materials are potentially of
2–2

concern in exempted products or materials. However, relatively few of these products have

been distributed to any significant extent in exempted products or materials, and only those

byproduct materials are considered in this section. Specifically, radioactive decay and

dosimetry data are presented only if the radionuclide is included in specific exemptions for

products or materials containing byproduct material or was distributed to a significant extent as

exempt concentrations or exempt quantities of byproduct material during the period of 1970 to

1991. An NRC license is required to distribute products containing exempt quantities of

byproduct material. Since all such products have been distributed under an NRC license, NRC

records contain all information pertaining to the total activity distributed in products. This is not

true, however, for concentrations of byproduct material that are authorized for exempt

distribution. In the latter case, exempt concentrations can be distributed by Agreement State

licensees. The activities of byproduct material distributed as exempt concentrations or exempt

quantities during this time were estimated from a review of distribution reports submitted to the

NRC by licensees under 10 CFR 32.12 and 32.20. To estimate the total activity it was assumed

that twice the quantity was distributed through Agreement State licensees as NRC licensees.

2.1.3 Decay Data for Selected Byproduct Materials

Radioactive decay data for the byproduct materials that have been used most frequently in

exempted products or materials are given in Table 2.1.1. These data include:

The half-life of the radionuclide.

The specific activity of the radionuclide, defined as the activity per unit mass.

Any short-lived radioactive decay products, their half-lives, and the branching fraction in

the decay of the parent radionuclide.

An identification of the principal decay modes for each radionuclide (i.e., beta, beta and

gamma, electron capture, positron, isomeric transition, or alpha).

Whenever a radioactive decay product is shorter lived than its parent radionuclide, the activity

of the decay product generally is assumed to be in equilibrium with the activity of the parent in

assessing radiological impacts on the public from exempted products or materials. For 131I,

however, the 131mXe decay product is longer lived than the parent radionuclide, and the activity

of the decay product never achieves equilibrium with the activity of the parent.

2.1.4 Dosimetry Data for Selected Byproduct Materials

Data that can be used to estimate external and internal dose from exposure to the selected

byproduct materials listed in Table 2.1.1 are given in Table 2.1.2. These data include:

The specific gamma-ray dose constant, which is defined as the dose equivalent rate per

unit activity at a distance of 1 meter from an unshielded point source in air and which is

a reasonable approximation to the effective dose equivalent (EDE) rate per unit activity

for radionuclides that emit high-energy photons.
2–3

The external dose coefficient for submersion in an atmospheric cloud, which is defined

as the external EDE rate per unit concentration in a uniformly contaminated,

semi-infinite volume of air. The skin dose component has been included using a 0.01

weighting factor.

The internal dose coefficient for ingestion, which is defined as the 50-year committed

EDE per unit activity intake by ingestion.

The internal dose coefficient for inhalation, which is defined as the 50-year committed

EDE per unit activity intake by inhalation.

The specific gamma-ray dose constant and external dose coefficient for air submersion are

listed in Table 2.1.2 only if a radionuclide emits photons of sufficient energy and intensity that

external exposure possibly could be of concern in assessing dose. Thus, these data are not

listed for radionuclides that are not photon emitters or that emit only very low-energy photons.

It also should be emphasized that the specific gamma-ray dose constant provides a

conservative estimate of the EDE from external exposure to a point source for radionuclides

with emitted photon energies substantially below about 100 keV. An important example is

241Am, which emits mainly 60-keV photons (Kocher, 1981). In any such cases, the specific

gamma-ray dose constant in Table 2.1.2 may be inappropriate for use in estimating external

dose. Radiation exposure codes such as MicroShield (Computer Codes, Grove Engineering,

1996) and CONDOS II (Computer Codes, O’Donnell et al., 1981) provide a better method for

estimating external dose.

For radionuclides that decay to shorter lived radioactive decay products, the dosimetry data for

the decay products are included only if the decay products contribute significantly to the dose

from exposure to the parent radionuclide and its decay products. The dosimetry data for each

short-lived decay product take into account the branching fraction in the decay of the parent

radionuclide given in Table 2.1.1. For 131I, the longer lived decay product 131mXe does not

contribute significantly to the dose from external or internal exposure to the parent radionuclide,

because the decay product is a noble gas and is produced only with a small branching fraction.

For some radionuclides, dose coefficients for ingestion are listed for more than one value of the

gastrointestinal-tract absorption fraction or dose coefficients for inhalation are listed for more

than one lung clearance class. Absent specific information on the chemical form of a

radionuclide in a particular product or material, the largest dose coefficient for ingestion or

inhalation is generally used in assessing dose. If the chemical form of a radionuclide in a

particular product or material is known, the appropriate dose coefficients for ingestion and

inhalation can be selected based on the assignments given in Table 2.1.3.

The dosimetry data in Table 2.1.2 are used, when appropriate, in the dose assessments for

byproduct material in the remainder of Section 2 and in Section 4. The dose coefficients for

ingestion and inhalation are used in all assessments of internal exposure to byproduct

materials. The external dose coefficient for air submersion also is used whenever this exposure

pathway is considered. The external dose coefficients provide conservative estimates of dose

for submersion in a finite atmospheric cloud.
2–4

However, the specific gamma-ray dose constant, which provides an indication of the potential

importance of external exposure, is used to estimate external dose only when exposure to an

unshielded point source is an appropriate assumption, i.e., when the dimensions of the source

are small compared with the distance between the source and receptor locations, and the

radionuclide of concern emits photons with energies above about 100 keV. For finite sources

that cannot be represented as a point and for radionuclides that emit only lower energy photons

(e.g., 241Am) , external dose rates normally are calculated using the CONDOS II (Computer

Codes, O’Donnell et al., 1981) or MicroShield (Computer Codes, Grove Engineering, 1996)

Computer Codes for the appropriate source geometry and amount of shielding between the

source and receptor locations.
2–5

Table 2.1.1 Decay Data for Selected Byproduct Materials

Radionuclide

Decay

Product Half-Lifea

Specific Activity

(curie (Ci)/g)b

Branching

Fractionc

Principal

Decay Modesd

3

H 12.28 yr 9.70×103 Beta

14C 5,730 yr 4.46 Beta

24Na 15.00 h 8.70×106 Beta/gamma

32P 14.29 d 2.87×105 Beta

35S 87.44 d 4.27×104 Beta

36Cl 3.01×105

 yr 3.30×102 Beta

45Ca 162.7 d 1.78×104 Beta

46Sc 83.83 d 3.38×104 Beta/gamma

51Cr 27.704 d 9.24×104 EC/gamma

54Mn 312.7 d 7.73×103 EC/gamma

55Fe 2.7 yr 2.41×103 EC

57Co 270.9 d 8.46×103 EC/gamma

58Co 70.80 d 3.29×104 Pos/gamma

59Fe 44.63 d 4.96×104 Beta/gamma

60Co 5.271 yr 1.13×103 Beta/gamma

63Ni 100.1 yr 5.68×101 Beta

65Zn 244.4 d 8.23×103 Pos/gamma

75Se 119.78 d 1.45×104 EC/gamma

82Br 35.30 h 1.08×106 Beta/gamma

85Kr 10.72 yr 3.93×102 Beta/gamma

90Sr 28.6 yr 1.39×102 Beta

90Y 64.1 h 1.0 Beta

99Tc 2.13×105

 yr 1.70×102 Beta

106Ru 368.2 d 3.35×103 Beta

106Rh 29.92 s 1.0 Beta/gamma

109Cd 464 d 2.59×103 EC

109mAg 39.6 s 1.0 IT

See end of table for footnotes.
2–6

Table 2.1.1 Decay Data for Selected Byproduct Materials (continued)

Radionuclide

Decay

Product Half-Lifea

Specific Activity

(curie (Ci)/g)b

Branching

Fractionc

Principal

Decay Modesd

110mAg 249.85 d 4.76×103 Beta/gamma

110Ag 24.57 s 0.0133 Beta/gamma

113Sn 115.1 d 1.01×104 EC/gamma

113In 1.658 h 1.0 IT/gamma

125I 60.14 d 1.74×104 EC

129I 1.57×107

 yr 1.77×104 Beta

131I 8.04 d 1.24×105 Beta/gamma

131mXe 11.84 de 8.41×104 0.01086 IT/gamma

133Ba 10.5 yr 2.56×102 EC/gamma

134Cs 2.062 yr 1.29×103 Beta/gamma

137Cs 30.17 yr 8.65×101 Beta

137mBa 2.552 min 0.946 IT/gamma

140La 40.22 h 5.58×105 Beta/gamma

147Pm 2.6234 yr 9.27×102 Beta

152Eu 13.6 yr 1.73×102 ECf

/gamma

182Ta 114.74 d 6.24×103 Beta/gamma

195Au 183 d 3.66×103 EC/gamma

198Au 2.696 d 2.45×105 Beta/gamma

203Hg 46.60 d 1.38×104 Beta/gamma

204Tl 3.779 yr 4.65×102 Beta/EC

206Tl 4.2 ming 1.0 Beta

210mBi 3.0×106

 yrg 5.69×104 Alpha/gamma

210Po 138.378 d 4.49×103 Alpha

241Am 432.2 yr 3.43 Alpha

a

 Values from Kocher (1981), except as noted.

b

 1 Ci/g = 0.037 terabecquerel (TBq)/g.

c

 Number of atoms of decay product per decay of parent radionuclide.

d

 EC = electron capture decay, Pos = positron decay, and IT = isomeric transition.

e

 Decay product is longer lived than parent radionuclide.

f

 Radionuclide decays by both electron capture/positron decay (72.2%) and beta decay (27.8%).

g

 Half-life obtained from Parrington et al. (1996).
2–7

Table 2.1.2 Dosimetry Data for Selected Byproduct Materials

Dose Coefficient

Radionuclidea

Γb

(rem/Ci-h)

Air Submersionc

(rem-m3/Ci-yr)

Ingestiond, e

(rem/Ci)

Inhalationd, f

(rem/Ci)

3

H 6.4×105 9.6×105g

14C 3.1×104 2.1×103 2.1×103h

2.9×106i

2.5×105j

24Na 1.9×106 2.6×101

 1.4×103 1.2×103 D

32P 6.4×102 8.8×103 1.6×102

 W

35S 3.7×104 4.5×104 (0.8)

7.3×104 (0.1)

3.0×104

 D

2.5×103

 W

36Cl 2.0×102 3.0×103 2.2×103

 D

2.2×102

 W

45Ca 1.8×103 3.2×103 6.6×103

 W

46Sc 1.2×106 1.2×101 6.4×103 3.0×102

 Y

51Cr 2.3×108 1.8×101 1.5×104 (0.1)

1.5×104 (0.01)

1.1×104

 D

2.6×104

 W

3.3×104

 Y

54Mn 5.1×107 4.8 2.8×103 5.3×103

 D

6.7×103

 W

55Fe 0 6.1×104 2.7×103

 D

1.3×103

 W

57Co 1.5×107 6.6×101 7.4×104 (0.05)

1.2×103 (0.3)

2.6×103

 W

9.1×103

 Y

58Co 6.1×107 5.6 3.0×103 (0.05)

3.6×103 (0.3)

6.4×103

 W

1.1×102

 Y

59Fe 6.6×107 7.1 6.7×103 1.5×102

 D

1.2×102

 W

60Co 1.4×106 1.5×101 1.0×102 (0.05)

2.7×102 (0.3)

3.3×102

 W

2.2×101

 Y

63Ni 0 5.8×104 3.1×103

 D

2.3×103

 W

65Zn 3.3×107 3.4 1.4×102 2.0×102

 Y

See end of table for footnotes.
2–8

Table 2.1.2 Dosimetry Data for Selected Byproduct Materials (continued)

Dose Coefficient

Radionuclidea

Γb

(rem/Ci-h)

Air Submersionc

(rem-m3/Ci-yr)

Ingestiond, e

(rem/Ci)

Inhalationd, f

(rem/Ci)

75Se 8.6×107 2.2 9.6×103

 (0.8)

1.8×103

 (0.05)

7.2×103

 D

8.5×103

 W

82Br 1.6×106 1.5×101 1.7×103 1.2×103

 D

1.5×103

 W

85Kr 1.6×109 2.9×102

90Sr+90Y 1.1×101 1.5×101

 (0.3)

1.3×102

 (0.01)

2.5×101

 D 

1.3 Y 

99Tc 4.6×1013 3.4×103 1.5×103 1.0×103

 D 

8.3×103

 W 

106Ru+106Rh 1.4×107 1.3 2.7×102 5.6×102

 D 

1.2×101

 W 

4.8×101

 Y 

109Cd+109mAg 5.9×102 1.3×102 1.1×101

 D

4.0×102

 W

4.5×102

 Y

110mAg 1.7×106 1.6×101 1.1×102 4.0×102

 D

3.1×102

 W

8.1×102

 Y

113Sn+113mIn 4.2×107 1.5 3.2×103 4.0×103

 D

1.1×102

 W

125I 6.3×102 3.8×102 2.4×102

 D

129I 4.6×102 2.8×101 1.7×101

 D

131I 2.8×107 2.2 5.3×102 3.3×102

 D

133Ba 4.6×107 2.1 3.4×103 7.8×103

 D 

134Cs 1.0×106 9.0 7.3×102 4.6×102

 D

137Cs+137mBa 3.8×107 3.2 5.0×102 3.2×102

 D 

See end of table for footnotes.
2–9

Table 2.1.2 Dosimetry Data for Selected Byproduct Materials (continued)

Dose Coefficient

Radionuclidea

Γb

(rem/Ci-h)

Air Submersionc

(rem-m3/Ci-yr)

Ingestiond, e

(rem/Ci)

Inhalationd ,f

(rem/Ci)

140La 1.2×106 1.4×101 8.4×103 3.5×103

 D 

4.8×103

 W 

147Pm 1.1×103 2.6×102

 W 

3.9×102

 Y 

152Eu 7.4×107 6.7 6.5×103 2.2×101

 W

182Ta 7.7×107 7.6 6.5×103 2.2×102

 W

4.5×102

 Y

195Au 8.7×108 3.8×101 1.1×103 4.3×104

 D

4.2×103

 W

1.3×102

 Y

198Au 2.9×107 2.3 4.2×103 1.4×103

 D

3.0×103

 W

3.3×103

 Y

203Hg 2.5×107 1.3 2.3×103

 (0.02)

1.1×102

 (1.0)

5.8×103

 (0.4)

4.1×103

 D k

7.3×103 D l

5.7×103 W

204Tl 2.1×102 3.4×103 2.4×103

 D

210mBi 2.0×107m 1.4 9.6×102 8.3×101

 D

7.6 W

210Po 5.3×1012 4.9×105 1.9 9.4 D

8.6 W

241Am 3.1×107n 9.7×102 3.6 4.4×102 W

See following page for footnotes.
2–10

Footnotes to Table 2.1.2

a

 If shorter lived radioactive decay product is listed with parent radionuclide, dosimetry data

include contributions from both radionuclides based on assumption of activity equilibrium and

branching fraction for decay product given in Table 2.1.1. If shorter lived decay product given

in Table 2.1.1 is not listed, decay product is not dosimetrically significant compared with parent.

b

 Specific gamma-ray dose constant obtained from Unger and Trubey (1981), except as noted,

gives dose-equivalent rate per unit activity at distance of 1 meter from an unshielded point

source in air. 1 rem/Ci-h = 270 millisieverts (mSv)/MBq-h. c

 Values for external exposure from submersion in uniformly contaminated, semi-infinite

atmospheric cloud obtained from EPA–402–R–93–081. The skin dose component has been

included using a 0.01 weighting factor. 1 rem-m3

/Ci-yr = 8.6×1015 Sv-m3/Bq-s.

d

 Values for internal exposure obtained from Tables 2.1 and 2.2 EPA–520/1–88–020. The skin

dose component has been included using a 0.01 weighting factor. 1 rem/Ci = 2.7×10–7 Sv/Bq. e

 If more than one value is given, entry in parentheses is corresponding gastrointestinal-tract

absorption fraction. Assigned absorption fraction for different chemical forms of element is

given in Table 2.1.3.

f

 Assumed lung clearance class is denoted by D for days, W for weeks, or Y for years. 

Assigned clearance class for different chemical forms of element is given in Table 2.1.3.

g

 Value applies to tritiated water and is increased by factor of 1.5 to take into account

absorption through the skin (ICRP 30).

h

 Value applies to labeled organic compounds.

i

 Value applies to carbon monoxide.

j

 Value applies to carbon dioxide.

k

 Value applies to chemical forms with gastrointestinal-tract absorption fraction of 0.02 (see

Table 2.1.3).

l

 Value applies to chemical forms with gastrointestinal-tract absorption fraction of 1.0 (see

Table 2.1.3).

m Value determined by correlating photon abundance and fluence to the dose rate.

n

 Value provides conservative overestimate of effective dose equivalent rate, because

radionuclide emits photons with energies substantially below 100 keV.
2–11

Table 2.1.3 Gastrointestinal-Tract Absorption Fractions (f1) and Lung Clearance

Classes for Chemical Compounds of Selected Elements a

Ingestion Inhalation

Element Compound f1 Compound f1/Class

S (Sulfur) All inorganic forms

Elemental

 0.8

 0.1

Sulfates and sulfides–

 See associated elements

Elemental

 0.8 D

 0.8 W

 0.8 W

Cl (Chlorine) All forms 1.0 See assignment of

 associated element

1.0 D

 1.0 W

Cr (Chromium) Trivalent state

Hexavalent state

 0.01

 0.1

Oxides and hydroxides

Halides and nitrates

All others

 0.1 Y

 0.1 W

 0.1 D

Mn (Manganese) All forms 0.1 Oxides, hydroxides,

 halides, and nitrates

All others

 0.1 W

 0.1 D

Fe (Iron) All forms 0.1 Oxides, hydroxides,

 and halides

All others

 0.1 W

 0.1 D

Co (Cobalt) Oxides, hydroxides,

 and trace inorganics

Organic complexes

 and other inorganics

 0.05

 0.3

Oxides, hydroxides,

 halides, and nitrates

All others

 0.05 Y

 0.05 W

Ni (Nickel) All forms 0.05 Oxides, hydroxides,

 and carbides

All others

 0.05 W

 0.05 D

Se (Selenium) Elemental

All others

 0.05

 0.8

Oxides, hydroxides,

 carbides, and elemental

All others

 0.8 W

0.8 D

Br (Bromine) All forms 1.0 See bromide assignment

 of associated element

 1.0 D

 1.0 W

Sr (Strontium) Soluble salts

SrTiO3

 0.3

 0.01

SrTiO3

All others

 0.01Y

 0.3 D

Tc (Technetium) All forms 0.8 Oxides, hydroxides,

 halides, and nitrates

All others

 0.8 W

 0.8 D

See end of table for footnote.
2–12

Table 2.1.3 Gastrointestinal-Tract Absorption Fractions (f1) and Lung Clearance

Classes for Chemical Compounds of Selected Elements a

 (continued)

Ingestion Inhalation

Element Compound f1 Compound f1/Class

Ru (Ruthenium) All forms 0.05 Oxides and hydroxides

Halides

All others

 0.05 Y

 0.05 W

 0.05 D

Ag (Silver) All forms 0.05 Oxides and hydroxides

Nitrates and sulfides

All others

 0.05 Y

 0.05 W

 0.05 D

Cd (Cadmium) All inorganic forms 0.05 Oxides and hydroxides

Sulfates, halides, and

 nitrates

All others

 0.05 Y

 0.05 W

 0.05 D

Sn (Tin) All forms 0.02 Oxides, hydroxides,

 halides, nitrates,

 sulfides, and Sn3(PO4)4

All others

 0.02 W

 0.02 D

La (Lanthanum) All forms 1×103 Oxides and hydroxides

All others

1×103 W

1×103

 D

Pm (Promethium) All forms 3×104 Oxides, hydroxides,

 carbides, and fluorides

All others

3×104

 Y

3×104

 W

Ta (Tantalum) All forms 1×103 Oxides, hydroxides,

 halides, carbides,

 nitrates, and nitrides

All others

1×103

 Y

1×103

 W

Au (Gold) All forms 0.1 Oxides and hydroxides

Halides and nitrates

All others

 0.1 Y

 0.1 W

 0.1 D

Hg (Mercury) All inorganic forms

Methyl mercury

Other organic

 forms

 0.02

 1.0

 0.4

Oxides, hydroxides,

 halides, nitrates, and 

 sulfides

Sulfates

Organic forms

 0.02 W

 0.02 D

 1.0 D

Bi (Bismuth) All forms 0.05 Nitrates

All others

 0.05 D

 0.05 W

See end of table for footnote.
2–13

Table 2.1.3 Gastrointestinal-Tract Absorption Fractions (f1) and Lung Clearance

Classes for Chemical Compounds of Selected Elements a

 (continued)

Ingestion Inhalation

Element Compound f1 Compound f1/Class

Po (Polonium) All forms 0.1 Oxides, hydroxides, and

 nitrates

All others

 0.1 W

 0.1 D

a

 Assignments of gastrointestinal-tract absorption fractions and lung clearance classes obtained

from Table 3 of EPA–520/1–88–020. For elements not listed in table, dose coefficients for

ingestion and inhalation in Table 2.1.2 apply to all chemical forms, except as noted for

inhalation of 3

H and 14C.

2–15

2.2 Concentrations of Byproduct Material 

2.2.1 Introduction

In 10 CFR 30.14, persons who receive, possess, use, transfer, own, or acquire products or

materials containing byproduct material in concentrations that do not exceed the values for

specific radionuclides in gaseous or liquid and solid form listed in Schedule A of 10 CFR 30.70

are exempted from licensing requirements. However, the exemption does not authorize the

import of byproduct material or products containing byproduct material and the exemption does

not apply to the transfer of byproduct material contained in any food, beverage, cosmetic, drug,

or other commodity or product designed for ingestion or inhalation by, or application to,

humans.

Introduction of byproduct material in exempt concentrations into a product or material must be

performed under a specific Nuclear Regulatory Commission (NRC) or Agreement State license

that authorizes transfer of the product or material for use under 10 CFR 30.14. Requirements

for licensees who introduce byproduct material in exempt concentrations into products or

materials, and requirements for transfer of ownership or possession of such products or

materials, are specified in 10 CFR 32.11. These regulations specify that a licensee must

provide a description of the product or material into which the byproduct material will be

introduced, the intended use of the byproduct material and the product or material into which it

is introduced, the method of introduction, the initial concentration of the byproduct material in

the product or material, control methods to assure that no more than the specified

concentration is introduced into the product or material, the estimated time interval between

introduction and transfer of the product or material, and the estimated concentration of the

radionuclides in the product or material at the time of transfer. The licensee must also provide

reasonable assurance that the concentrations of byproduct material at the time of transfer will

not exceed the concentrations in 10 CFR 30.70, that reconcentration of the byproduct material

in concentrations exceeding those in 10 CFR 30.70 is unlikely, that use of lower concentrations

is not feasible, and that the product or material is not likely to be incorporated in any food,

beverage, cosmetic, drug, or other commodity or product designed for ingestion or inhalation

by, or application, to a human being.

This exemption was proposed on October 31, 1958 (23 FR 8428), and issued as a final rule on

August 17, 1960 (25 FR 7875). An additional exempt concentration for 85Sr was proposed on

November 13, 1969 (34 FR 18178), and issued as a final rule on March 3, 1970 (35 FR 3982).

The basis for the exempt concentrations of byproduct material is described in the Federal

Register notice from 1960 cited above. For each radionuclide, the exempt concentration for

gases or solids and liquids is the lowest value of the maximum permissible concentration (MPC)

in air or water, respectively, for occupational exposure over a 168-hour week given in Table 1 of

National Bureau of Standards (NBS) Handbook 69 (NBS, 1959). The values selected are those

for soluble chemical forms, which in general are lower than for insoluble forms. The exempt

concentrations are high enough to make quality control applications feasible from a

measurement standpoint and low enough to assure safety of the public.

Shortly thereafter, these MPCs were adopted into 10 CFR 20 as the limits for concentrations of

byproduct material in air and water and that would meet the annual dose limits for workers,
2–16

without consideration of external exposure, which was separately limited at the time. These

concentrations, if inhaled or ingested continuously over 1 year, were intended to correspond to

annual committed dose equivalents of approximately 0.05 sievert (Sv) (5 rem) to the whole

body or the gonads, 0.3 Sv (30 rem) to the thyroid, or 0.15 Sv (15 rem) to any other organ. The

MPCs in NBS Handbook 69 were calculated using the dosimetric and metabolic models in

Publication 2 of the International Commission on Radiological Protection (ICRP) (see

references). In adopting MPCs for occupational exposure to define exempt concentration of

byproduct material, the Atomic Energy Commission (AEC) reasoned that exempt

concentrations of by product materials would not generally be inhaled or ingested and that

continuous exposure over a year is highly unlikely. Therefore, in the AEC’s judgment, it is

highly improbable that any member of the public exposed to byproduct material in

concentrations less than the limits for exemption would receive an annual dose equivalent in

excess of a small fraction of 5 mSv (0.5 rem), which was the existing dose criterion for limiting

external exposure to members of the public (AEC, 25 FR 10914).

2.2.2 Description of Products or Materials

The Federal Register notice for the final rule cited previously indicates that exempt

concentrations of byproduct material would be permitted in such products or materials as oil,

gasoline, plastics, and similar commercial or industrial items that are unlikely to be ingested or

inhaled. However, as long as ingestion or inhalation by, or application to, humans is unlikely,

no restrictions are placed on products or materials to which the exemption can apply. Particular

examples of products or materials containing exempt concentrations of byproduct material in

use currently include irradiated topaz gemstones, irradiated silicon semiconductor materials,

engine oil, steel contaminated during production from use of byproduct material in blast furnace

refractory lining to provide an indication of wear, and luggage and cargo that have been

activated by irradiation with neutrons in an airport explosive detection system.

Recent information on the quantities of different radionuclides in various products or materials

containing exempt concentrations of byproduct material is summarized in Tables 2.2.1 to 2.2.5. 

The information in Tables 2.2.1 to 2.2.4 represents data submitted to the NRC in materials

licensee reports or other communications, whereas the information in Table 2.2.5 gives the

results of calculations based on an assumption that 1 kg of various elements would be

irradiated by neutrons in an airport explosive detection system. No other products or materials

containing exempt concentrations of byproduct material have been reported to the NRC in

recent years by materials licensees. Additional information on the quantities of different

radionuclides distributed as exempt concentrations is given in Table 2.2.6. These data

represent quantities reported to the NRC over a 20-year period prior to 1990.

The data in Tables 2.2.1 to 2.2.4 and 2.2.6 suggest that only a few of the more than 150

radionuclides for which exempt concentrations are given in Schedule A of 10 CFR 30.70 have

been distributed in significant amounts under this exemption. Furthermore, in only a few cases

is the concentration of a radionuclide distributed in a particular product or material within an

order of magnitude of the exempt concentration. In regard to the results in Table 2.2.5, the

actual concentrations of most of the listed radionuclides in irradiated luggage should be

considerably less than the calculated values, because the quantity of the parent stable element

in luggage normally should be considerably less than 1 kg.
2–17

It is difficult to estimate the total activity of various radionuclides that has been distributed as

exempt concentrations based on the available information. Not only is information on total

activity distributed not given in some of the materials licensee reports sent to the NRC, but

these reports do not take into account radionuclides distributed under this exemption by

Agreement State licensees. In the case of exempt quantities of byproduct material which can

be distributed either by an NRC license or an Agreement State license, it was assumed that

twice the quantity was distributed by Agreement State licensees as NRC licensees. 

2.2.3 Summary of Previous Analyses and Assessments

As discussed in Section 2.2.1, the Federal Register notices establishing this exemption

indicated that dose equivalents to members of the public from exposure to products or

materials containing exempt concentrations of byproduct material should be a small fraction of

the criterion for limiting dose to members of the public. However, quantitative analyses of

doses for particular products or materials and particular exposure scenarios were not

presented. In addition, it is not evident that external exposure to exempt concentrations of

byproduct material was considered, even though external exposure should be more important

than internal exposure for products or materials containing exempt concentrations of

photon-emitting radionuclides.

Studies by other investigators have provided information on doses to the public from exposure

to particular products or materials containing exempt concentrations of byproduct material,

including irradiated topaz gemstones, irradiated silicon semiconductor materials, steel

contaminated with 60Co used in blast furnace refractory lining to provide an indication of wear,

and the contents of baggage that has been activated by irradiation with neutrons in an airport

explosive detection system. The following sections summarize the available information on

doses for these particular products or materials.

2.2.3.1 Irradiated Topaz Gemstones

External doses to individuals and populations while wearing topaz gemstones during the first

year after irradiation have been estimated by Nelson and Baum (NUREG/CR–5883). Doses in

the first year are substantially higher than in subsequent years, because of the relatively short

half-lives of the radionuclides of concern (see Table 2.2.1). Nelson and Baum calculated the

doses by assuming that the gemstones contain exempt concentrations of various radionuclides

and are worn 8 h/day for 365 days/yr. The maximum individual dose from photon exposure was

calculated by assuming that an individual wears a single 30-carat (6-g) stone, and the

maximum individual dose from beta exposure was calculated by assuming that an individual

wears six 5-carat (1-g) stones. The collective dose from photon exposure was calculated by

assuming, based on recent distribution data for gemstones, that the exposed population is

2.25 million and that an average individual in this population wears a single 5-carat (1-g) stone.

The external doses to individuals calculated by Nelson and Baum were 3.6×103

 mSv

(0.36 mrem), effective dose equivalent (EDE), and 0.03 Sv (3 rem), skin dose equivalent to a

small area (1 cm2

). The calculated doses for beta exposure to the skin are conservative

because they were equated to the dose internal to the topaz gemstone and did not consider

any shielding that might be provided by a gemstone mounting and, did not include the

0.7 mg/cm2

 dead skin layer.
2–18

2.2.3.2 Irradiated Silicon Semiconductor Materials

Individual doses from exposure to irradiated silicon semiconductor materials have been

estimated by the NRC staff (NRC, Memoranda, Paperiello, 1994). Doses were estimated for

the following exposure scenarios: (1) external and internal exposure to workers at unlicenced

facilities during processing and assembly of irradiated materials into electronic components,

(2) exposure to members of the public from disposal into sanitary sewer systems of silicon fines

generated during cutting and lapping operations, and (3) exposure to members of the public

during a fire at a facility for processing and assembly of irradiated materials.

The dose estimates obtained by the NRC staff (NRC, Memoranda, Paperiello, 1994) for the

assumed exposure scenarios are summarized in Table 2.2.7. These estimates were based on

data on the concentrations of impurity byproduct materials in the irradiated semiconductor

materials provided by the Missouri University Research Reactor (MUIR) facility. With few

exemptions, the upper bounds on the reported concentrations of byproduct materials in the

irradiated materials were about 10% or less of the corresponding exempt concentrations, in

agreement with other data given in Tables 2.2.2 and 2.2.3. Based on the MUIR data, the NRC

concluded that the dose in all scenarios would result primarily from exposure to 152Eu, which

has a half-life of 13.6 years.

Potential inhalation and external exposure to workers at unlicenced facilities that process

irradiated silicon semiconductor materials also were considered by Morris (1993). Inhalation

doses for various radionuclides were calculated by assuming that each radionuclide would be

present in the materials at its exempt concentration and that the airborne concentration of

silicon during the work activities would be 10 mg/m3

. The latter is a level above which use of

respiratory protection has been recommended by the American Conference of Governmental

Industrial Hygienists. Based on these assumptions, the highest annual EDE from inhalation

was 0.035 mSv (3.5 mrem) for 152Eu, and the calculated doses for most of the other

radionuclides were less than 0.01 mSv (<1 mrem). The doses calculated by Morris (1993)

should be quite conservative, because the concentrations of each radionuclide in the irradiated

materials normally are considerably less than the exempt concentrations (e.g., see Tables 2.2.2

and 2.2.3) and the airborne concentrations of silicon would not normally be as high as 10 mg/m3

when the materials are processed in wet form, which is the usual procedure.

Morris (1993) also calculated dose to the skin from external exposure to 32P. Based on the

conservative assumptions that this radionuclide would be present at its exempt concentration,

that half of the beta particles in sources near the surface of the silicon would be emitted

perpendicular to the surface, and that a worker would be in direct contact with the silicon for a

working year of 2000 hours, the estimated annual dose equivalent to the skin was less than

0.01 mSv (<1 mrem).

2.2.3.3 Steel Contaminated with 60Co

Dose rates from external exposure to steel contaminated with 60Co used in blast furnace

refractory lining have been estimated by Leoben (NRC, Memoranda, 1996). External dose

rates near a steel slab measuring 1.3 m × 1.3 m × 0.66 m were estimated based on a variety of

measurements and calculations, and the results are summarized in Table 2.2.8.
2–19

The exempt concentration for 60Co is 19 becquerel (Bq)/g (500 picocurie (pCi)/g). Based on the

results in Table 2.2.8, Leoben concluded that external exposure at locations near contaminated

steel containing the exempt concentration of 60Co could result in dose-equivalent rates on the

order of 0.01 mSv/h (1 mrem/h), and that the resulting doses to members of the public from

exposure to contaminated steel thus could be unacceptably high.

In practice, however, the presence of 60Co in contaminated steel at levels approaching the

exempt concentration of 19 Bq/g (500 pCi/g) is quite unlikely. For example, the data in

Table 2.2.4 indicate that the concentrations of 60Co in contaminated steel normally would be

about three orders of magnitude less than the exempt concentration, and the external dose

rates and doses to members of the public would be reduced accordingly. In addition, for a

given concentration of 60Co in steel, the dose rate from a large slab overestimates the dose rate

from any smaller sources. Therefore, the dose rate would be reduced somewhat for steel

products that are much less massive than the steel slab considered by Leoben (NRC,

Memoranda, 1996), including, for example, appliances, furniture, and parts used in

automobiles.

2.2.3.4 Baggage Irradiated by Airport Explosive Detection System

Individual and collective doses to the public from exposure to baggage that has been irradiated

by neutrons from spontaneous fission of 252Cf in an airport explosive detection system (SAIC,

1988) have been estimated by Randolph and Simpson (1988). Doses from external exposure

to the 252Cf source itself also could occur, but these doses are not of concern for this

exemption. Doses were estimated for external exposure to the contents of baggage and for

ingestion of activated food in the baggage. Doses from external exposure are limited by the

presence of a detection and alarm system that prevents delivery of irradiated baggage for

loading on aircraft when the external dose rate exceeds 0.005 mSv/h (0.5 mrem/h) (SAIC,

1988).

External dose rates near baggage from exposure to different radionuclides were calculated by

Randolph and Simpson (1988) based on an assumption that 1 kg of various stable elements is

irradiated, and the results are given in Table 2.2.9. Neutron activation of the important

elements in clothing (i.e., hydrogen, carbon, nitrogen, and oxygen) is insignificant. By

considering the radionuclides that could be produced in significant amounts (i.e., the stable

elements with substantial cross-sections for neutron activation that also could be present in

significant amounts in baggage and its contents), external doses were estimated for the

following exposure scenarios: (1) exposure to a baggage handler to 28Al at 30 seconds after

irradiation over a normal working year of 2000 hours, (2) exposure to a passenger to 56Mn from

reclaiming of luggage 1 hour after irradiation and exposure for 1 hour thereafter, (3) exposure

to a passenger to 152mEu in concentrations that would just pass the baggage release criterion

for the explosive detection system of 0.005 mSv/h (0.5 mrem/h) during a 3-hour car trip and

from placement of the luggage near the individual for the next 12 hours, (4) exposure to the

skin while wearing a 40-g gold medallion continuously for 10 days after luggage is reclaimed,

and (5) exposure to the skin from application of cosmetics 1 hour after irradiation.

The dose estimates for the scenarios for external exposure described above are summarized in

Table 2.2.10. The dose estimate for baggage handlers should be very conservative, because it

assumes continuous exposure at a distance of 30 cm throughout a normal working year and

that all luggage contains 1 kg of aluminum. The dose estimate for passengers exposed to 56Mn
2–20

would be conservative for luggage containing less than 1 kg of manganese, which normally

should be the case. The dose estimate for exposure to 152mEu should be very conservative,

given the small amounts of europium that normally should occur in luggage. The dose estimate

for this scenario is intended primarily to provide an upper bound on doses from exposure to

longer lived activation products. It also should be noted that the assumed concentration of

152mEu for this scenario, a concentration that would just pass the baggage release criterion for

the explosive detection system, is nearly three orders of magnitude greater than the exempt

concentration.

Randolph and Simpson (1988) also estimated internal doses resulting from ingestion of

irradiated food. The dose for this scenario was estimated by assuming that a 1-day supply of

food was packed in a suitcase and then consumed 1 hour after irradiation. The estimated

committed EDE for this scenario also is given in Table 2.2.10. The dose for this scenario is due

primarily to irradiation of salt. Therefore, this scenario could be appropriate for individuals

consuming salt pills or highly salted food carried in their luggage.

2.2.4 Present Exemption Analysis

A rigorous quantitative assessment of potential radiological impacts on the public from use of

products or materials containing exempt concentrations of byproduct material is a difficult

undertaking, due in part to the intentional lack of specificity regarding the particular products or

materials to which this exemption may be applied. It is particularly difficult to obtain reasonable

bounding estimates of individual dose when there are no specified limits on volume, total

activity, or external dose for this exemption, and it is difficult to obtain estimates of collective

dose when complete information on the distribution and activity content of different products or

materials containing exempt concentrations of byproduct material is lacking.

The difficulty in obtaining reasonable bounding estimates of individual dose from exposure to

exempt concentrations of byproduct material is illustrated by the data in Table 2.2.8 on external

dose rates near a steel slab contaminated with 60Co, as discussed in Section 2.2.3.3. Based on

these data, the annual dose equivalent from external exposure for an individual who might be

located for 2000 h/yr near a source (about 0.3 meter) containing an exempt concentration of

60Co of 19 Bq/g (500 pCi/g) would be 10 to 20 mSv (1 to 2 rem). Using the MicroShield code

(Computer Codes, Grove Engineering, 1996), the calculated annual dose for a 1-meter

exposure distance would be about 4 mSv (400 mrem). Such an estimate presumably provides

an upper bound on the dose from external exposure to the maximum allowed exempt

concentration of 60Co in a material. However, this estimate does not provide a reasonable

upper bound on external dose from 60Co in steel that is actually distributed primarily because,

as indicated in Table 2.2.4, the concentrations of 60Co that actually occur in contaminated steel

are at least three orders of magnitude less than the exempt concentration limit.

In this assessment, simple scenarios are used to estimate dose from external and internal

exposure to specific products or materials containing exempt concentrations of byproduct

material. These scenarios are intended to provide reasonable upper bounds on doses that

might be experienced by individual members of the public from routine use or from accidents

and misuse, based on current practices. This assessment did not attempt to estimate the

doses from external and internal exposures to the maximum allowed concentrations of
2–21

byproduct material authorized in 10 CFR 30.14 where such concentrations are not known to be

distributed.

2.2.4.1 External Exposure to Individuals During Routine Use

As an example of potential external doses that might be received by individual members of the

public from exposure to products or materials containing exempt concentrations of byproduct

material, this analysis considers exposure to topaz gemstones that have been activated by

neutron irradiation for the purpose of enhancing color and appearance. As indicated in

Table 2.2.1, irradiated topaz gemstones commonly contain concentrations of some

radionuclides that are within an order of magnitude of the corresponding exempt

concentrations. Furthermore, gemstones often are worn for extended periods of time, and they

are located very close to the body while they are worn. Thus, an assessment of external dose

while wearing irradiated topaz gemstones should provide a reasonable upper bound on external

dose to individuals during routine use for any of the reported products or materials containing

exempt concentrations of byproduct material.

The individual dose from external exposure to a large topaz gemstone containing exempt

concentrations of various radionuclides has been estimated and the results are summarized in

Table 2.2.11. Doses were calculated using MicroShield (Computer Codes, Grove Engineering,

1996), assuming an individual wears a single 30-carat (6-g) stone for 8 h/day, 365 days/yr. The

beta skin dose was calculated using VARSKIN (Computer Codes, Durham, 1992) with the

same exposure assumptions. The EDE (photon) was calculated at a body depth of 10 cm,

which, as discussed in Appendix A.4, is considered a reasonable approximation for the average

depth of the body organs relative to a small source on the body surface. The dose equivalent

to the skin was calculated for a spherical source in contact with the skin.

An upper bound estimate of external dose can be obtained directly from the results in

Table 2.2.11, because the calculations assume that a gemstone contains an exempt

concentration of each radionuclide. Based on the results the EDE during the first year after

irradiation would be 0.01 mSv (1 mrem). The annual dose would decrease after the first year

because of the short half-lives of the radionuclides of concern. 

A more realistic estimate of dose from wearing of irradiated topaz gemstones can be obtained

by combining the calculations for exempt concentrations in Table 2.2.11 with the data in

Table 2.2.1 on the estimated concentrations of various radionuclides in gemstones. With the

assumption that a gemstone contains the maximum concentrations of the five radionuclides

listed in Table 2.2.1, the EDE during the first year after irradiation would be about 3×104

 mSv

(0.03 mrem).

An important consideration for this exemption is the so-called sum-of-fractions rule, which

states that for mixtures of radionuclides in any material, the sum over all radionuclides of the

ratio of the concentration of each radionuclide to its exempt concentration may not exceed unity

(see Note 2 to Schedule A of 10 CFR 30.70). Based on the maximum reported concentrations

in Table 2.2.1, the sum of fractions for the dose estimate obtained above is unity. Therefore,

the more realistic estimate of dose should be somewhat conservative for an average gemstone

containing exempt concentrations.
2–22

Estimates of dose to the skin from beta exposure while wearing irradiated topaz gemstones

have been calculated using VARSKIN and are summarized in Table 2.2.11. For a gemstone

containing an exempt concentration of 182Ta, which gives the highest estimate of dose to the

skin for any of the radionuclides listed in Table 2.2.11 and, thus, provides an upper bound on

dose to the skin from wearing of gemstones containing mixtures of radionuclides at levels less

than their exempt concentrations, the annual dose equivalent to the irradiated portion of the

skin would be about 0.3 mSv (30 mrem) and the annual dose equivalent to the whole skin

would be less than 2×104

 mSv (0.02 mrem). The skin dose component has been included

assuming an exposed area of 10 cm2

. If the dose to the whole skin is included in the EDE with

a weighting factor of 0.01 (ICRP 60), the contribution from exposure to the skin would be less

than 1 mSv (<0.001 mrem). Thus, the dose to the skin from beta exposure from wearing of

irradiated topaz gemstones is insignificant compared with the EDE from photon exposure to the

whole body.

Estimates of external dose to individuals from routine exposure to other products or materials

containing exempt concentrations of byproduct material can be obtained from the previous

assessments discussed in Section 2.2.3. First, as estimated by Paperiello (NRC, Memoranda,

1994) and summarized in Table 2.2.7, the annual EDE to individual workers during wet

processing and assembly of irradiated silicon semiconductor materials into electronic

components would be 0.008 mSv (0.8 mrem). The dose estimate for wet processing in

Table 2.2.7 is adopted because dry processing of irradiated semiconductor materials is not

normally practiced (NRC, Memoranda, Paperiello, 1994). This dose estimate, which was based

on reported concentrations of various radionuclides in the irradiated materials and includes a

relatively small contribution from internal exposure, should be conservative, because it assumes

that all of the silicon that would be irradiated each year in a single reactor would be processed

in the same facility. Doses to other members of the public from use of the irradiated

semiconductor materials should be far less.

Second, based on the reported concentrations of 60Co in steel (Kobrick, 1991) summarized in

Table 2.2.4 and MicroShield (Computer Codes, Grove Engineering, 1996), the annual EDE to

an individual who is assumed to be exposed for 2000 h/yr at an average distance of 1 meter

from the source would be on the order of 0.004 mSv (0.4 mrem). The assumed exposure time

and distance from the source should be conservative for most exposure situations.

Finally, estimates of EDEs to individuals from exposure to checked luggage that has been

irradiated in an airport explosive detection system can be based on the analysis of Randolph

and Simpson (1988) summarized in Table 2.2.10. However, some of the results in this table do

not represent doses that reasonably could be experienced. For example, the dose from

exposure to 152mEu is unreasonable because it assumes that luggage would contain about

100 kg of europium (see Footnote e of Table 2.2.10). If the results for exposure to airline

passengers to 56Mn in irradiated luggage are assumed to be reasonably representative, the

annual EDE to an individual who is assumed to travel by plane twice a week would be about 

0.002 mSv (0.2 mrem). The assumed exposure time for this scenario should be conservative

for most individuals. Table 2.2.10 also gives a dose estimate for baggage handlers. However,

this estimate appears to be unreasonably conservative, given the short half-life of 28Al

(2.24 min), the assumed exposure time of 2000 h/yr, and the assumption that all luggage

contains 1 kg of aluminum. A more reasonable dose estimate for baggage handlers probably

would be at least an order of magnitude lower, i.e., less than 0.01 mSv/yr (<1 mrem/yr).
2–23

The foregoing assessment indicates that external doses to individual members of the public

from routine use of currently distributed products or materials containing exempt concentrations

of byproduct material should be no more than 0.01 mSv/yr (1 mrem/yr) in the worst credible

cases and probably are considerably less for many realistic exposure situations. Although the

exempt concentrations of byproduct material were not based on considerations of external

dose, the values are relatively low for most photon-emitting radionuclides in the concentrations

currently distributed, because these radionuclides also tend to have relatively high doses per

unit intake by inhalation or ingestion and the relatively low exempt concentrations for photon

emitters serve to limit doses from external exposure. External doses from photon-emitting

radionuclides were not calculated at the maximum allowed concentrations (except for 60Co in

steel), because no current uses are known to exist. Doses from external exposure are further

limited by two additional considerations. First, for products or materials that contain

concentrations of byproduct material approaching the exempt concentrations (e.g., irradiated

topaz gemstones), the source volumes generally are small and the total activity per source thus

is low. Second, for products or materials with larger volumes (e.g., steel products), the reported

concentrations of byproduct material appear to be much less than the exempt concentration

limits.

2.2.4.2 Internal Exposure to Individuals During Routine Use

As indicated in Section 2.2.1, potential inhalation or ingestion exposures to exempt

concentrations of byproduct material are inherently limited by two factors. The first is the

condition that exempt concentrations of byproduct material should not be contained in any

product or material designed for intake by, or application to, humans. Thus, continuous internal

exposure over a year would be highly unlikely. The second factor is the definition of exempt

concentrations in terms of MPCs in air or water for occupational exposure.

In this assessment, estimates of internal dose to individuals from routine exposure to particular

products or materials containing exempt concentrations of byproduct material are obtained from

the previous dose assessments discussed in Section 2.2.3. First, based on the concentrations

of byproduct material in irradiated silicon semiconductor materials reported by a materials

licensee, Paperiello (NRC, Memoranda, 1994) estimated that the annual EDE to individual

workers from inhalation and ingestion of silicon dust during processing and assembly of the

materials into electronic components could be as high as 0.004 mSv (0.4 mrem) (see

Footnote c of Table 2.2.7). The estimated internal dose in this case is due almost entirely to

inhalation (NRC, Memoranda, Paperiello, 1994), and the dose from ingestion is negligible. This

dose estimate should be conservative, because it is based on the assumptions that lapping and

cutting of the silicon take place under dry conditions, rather than the usual practice of wetting

the material to reduce dust generation, and that the resulting concentration of silicon in air

would be equal to the limit for occupational exposure to 10 mg/m3

 recommended by the

American Conference of Governmental Industrial Hygienists. A more reasonable upper bound

on the internal dose in this case thus might be on the order of 0.001 mSv (0.1 mrem).

Second, estimates of dose to individuals from ingestion of food irradiated by neutrons in an

airport explosive detection system can be based on the analysis of Randolph and Simpson

(1988) summarized in Table 2.2.10. For an individual who is assumed to travel by plane twice a

week and to carry a 1-day supply of food in irradiated luggage on each trip, consumed 1 hour

after irradiation, the annual EDE would be about 3×105

 mSv (0.003 mrem). This dose estimate

should be quite conservative for most travelers.
2–24

The foregoing assessment indicates that doses from inhalation and ingestion of products or

materials containing exempt concentrations of byproduct material for routine exposure

situations should be no more than 0.001 mSv/yr (0.1 mrem/yr) and usually should be

considerably less. The assessment also indicates that doses from inhalation and ingestion

exposure normally should be considerably less than doses from external exposure. This is a

reasonable result, given the condition for the exemption that products or materials containing

exempt concentrations of byproduct material should be unlikely to be inhaled or ingested and

the definition of exempt concentrations in terms of MPCs in air or water for occupational

exposure.

2.2.4.3 Collective Dose During Routine Use

The collective dose during routine use of all products or materials containing exempt

concentrations of byproduct material is difficult to estimate, due primarily to the lack of

information on the total amounts of byproduct materials distributed under this exemption. 

However, the collective dose from routine use of particular products or materials can be

estimated based on available information, as discussed below.

At the present time, topaz gemstones appear to be the most commonly used product containing

exempt concentrations of byproduct material. Therefore, estimates of collective dose in this

case should not seriously underestimate the total collective dose associated with this

exemption. In this assessment, the collective dose from wearing of the 2.25 million irradiated,

5-carat topaz gemstones distributed per year is estimated by adjusting the individual dose for

each radionuclide in Table 2.2.11 based on the upper bound of the average reported

concentrations of radionuclides in gemstones, as obtained from the range of concentrations in

Table 2.2.1. Using this procedure, the estimated collective EDE during the first year is

0.6 person-Sv (60 person-rem). If the gemstones are assumed to be worn for 10 years (i.e.,

several half-lives of the radionuclides of concern), the integrated collective dose for 2.25 million

gemstones distributed per year would be about 0.9 person-Sv (90 person-rem). About

two-thirds of the total collective dose is due to 182Ta and 134Cs, with the remaining one-third due

to 65Zn, 54Mn and 46Sc.

Crude estimates of collective dose from routine use of other products or materials containing

exempt concentrations of byproduct material can be obtained as described in the following

paragraphs.

First, during processing of irradiated silicon semiconductor materials, the estimated annual EDE

to individual workers from external and internal exposure obtained from Sections 2.2.4.1

and 2.2.4.2 is 0.01 mSv (1 mrem). This estimate was based on an assumption that all

materials irradiated in a single reactor would be processed in a single facility. Therefore, if it is

arbitrarily assumed that 100 workers would be exposed at a single facility during processing of

the irradiated materials, the resulting annual collective EDE would be 0.001 person-Sv

(0.1 person-rem).

Second, for external exposure to contaminated steel containing 60Co, the annual EDE to an

individual estimated in Section 2.2.4.1 is 0.004 mSv (0.4 mrem). This estimate applies to a

concentration of 60Co in a large steel slab of about 18 mBq/g (5×107 Ci/g) (see Table 2.2.4)

and an exposure time of 2000 h/yr at an average distance from the source of 1 meter. For the

purpose of estimating collective dose, it is assumed that a single individual is exposed to a
2–25

contaminated steel slab containing the entire amount of 60Co introduced into steel in a year by a

large steel producer of 30 GBq (0.8 Ci) (see Table 2.2.4). For the assumed dimensions of the

slab given in Section 2.2.3.3 and Table 2.2.8 and an assumed density of steel of 7.9 g/cm3

, the

resulting concentration of 60Co in the steel would be 3.3 kBq/g (0.09 Ci/g). Then it is assumed

that an average individual in the exposed population would be located for 1000 h/yr at an

average distance from the source of 2 meters. Based on these assumptions, the collective

EDE in the first year would be about 0.1 person-Sv (10 person-rem). If the assumption is a

useful lifetime for the steel product of 10 years and it takes into account the half-life of 60Co, the

collective EDE from 1 year’s distribution of 60Co in contaminated steel would be about

0.6 person-Sv (60 person-rem). This estimate, although quite uncertain, is believed to be

conservative, because steel often is used in products or materials (e.g., bridges) that are not

located as near to members of the public, on the average, as the distance of 2 meters assumed

in this assessment. In addition, as noted in Section 2.2.3.3, the estimated individual dose is

conservative for sources that are considerably smaller than a large slab. Actual doses would

be expected to be a small fraction of this estimate. This result mainly indicates that the

collective dose from routine exposure to exempt concentrations of 60Co in contaminated steel

should be less than the collective dose from exposure to irradiated topaz gemstones.

Third, based on the data in Table 2.2.6, it is assumed that 0.74 TBq (20 Ci) of tritium (3

H) per

year is distributed as exempt concentrations. With the assumption that the average release

rate of 3

H from a product or material is 1 ppm/h (see Appendix A.3), the total annual release of

3

H would be about 1% of the total inventory, or about 7.4 GBq/yr (0.2 Ci/yr). For the purpose of

estimating collective dose, it can be assumed that this release occurs in a single laboratory in

which a single individual is located, and that inhalation and absorption through the skin are the

only significant exposure pathways. For a spill of 3

H in a laboratory, Table A.1.8 of

Appendix A.1 gives an EDE for inhalation of 2.6×108

 Sv/GBq (9.6×1011 rem/Ci). This doseto-source ratio assumes a release fraction of 0.1%, so the value needs to be increased by a

factor of 1000 for application to the scenario described above. Based on these assumptions,

the collective EDE from 1 year’s distribution of 3

H would be about 2×104 person-Sv

(0.02 person-rem). This result, although quite uncertain, indicates that the collective dose from

routine exposure to exempt concentrations of 3

H should be insignificant compared with the

collective dose from exposure to irradiated topaz gemstones.

Finally, in an assessment of doses from irradiation of luggage in an airport explosive detection

system, Randolph and Simpson (1988) estimated the collective dose to passengers who

reclaim luggage 1 hour after irradiation. As summarized in Table 2.2.10, the estimated annual

collective EDE from external exposure is about 0.02 person-Sv (2 person-rem) per million

passengers. This estimate should be quite conservative because it assumes that all irradiated

luggage contains 1 kg of manganese. The annual collective EDE from wearing of irradiated

gold medallions, as obtained from Table 2.2.10, by applying the weighting factor for skin of 0.01

(ICRP 60) would be 0.004 person-Sv (0.4 person-rem) per million passengers per year. The

collective dose from consumption of irradiated food carried in luggage estimated by Randolph

and Simpson (1988) is about four orders of magnitude less than the collective dose from

external exposure to irradiated luggage. These results indicate that the collective dose from

routine exposure to irradiated luggage should be small compared with the collective dose from

exposure to irradiated topaz gemstones.
2–26

Based on the foregoing assessment, the collective EDE from 1 year’s distribution of exempt

concentrations of byproduct material in accordance with current practices appears to be less

than 1 person-Sv (<100 person-rem), due primarily to exposure to irradiated topaz gemstones. 

2.2.4.4 Doses During Distribution and Transport

In this assessment, doses during distribution and transport of exempt concentrations of

byproduct material are estimated for the case of irradiated topaz gemstones. As noted

previously, such gemstones appear to be the most commonly used product containing exempt

concentrations of byproduct material at the present time. Therefore, dose estimates for this

case should provide reasonable representations of doses from distribution and transport of

exempt concentrations of byproduct material in all products or materials.

The following assumptions are used in the dose assessment for distribution and transport of

irradiated topaz gemstones. First, as in the assessment of collective dose from wearing of

gemstones in the previous section, 2.25 million gemstones are assumed to be distributed per

year. Second, the gemstones are assumed to be irradiated at two facilities, with each facility

thus distributing more than 1 million gemstones per year. Third, the gemstones are assumed to

be distributed equally among 10,000 retail stores; i.e., each store is assumed to receive 225

irradiated gemstones each year. Fourth, the concentrations of different radionuclides in each

gemstone are assumed to be the average of the range of values given in Table 2.2.1. Finally,

the mass of an average gemstone is assumed to be 1 g (see Footnote d of Table 2.2.11). 

Based on the last two assumptions, the activity of the radionuclides in a single gemstone is

assumed to be 1.5 Bq (4×105 Ci) for 46Sc, 1.3 Bq (3.5×105 Ci) for 54Mn, 1.9 Bq (5×105 Ci)

for 65Zn, 0.56 Bq (1.5×105 Ci) for 134Cs, and 4.8 Bq (1.3×104 Ci) for 182Ta.

The following distribution and transportation system is assumed in this assessment. First, the

gemstones are assumed to be shipped primarily by ground parcel delivery. Second, a local

parcel delivery driver in a small truck is assumed to pick up the gemstones from the irradiation

facility and transport them to a local terminal, and similarly for transport to retail stores. Finally,

it is assumed that semi-trucks are used to transport the gemstones between terminals, and that

the gemstones are transported to an average of four regional terminals before delivery to retail

stores.

Individual and collective doses during distribution and transport are estimated using the generic

methodology in Appendix A.3. Based on the assumptions described above, the highest

individual doses would be received either by the truck driver that picks up the gemstones from

the irradiation facility or by workers in retail stores. Doses to other individuals would be less,

either because the number of gemstones present would be reduced (e.g., for truck drivers in

the transportation legs after the initial transport from the irradiation facility to a local terminal) or

because the exposure times would not be as high.

Doses to individual truck drivers during local delivery in a small truck can be obtained based on

the results for 46Sc in Table A.3.1 of Appendix A.3. The dose from the other radionuclides

assumed to be present in the gemstones can be estimated by scaling the dose estimate for

46Sc in accordance with the specific gamma-ray dose constants given in Table 2.1.2. If a single

truck driver is assumed to be exposed to all 1 million of the gemstones distributed per year by a

single irradiation facility, the annual EDE is estimated to be 0.005 mSv (0.5 mrem), half of which

results from exposure to 182Ta, about one-fourth to 46Sc, and about one-fourth to 54Mn, 65Zn, and
2–27

134Cs. This dose estimate should be conservative because a single individual is assumed to be

exposed to all gemstones distributed by a single facility. However, the dose estimate may not

be extreme, because there are few irradiation facilities and a single driver could be exposed to

a large fraction of the total number of gemstones shipped from one facility.

Similarly, the dose to individuals working in small retail stores can be obtained based on the

results for 46Sc in Table A.3.7 of Appendix A.3. Recalling that each store is assumed to contain

225 irradiated gemstones, the annual EDE to individual workers is estimated to be less than

1×105

 mSv (<0.001 mrem). Thus, the dose to individual workers in retail stores should be

much less than the dose to individual truck drivers during shipment from an irradiation facility to

a local terminal.

As described previously, the collective dose during distribution and transport of gemstones is

estimated by assuming two shipments in small express delivery trucks (i.e., the initial pickup

from irradiation facilities and the final delivery to retail stores), three shipments between

terminals in semi-trucks, and temporary storage in four terminals (i.e., large warehouses). For

the assumed annual distribution of 2.25 million gemstones, each containing the activities of the

various radionuclides listed previously, the annual collective EDE is estimated to be about

0.01 person-Sv (1 person-rem). The collective dose during distribution and transportation

results almost entirely from exposures in retail stores.

Thus, in summary, based on information about the amounts of various radionuclides that are

present in irradiated topaz gemstones, the following estimates of dose during distribution and

transport of 2.25 million such gemstones per year are obtained:

The annual EDE to individual truck drivers during the initial shipment of gemstones from

irradiation facilities to a local terminal could be as high as 0.005 mSv (0.5 mrem).

The annual collective EDE, most of which would be from exposures in retail stores,

would be about 0.01 person-Sv (1 person-rem).

The estimated dose for individual truck drivers is more than an order of magnitude higher than

the best estimate of individual dose to a wearer of gemstones given in Section 2.2.4.1. 

However, the dose estimate for truck drivers should be conservative, because it is based on an

assumption that a single driver would be exposed to all 1 million gemstones distributed per year

by a single irradiation facility. On the other hand, the estimated annual collective dose during

distribution and transport is a factor of 60 less than the estimate for wearers of gemstones

given in Section 2.2.4.3. This is a reasonable result, given the greater exposure times and

smaller distances from the source for wearers of gemstones.

Doses during distribution and transport of contaminated steel containing 60Co also could be of

concern. However, the following arguments indicate that individual and collective doses in this

case should be significantly less than the doses for irradiated topaz gemstones obtained in this

assessment and given above.

As discussed in Section 2.2.4.1, the annual EDE to an individual who is assumed to be exposed

for 2000 hours at a distance of 1 meter from a source of contaminated steel containing 60Co

should be less than about 0.004 mSv (<0.4 mrem). Therefore, since the average distance from

a source during distribution and transport normally would be at least 1 meter (see
2–28

Appendix A.3), the annual individual dose during distribution and transport normally should also

be less than about 0.004 mSv (<0.4 mrem). Furthermore, the total activity of 60Co introduced

into steel annually, as estimated from the data in Table 2.2.4, is significantly less than the total

activity of the radionuclides in the 2.25 million irradiated topaz gemstones assumed to be

distributed annually. Therefore, the collective dose during distribution and transport of steel

containing 60Co also should be significantly less than the estimate for gemstones obtained in

this assessment.

The assessment for 60Co described above supports the previous assertion that individual and

collective doses from distribution and transport can be represented by the results for irradiated

topaz gemstones. Doses from distribution and transport should not be an important concern for

any of the other products or materials considered in this assessment.

2.2.4.5 Doses from Disposal

For many products or materials containing exempt concentrations of byproduct material, the

individual and collective doses from disposal would be much less than the doses during routine

use or distribution and transport, primarily because the useful lifetime of the products or

materials should be much greater than the half-lives of the most important radionuclides of

concern. This is the case, for example, with irradiated topaz gemstones and irradiated silicon

semiconductor materials, which contain mostly short-lived radionuclides (see Tables 2.2.1 to

2.2.3).

In this assessment, individual and collective doses from disposal of products or materials

containing exempt concentrations of byproduct material are estimated using the generic

methodology described in Appendix A.2 and data on the quantities of different radionuclides

distributed as exempt concentrations. Based on the data in Table 2.2.6 for the years 1970 to

1989 and the data for particular products or materials in later years in Tables 2.2.1 to 2.2.4, the

most important radionuclides distributed as exempt concentrations in regard to potential doses

from disposal appear to be 3

H, 14C, 60Co, and 85Kr. The doses from disposal of any other

radionuclides should be insignificant by comparison, because of their shorter half-lives and

lower total activities distributed.

Except for 60Co, the dose assessment for disposal is based on an assumption that the total

activities given in Table 2.2.6 represent the total distributions for a 20-year time period and that

each radionuclide is distributed uniformly over time. Therefore, the annual distributions are

assumed to be 1.8 TBq (4.8×107 Ci) for 3

H, 1.0 GBq (2.6×104 Ci) for 14C, and 0.1 TBq

(2.7×106 Ci) for 85Kr, and the same quantities of these radionuclides are assumed to be

disposed each year. Disposal of 3

H, 14C, and 85Kr in landfills and by incineration is assumed to

occur, with 80% of all disposals of these radionuclides going to landfills and 20% to

incinerators.

The dose assessment for disposal of 60Co is based on assumptions that differ somewhat from

those for 3

H, 14C, and 85Kr described above. First, the average annual distribution for 60Co of

0.3 TBq (8.2×106 Ci) obtained from the data in Table 2.2.6 is assumed to be inappropriate,

because nearly all of the reported distribution over the 20-year period occurred in a single year. 

In this assessment, the annual distribution of 60Co is assumed to be 30 GBq (8×105 Ci), or a

factor of 10 less, based on the reported distribution in 1 year by a large steel producer, provided

in Table 2.2.4. Second, radioactive decay between the time of distribution and the time of
2–29

disposal is taken into account for 60Co by assuming that disposal occurs at two half-lives

(i.e., about 10 years) after distribution. Third, disposal is assumed to occur in landfills but not in

incinerators, because the 60Co is assumed to be contained in steel forms that would not

normally be incinerated. Finally, because the 60Co is assumed to be contained in steel,

recycling also is considered as a disposal option.

2.2.4.5.1 Disposal in Landfills

Based on the generic methodology in Appendix A.2 and the assumptions described above, the

following estimates of individual and collective dose from disposal in landfills of the assumed

quantities of byproduct materials are obtained.

For 3

H, the annual EDE to individual waste collectors, individual landfill workers and other

members of the public would be less than 1×105

 mSv (<0.001 mrem). The collective EDE from

1 year’s disposals would be 4×104

 person-Sv (0.04 person-rem), due almost entirely to

exposure to off-site residents from releases to groundwater more than 1000 years after

disposal.

For 14C, the annual EDE to individual waste collectors would be 20 pSv (2 nrem) and the annual

doses to individual landfill workers or other members of the public would be considerably less. 

The collective EDE from 1 year’s disposals would be 1×105

 person-Sv (1×103 person-rem),

due almost entirely to exposure to off-site residents from releases to groundwater more than

1000 years after disposal.

For 60Co, the annual EDE to individual waste collectors would be 0.004 mSv (0.4 mrem), and

the annual doses to individual landfill workers or other members of the public would be

considerably less. The collective EDE from 1 year’s disposals would be 0.02 person-Sv

(2 person-rem), due almost entirely to exposure to waste collectors and workers at landfills. 

These dose estimates should be conservative for disposal of steel forms containing 60Co,

because the self-shielding provided by the steel is not taken into account. The dose estimates

also would be conservative if the materials containing 60Co were used significantly longer than

10 years prior to disposal.

For 85Kr, the annual EDE to individual waste collectors would be 4×105

 mSv (0.004 mrem), and

the annual doses to individual landfill workers or other members of the public would be

considerably less. The collective EDE from 1 year’s disposals would be 2×104

 person-Sv

(0.02 person-rem), due primarily to exposure to waste collectors and landfill workers.

Thus, in summary, based on data on the distribution of exempt concentrations of byproduct

material, the dose from disposal in landfills would be due almost entirely to the distribution of

60Co, and the following dose estimates are obtained:

The annual EDE to individuals, i.e., waste collectors, would be about 0.004 mSv

(0.4 mrem).

The collective EDE from 1 year’s disposals, which would be received almost entirely by

waste collectors and workers at landfills, would be about 0.02 person-Sv (2 person-rem).
2–30

2.2.4.5.2 Disposal in Incinerators

Based on the generic methodology in Appendix A.2 and the assumptions described at the

beginning of Section 2.2.4.5, including the assumption that steel containing 60Co would not be

incinerated, the following estimates are obtained of individual and collective dose from disposal

in incinerators of the assumed quantities of byproduct materials.

For 3H, the annual EDE to individual waste collectors, individual incinerator workers and other

members of the public would be less than 1×105

 mSv (<0.001 mrem). The collective EDE from

1 year’s disposals would be 5×105

 person-Sv (5×103 person-rem), due almost entirely to

exposures to off-site members of the public near waste incinerators.

For 14C, the annual EDE to individual waste collectors, individual incinerator workers and other

members of the public would be less than 1×105

 mSv (<0.001 mrem). The collective EDE from

1 year’s disposals would be 2×108

 person-Sv (2×106 person-rem), due almost entirely to

exposure to waste collectors.

For 85Kr, the annual EDE to individual waste collectors would be 2×104

 mSv (0.02 mrem), and

the annual doses to individual incinerator workers or other members of the public would be

considerably less. The collective EDE from 1 year’s disposals would be 3×105

 person-Sv

(3×103

 person-rem), due almost entirely to exposure to waste collectors.

Thus, in summary, based on data on the distribution of exempt concentrations of byproduct

material and the assumption that exempted materials containing 60Co would not be incinerated,

the dose from disposal in incinerators would be due almost entirely to the distribution of 3

H and

85Kr, and the following dose estimates are obtained:

The annual EDE to individuals, i.e., waste collectors, would be about 2×104

 mSv

(0.02 mrem).

The collective EDE from 1 year’s disposals, which would be received almost entirely by

waste collectors and off-site members of the public near incinerators, would be about

8×105

 person-Sv (0.008 person-rem).

2.2.4.5.3 Recycling of Contaminated Steel

If exempt concentrations of 60Co are assumed to be contained primarily in steel, recycling of the

60Co is a credible scenario for disposal. However, the concentrations of 60Co in recycled

materials generally would be less than the concentrations in the original materials, due to

dilution by mixing with uncontaminated materials, and the total activity of 60Co in the recycled

materials would be less than in the original materials at the time they are produced, due to

radioactive decay. In addition, it presumably is unlikely that all of the 60Co distributed under this

exemption would be recycled. Therefore, it is reasonable to conclude that individual and

collective doses from use of recycled materials containing 60Co would be substantially less than

the doses from use of the original materials prior to disposal, as discussed in Sections 2.2.4.1

and 2.2.4.3.
2–31

2.2.4.6 Accidents and Misuse

Potential doses from accidents and misuse involving products or materials containing exempt

concentrations of byproduct material are inherently limited by the definition of exempt

concentrations in terms of maximum permissive concentrations for occupational exposure,

which results in low values of the exempt concentrations. Doses from accidents and misuse

appear to be limited further, in practice, by the small volumes of the more common products or

materials containing relatively high concentrations of byproduct material. In this assessment,

the following scenarios considered illustrate the low doses that could result from accidents or

misuse involving products or materials containing exempt concentrations of byproduct material.

First, as described in Section 2.2.3.2 and summarized in Table 2.2.7, the EDE to an individual

during a fire at a facility for processing and assembly of irradiated silicon semiconductor

materials would be about 0.001 mSv (0.1 mrem) (NRC, Memoranda, Paperiello, 1994). This

dose estimate was based on reported concentrations of byproduct material in the irradiated

materials of about 10% or less of the exempt concentrations. However, the estimate should be

conservative because an entire year’s supply of semiconductor materials that would be

irradiated in a single reactor facility was assumed to be present in the processing facility during

the fire and all of the activity was assumed to be released into the air.

Second, misuse of a large ampule of tritiated water is considered, resulting in release of the

entire contents of the ampule into the air in a room the size of a small laboratory. If it is

assumed that 3

H at its exempt concentration of 1.1 kBq/mL (0.03 Ci/mL) is contained in a

50-mL ampule, the volume of the laboratory is 75 m3

, the air turnover rate in the room is 1/h,

and the breathing rate for an individual is 1.2 m3

/h (see Appendix A.1), the EDE from exposure

over the next 8 hours would be about 2×105

 mSv (0.002 mrem), based on the committed EDE

per unit activity intake in Table 2.1.2, which takes into account absorption of 3

H through the skin

as well as inhalation.

Third, the possibility is considered that the entire ampule of tritiated water described above

would be ingested inadvertently. For an assumed activity of 3

H in the ampule of 56 kBq

(1.5 Ci), and using the ingestion dose conversion factor in Table 2.1.2, the resulting EDE

would be about 1 Sv (0.1 mrem).

Finally, based on the generic methodology for accidents in Appendix A.1, doses to firefighters

and individuals cleaning up after a fire are considered. Doses to these individuals should be

considerably greater than doses to other members of the public who might be located near a

fire. For the byproduct materials for which dose estimates were obtained, the EDE per unit

activity available for any type of fire is always less than 2.7×106

 Sv/GBq (<108 rem/Ci) (see

Tables A.1.4 to A.1.6). Therefore, doses approaching 0.01 mSv (1 mrem) would be obtained

only if the total activity available during a fire were about 4 GBq (0.1 Ci) or greater. Since the

exempt concentrations of byproduct material in liquid or solid form are 1 kBq (0.03 Ci/g) or

less, the mass of material involved in a fire would need to be about 3 Mg or greater. Based on

available information about the types and quantities of the more common products or materials

containing exempt concentrations of byproduct material (e.g., as discussed in Section 2.2.2),

the required mass appears to greatly exceed the total mass of materials that are distributed

annually under this exemption. Therefore, it appears quite unlikely that individual doses

resulting from fires involving exempt concentrations of byproduct material could approach

0.01 mSv (1 mrem).
2–32

Based on the foregoing analysis, potential doses from accidents or misuse involving products or

materials containing exempt concentrations of byproduct material appear to be very low. 

Indeed, for current practices, it does not appear that a credible scenario for accidents or misuse

could result in EDEs exceeding about 0.001 mSv (0.1 mrem).

2.2.5 Effect of Changes in Dose Limits and Internal Dosimetry Models

As noted in Section 2.2.1, the exempt concentrations of byproduct material are derived from

limits on annual committed dose equivalents for internal exposure to workers of 0.05 Sv (5 rem)

to the whole body or the gonads, 0.3 Sv (30 rem) to the thyroid, or 0.15 Sv (15 rem) to any

other organ (NBS 69) and the dosimetric and metabolic models for inhalation and ingestion of

radionuclides in ICRP Publication 2 (ICRP, 1960). In 1991, 10 CFR 20 was revised (NRC,

1991) to limit occupational exposure to an annual effective dose equivalent of 0.05 Sv (5 rem)

and to utilize the dosimetric and metabolic models in ICRP Publication 30 (ICRP, 1979). In this

section, the effects on the exempt concentrations are investigated, if they were to be revised to

reflect these changes to 10 CFR 20. 

Assuming an internal dose of 0.05 Sv (5 rem) annually to persons using exempt concentrations

as the basis for the concentration values, the effect of the changes described above can be

investigated by comparing the existing exempt concentrations for liquids and solids in

Column II, of Schedule A, 10 CFR 30.70 with 100 times the limits on effluent concentrations in

water in Table 2, Column 2 of Appendix B of the revised 10 CFR 20 (NRC, 1991). The limits on

effluent concentrations in 10 CFR 20, which were calculated using the dosimetric and metabolic

models in ICRP Publication 30 (ICRP, 1979), are increased by a factor of 100 because they are

based on an annual committed EDE of 0.5 mSv (0.05 rem) for members of the public, rather

than the limit of 0.05 Sv (5 rem) for occupational exposure). This comparison is useful even

though the dose limit for occupational exposure in the revised 10 CFR 20 applies to the sum of

internal and external exposure, because the existing exempt concentrations and the limits on

effluent concentrations in 10 CFR 20 both are based only on considerations of internal

exposure.

In adopting MPCs for occupational exposure to define exempt concentrations of byproduct

material, the Atomic Energy Commission (AEC) reasoned that exempted products or materials

would not generally be inhaled or ingested and that continuous exposure over a year is highly

unlikely. Therefore, in the AEC’s judgement, it is highly improbable that any member of the

public exposed to byproduct material in concentrations less than the limits for exemption would

receive an annual dose equivalent in excess of a small fraction of 0.5 rem (5 mSv), which was

the existing dose criterion for limiting external exposure of members of the public (AEC, 1960). 

For the radionuclides listed in Table 2.2.6, a comparison of the existing exempt concentrations

with the values recalculated as described above is given in Table 2.2.12. These radionuclides

presumably have been distributed in the greatest amounts under this exemption. For most of

the radionuclides, the changes in the dose limit for workers and the internal dosimetry models

would increase the exempt concentration, but the increase is less than an order of magnitude in

most cases. For a few radionuclides (i.e., 14C and 60Co), the recalculated exempt concentration

is less than the existing value, but the decrease is less than a factor of three.
2–33

2.2.6 Summary

Exempt concentrations are specified for a large number of byproduct materials and, except for

the provision that exempted products or materials should not be readily inhaled or ingested, or

designed for application to the human body, there are no restrictions on the types of products or

materials into which exempt concentrations of byproduct material can be incorporated. Also,

the NRC may grant exceptions to this provision, and it has in the case of gemstones. There are

no restrictions on the total volume or mass of exempted products or materials and, thus, the

total activity of byproduct materials, although there is a provision to show that lower

concentrations are not feasible for the particular application. Therefore, a rigorous assessment

of individual doses from routine uses of products or materials containing exempt concentrations

of byproduct material is a difficult undertaking. It also is difficult to estimate the collective dose

from all routine uses of such products or materials, because the available information on the

total activities of byproduct materials distributed under this exemption and the extent of various

individual practices is incomplete. In the case of exempt quantities of byproduct material which

can be distributed either by an NRC license or an Agreement State license, it was assumed that

twice the quantity was distributed by Agreement State licensees as NRC licensees.

However, in spite of the difficulties in estimating individual and collective doses, it appears that

the doses associated with this exemption are low. Important factors in limiting doses noted in

this assessment include the low values of the exempt concentrations for photon-emitting

radionuclides, the small source volumes and, thus, the low total activities of radionuclides in

products or materials containing concentrations of byproduct material approaching the

maximum exempt concentrations, and the very low reported concentrations of byproduct

material in products or materials with larger source volumes. These factors inherently limit

doses from external and internal exposure. An additional important factor in limiting doses is

the requirement that the intended use of any product or material distributed under this

exemption and the feasibility of using lower concentrations of byproduct materials must undergo

regulatory review. The licensing review prior to distribution of exempted products or materials

and the requirements for licensees who introduce byproduct material in exempt concentrations

into products or materials, as specified in 10 CFR 32.11 and similar Agreement State

regulations, have served to limit plausible exposure conditions. As a consequence, in the many

years of experience with this exemption, no approved practices have produced doses near any

theoretical limit, such as that discussed in Section 2.2.3.3 for exposure to 60Co in contaminated

steel.

In this assessment, estimates of individual and collective dose to the public from routine use,

distribution and transport, and disposal of products or materials containing exempt

concentrations of byproduct material were obtained based on available information on the types

of products or materials that are most commonly distributed under this exemption at the present

time and the total quantities of byproduct material that have been distributed under this

exemption. Doses from accidents and misuse involving exempted products or materials also

were considered. The results of this assessment are summarized in Table 2.2.13.
2–34

Based on this assessment, the following general conclusions about radiological impacts on the

public associated with this exemption can be obtained:

Maximum credible external doses during routine use appear to be considerably higher

than maximum credible internal doses. This result is due in large part to the presence of

photon-emitting radionuclides in many of the most commonly distributed products or

materials. Even if byproduct materials were routinely released into the air or ingested

(e.g., during lapping and cutting of irradiated silicon semiconductor materials), the

maximum credible internal dose should be considerably less than the maximum credible

external dose.

There do not appear to be any credible scenarios for accidents or misuse involving

products or materials containing exempt concentrations of byproduct material that could

result in doses substantially higher than the estimates of individual dose during routine

use. Doses from accidents or misuse, which generally would involve inhalation or

ingestion exposure, are inherently limited by the low values of the exempt

concentrations and, thus, the low activities of byproduct material that could be inhaled or

ingested in any plausible scenario.

The upper bound estimates of individual dose from distribution and transport and disposal of

exempt concentrations of byproduct material given in Table 2.2.13 are comparable to or greater

than the best estimate of individual dose from wearing of irradiated topaz gemstones, which is a

common exposure scenario for this exemption. In practice, however, the individual doses from

distribution and transport and disposal probably are less than the individual doses during

routine use, because the former are based on important assumptions that are likely to be

conservative. In particular, the estimated individual dose from distribution and transport

assumes that a single truck driver would be exposed to half of the total number of irradiated

topaz gemstones distributed in a year, and the estimated individual dose from disposal, which is

due almost entirely to disposal of 60Co contained in steel, does not take into account the selfshielding provided by the source. These kinds of conservative assumptions are not

incorporated in the individual dose assessment for routine wearing of irradiated topaz

gemstones.
2–35

Table 2.2.1 Quantities of Byproduct Material Transferred in Irradiated Topaz Gemstones

by Single Materials Licensee a

Radionuclide Half-Life

Activity Transferredb

(Ci)

Concentrationc

(Ci/g)

Exempt

Concentrationd

(Ci/g)

46Sc 83.83 days 0.9 (0.9-7)×105 4×104

54Mn 312.7 days 0.7 (2-5)×105 1×103

65Zn 244.4 days 0.2 5×105 1×103

134Cs 2.062 yr 0.03 (1-2)×105 9×105

182Ta 114.74 days 7.5 (0.6-2)×104 4×104

a

 Data reported by Brightwell (1994) for the period June 20, 1991, through January 26, 1993.

b

 Total activity transferred in all irradiated gemstones; 1 Ci = 37 kBq. c

 Range of activity concentrations in irradiated gemstones.

d

 Value from Schedule A of 10 CFR 30.70.
2–36

Table 2.2.2 Quantities of Byproduct Material Transferred in Irradiated Semiconductor

Materials by Single Materials Licensee a

Radionuclide Half-Life

Concentrationb

(Ci/g)

Exempt

Concentrationc

(Ci/g)

24Na 15.00 h <2×1018 2×103

31Si 157.3 min <2×109 9×103

32P 14.29 days (0.003-2)×106 2×104

47Sc 3.422 days (0.005-7)×105 9×104

51Cr 27.704 days 1.0×104 2×102

58Co 70.80 days (0.9-2)×105 1×103

64Cu 12.701 h <2×1020 3×103

65Zn 244.4 days (2-4)×105 1×103

76As 26.32 h <2×1012 2×104

82Br 35.30 h <7×1011 3×103

122Sb 2.70 days (2-5)×108 3×104

124Sb 60.20 days (0.9-1)×105 2×104

198Au 2.696 days (0.8-1)×108 5×104

a

 Data reported by Borza (1995) for the period August 31, 1990, through June 30, 1995.

b

 Range of activity concentrations in irradiated semiconductor materials; 1 Ci = 37 kBq. Total

activity transferred by the licensee cannot be determined from the information provided.

c

 Value from Schedule A of 10 CFR 30.70.
2–37

Table 2.2.3 Quantities of Byproduct Material Introduced Into High-Purity Silicon

Semiconductor Materials by Single Materials Licensee a

Radionuclide Half-Life

Concentrationb

(Ci/g) 

Exempt

Concentrationc

(Ci/g) 

24Na 14.96 h 6×105 2×103

42K 12.36 h 2×105 3×103

45Ca 165 days 3×105 9×105

47Ca 4.54 days 1×106 5×104

51Cr 27.7 days 2×106 2×102

59Fe 44.5 days 4×107 6×104

60Co 5.27 yr 6×105 2×103

64Cu 12.7 h 6×106 3×103

65Zn 244 days 3×107 1×103

76As 26.3 h 9×106 2×104

75Se 120 days 3×107 3×103

95Zr 64.0 days 6×105 2×103

82Br 35.3 h 2×106 3×103

97Zr 16.74 h 9×107 2×104

99Mo 65.94 h 6×107 2×103

110mAg 250 days 3×107 3×104

115Cd 55.46 h 2×106 3×104

122Sb 2.70 days 8×106 3×104

124Sb 60.2 days 3×107 2×104

134Cs 2.06 yr 2×107 9×105

140La 40.27 h 1×106 2×104

141Ce 32.5 days 5×107 9×104

147Nd 10.98 days 2×108 6×104

152mEu 9.32 h 4×105 6×104

See following page for footnotes.
2–38

Table 2.2.3 Quantities of Byproduct Material Introduced Into High-Purity Silicon

Semiconductor Materials by Single Materials Licensee a

 (continued)

Radionuclide Half-Life

Concentrationb

(Ci/g) 

Exempt

Concentrationc

(Ci/g) 

152Eu 13.33 yr 7×107 7×104

159Gd 18.56 h 1×106 8×104

160Tb 72.3 days 7×107 4×104

169Yb 32.0 days 3×106 6×104

175Yb 4.19 days 1×104 1×103

177Lu 6.71 days 1×104 1×103

181Hf 42.39 days 2×107 7×104

186Re 90.64 h 6×105 9×104

187W 23.9 h 6×106 7×104

188Re 16.98 h 3×105 6×104

192Ir 73.83 days 3×107 4×104

194Ir 19.15 h 7×107 3×104

197Hg 64.1 h 5×105 3×103

198Au 2.70 days 2×105 5×104

203Hg 46.61 days 2×106 2×104

239Np 2.36 days 2×105 2×103

a

 Data reported by Morris (1993).

b

 Values are upper limits based on maximum concentrations or detection limits of impurity

elements in high-purity silicon, and are based on an assumed 27-hour irradiation at a thermal

neutron flux of 1×1013/cm2

-s and a decay time of approximately 56 hours; 1 Ci = 37 kBq. c

 Value from Schedule A of 10 CFR 30.70. If the exempt concentration is not listed for a

radionuclide in Schedule A of 10 CFR 30.70, the value was calculated by the licensee based on

the annual limit of intake for ingestion in Table 1, Column 1 of Appendix B to 10 CFR 20.1001

to 20.2401 (NRC, 56 FR 23360) and assumed daily water intake of 3,000 g for 365 days/yr.
2–39

Table 2.2.4 Quantities of Byproduct Material Introduced Into Engine Oil or Steel

by Materials Licensees

Radionuclide/Product Half-Life

Activity

Introduceda Concentrationb

Exempt

Concentrationc

3

H/Engine oild 12.28 yr 3.2 Ci (0.04-2)×102 Ci/mL 3×102 Ci/mL

60Co/Steele 5.271 yr 0.79 Ci (3-5)×107 Ci/g 5×104 Ci/g

a

 Total activity introduced into all products; 1 Ci = 37 kBq and 1 Ci = 37 GBq.

b

 Range of activity concentrations in products.

c

 Value from Schedule A of 10 CFR 30.70.

d

 Data reported by Hamelink (1990) for the period June 30, 1987, through June 30, 1990.

e

 Data reported by Kobrick (1991) for the period through December 31, 1990.
2–40

Table 2.2.5 Estimated Concentrations of Byproduct Material Introduced Into Baggage

Contents From Neutron Irradiation of 1-kg Masses of Various Elements

in Airport Explosive Detection System a

Radionuclide Half-Life

Concentration

(Ci/g)

Exempt

Concentrationb

(Ci/g)

24Na 15.00 h 4×106 2×103

27Mg 9.458 min 1×106 —

28Al 2.240 min 5×104 —

38Cl 37.21 min 1×105 4×103

46mSc 18.72 s 3×102 —

52V 3.75 min 4×103 —

56Mn 2.5785 h 2×104 1×103

66Cu 5.10 min 3×104 —

69Zn 55.6 min 8×106 2×102

77mSe 17.4 s 4×103 —

80Br 17.4 min 6×104 —

104mRh 4.36 min 7×103 —

108Ag 2.37 min 1×102 —

110Ag 24.57 s 8×102 —

116mIn 54.15 min 3×103 —

152mEu 9.32 h 3×103 c 6×104

165mDy 1.26 min 3×101 —

187W 23.83 h 8×106 7×104

198Au 2.696 days 3×105 5×104

a

 Concentrations calculated by Randolph and Simpson (1988), based on the description of the

explosive detection system by Science Applications International Corporation (SAIC, 1988);

1 Ci = 37 kBq.

b

 Value from Schedule A of 10 CFR 30.70. Blank entry indicates that an exempt concentration

has not been established other than the broad provision in Schedule A for beta and/or gamma

emitting byproduct material with a half-life less than 3 years.

c

 Assumed concentration exceeds the exempt concentration.
2–41

Table 2.2.6 Quantities of Radionuclides Distributed as Exempt Concentrations

of Byproduct Material During 1970 to 1989 a

Radionuclide Half-Lifeb

Activity

(Ci)c Radionuclide Half-Lifeb

Activity

(Ci)c

110mAg 250 days 0.37 55Fe 2.7 yr 0.007

198Au 2.7 days 0.21 59Fe 44.6 days 6.48

82Br 35.3 s 0.025 3

H 12.28 yr 963

14C 5730 yr 0.51 203Hg 46.6 days 1.20

45Ca 163 days 0.010 131I 8.04 days 4.71

57Co 271 days 0.31 85Kr 10.72 yr 53.0

58Co 70.8 days 1.06 24Na 15.00 h 0.018

60Co 5.271 yr 164 32P 14.29 days 2.88

51Cr 27.7 days 0.015 46Sc 83.83 days 0.002

134Cs 2.06 yr 0.15 113Sn 115.1 days 0.008

a

 Data summarized from unpublished NRC report, M. L. Janney, 1990 (see references). For all

other radionuclides for which exempt concentrations have been established in Schedule A of

10 CFR 30.70, the quantity distributed was less than 0.01 mCi (<0.37 MBq).

b

 Values obtained from Kocher (1981).

c

 1 Ci = 0.037 TBq or 37 GBq.
2–42

Table 2.2.7 Estimates of Dose From Processing and Assembly of Irradiated Silicon

Semiconductor Materials a

Exposure Scenario Individual Effective Dose Equivalent

Exposure to workers during processing

 and assembly of irradiated materials

 into electronic components

0.8 mrem/yr for wet processing of materialsb

1.0 mrem/yr for dry processing of materialsc

Exposure to members of the public

 resulting from disposal of fines from

 processing of materials into sanitary

 sewer systemsd

0.2 rem/yr

Exposure to members of the public

 during fire at facility for the processing

 of materials

0.1 mrem

a

 Doses estimated by Paperiello (NRC, Memoranda, 1994) for irradiated materials assumed to

contain concentrations of impurity byproduct materials reported by materials licensee;

1 mrem = 0.01 mSv; 1rem = 0.01 Sv. Assumed concentrations for radionuclides contributing

significantly to dose were about 10% or less of corresponding exempt concentrations. Dose in

all scenarios was assumed to result primarily from exposure to 152Eu.

b

 Estimated dose results almost entirely from external exposure; estimated dose from internal

exposure is 0.0001 mSv/yr (0.01 mrem/yr).

c

 About two-thirds of the estimated dose results from external exposure and about one-third

from internal exposure.

d

 Individuals receiving the highest doses were assumed to be sewer sludge operators at waste

water treatment facility or equipment operators at landfill for disposal of sewage sludge.
2–43

Table 2.2.8 Estimates of External Dose Rates Near Steel Slab Contaminated With 60Co a

Location

Dose-Equivalent Rateb

(rem/h per pCi/g) 

Contact 2.3-6.3

Distance of 1 footc 1.3-2.1

a

 Dose rates reported by Leoben (NRC, Memoranda, 1996) for a uniformly contaminated

steel slab of dimensions 1.3 m × 1.3 m × 0.66 m; 1 rem/h per pCi/g = 0.27 nSv/h per mBq/g.

b

 Range in estimated dose rates is based on differences among various

calculations and measurements.

c

 Corresponds to a distance of about 0.3 meters.
2–44

Table 2.2.9 Estimates of External Dose Rates From Exposure to Various Radionuclides

in Baggage Contents Irradiated by Neutrons in Airport Explosive Detection System a

Dose-Equivalent Ratec

(mrem/h)

Radionuclide 

Concentrationb

(Ci/g) 30 s After Irradiation 1 h After Irradiation

24Na 4×106 1×104 9×105

27Mg 1×106 8×106 1×107

28Al 5×104 5×103 6×1011

38Cl 1×105 1×104 3×105

46mSc 3×102 3×102 —

52V 4×103 3×102 5×107

56Mn 2×104 3×103 2×103

66Cu 3×104 2×104 5×108

69Zn 8×106 2×108 7×109

77mSe 4×103 2×103 —

80Br 6×104 3×104 3×105

104mRh 7×103 2×103 2×107

108Ag 1×102 1×103 3×1011

110Ag 8×102 2×102 —

116mIn 3×103 4×102 2×102

152mEu 3×103 5×103 5×103

165mDy 3×101 3×102 2×1016

187W 8×106 2×105 2×105

198Au 3×105 6×105 6×105

a

 Doses estimated by Randolph and Simpson (1988).

b

 Estimated concentrations from irradiation of 1-kg masses of various stable elements given in

Table 2.2.5; 1 Ci = 37 kBq. c Dose rates at a distance of 30 cm from source; 1 mrem = 0.01 mSv.
2–45

Table 2.2.10 Estimates of Dose From Exposure to Baggage Contents Irradiated

by Neutrons in Airport Explosive Detection System a

Exposure Scenario Radionuclideb Annual Dose Equivalentc

External exposure to baggage handler for

 2,000 h/yr at 30 s after irradiation

28Al 10 mrem

External exposure to passenger for 1 hour

 beginning at 1 hour after irradiation

56Mn 2 rem;

2.1 person-remd

External exposure to passenger during a

 3-hour car trip and placement of bag near

 individual for the next 12 hours

152mEue 0.8 mrem

External exposure from wearing of 40-g

 gold medallion continuously for 10 days

 after irradiation

198Au 0.7 mrem;f

40 person-remf, g

External exposure from application of

 cosmetics 1 hour after irradiation

56Mnh 3 remf, i

Internal exposure from consumption of

 1 day’s food supply 1 hour after irradiation

24Naj 0.03 rem;k

0.0003 person-remk, l

a

 Doses estimated by Randolph and Simpson (1988).

b

 Unless otherwise noted, assumed concentration of radionuclide was value resulting from

irradiation of 1 kg of stable element given in Table 2.2.5.

c

 Unless otherwise noted, value is effective dose equivalent (EDE) to individuals from photon

exposure. 1 rem = 0.01 Sv; 1 mrem = 0.01 mSv; 1 rem = 0.01 Sv.

d

 Collective dose for exposure to 1.1 million passengers per year.

e

 Assumed concentration of radionuclide was value that would give a dose rate of 0.005 mSv/h

(0.5 mrem/h) at surface of luggage, which is release criterion for explosive detection system. 

Assumed concentration is two orders of magnitude greater than the value that would result from

irradiation of 1 kg of Eu given in Table 2.2.5 and, thus, is nearly three orders of magnitude

greater than the exempt concentration.

f

 Dose equivalent to the whole skin from beta exposure.

g

 Collective dose based on the assumption that 0.5% of 1.1 million passengers per year carry

gold medallions in luggage.

h

 Assumed concentration was based on the reported amount of Mn in common cosmetic

materials.

i

 Dose per gram of cosmetics applied. Estimated doses from other radionuclides that could

occur in common cosmetic materials are at least two orders of magnitude lower.

j

 Radionuclide contributes more than 75% of the estimated dose from the consumption of

irradiated food. Assumed concentrations of all radionuclides in irradiated food were based on

the reported amounts of stable elements in normal daily diet.

k

 Committed EDE from ingestion.

l

 Collective dose based on the assumption that 1% of 1.1 million passengers per year carry salt

tablets or highly salted food in luggage.
2–46

Table 2.2.11 Estimates of External Dose While Wearing Irradiated Topaz Gemstones

Containing Exempt Concentrations of Byproduct Material 

Photon Exposure to

Whole Body

Beta Exposure to

Skin

Radionuclide

Exempt

Concentrationa

(Ci/g) 

Individual

Annual

Effective

Dose

Equivalentb

(mrem)

Collective

Dose

Equivalentc

(person-rem)

Individual Annual

Dosed

(mrem)

46Sc 4×104 0.2 12 10

54Mn 1×103 0.4 8 0

65Zn 1×103 0.3 5 1

134Cs 9×105 0.09 8 10

182Ta 4×104 0.1 30 30

Total 60

a

 Value from Schedule A of 10 CFR 30.70; 1 Ci/g = 37 kBq/g.

b

 Effective dose in the first year (ICRP 60) while wearing a 30-carat (6-g) gemstone containing

exempt concentration for 8 h/day and 365 days/yr. Dose to individual in first year for a

gemstone with average concentrations from Table 2.2.1 is 0.03 mrem. 1 mrem = 0.01 mSv.

c

 Collective effective dose for a population of 2.25 million, each wearing a single 5-carat (1-g)

gemstone for 8 h/day and 365 days/yr with the average concentrations from Table 2.2.1. 

1 person-rem = 0.01 person-Sv.

d

 Dose equivalent to the irradiated portion of skin while wearing 30-carat (6-g) gemstones

containing exempt concentration for 8 h/day and 365 days/yr. Average dose to the whole skin

is obtained by dividing 10 cm2

 exposed area by the area of whole skin of 18,000 cm2. 

1 mrem = 0.01 mSv.
2–47

Table 2.2.12 Comparison of Existing Exempt Concentrations Calculated to Result in

5 Rem Internal Annual Dose Based on Revised 10 CFR 20

Radionuclidea

Exempt Concentration

(Ci/g)b

Dose Rate

Relative to 60Co

For Existing

Concentrations Existing e c Recalculatedd

110mAg 3×104 6×104 0.7

198Au 5×104 2×103 0.2

82Br 3×103 4×103 7

14C 8×103 3×103

45Ca 9×105 2×103

57Co 5×103 6×103 1

58Co 1×103 2×103 0.9

60Co 5×104 3×104 1

51Cr 2×102 5×102 0.7

134Cs 9×105 9×105 0.1

55Fe 8×103 1×102

59Fe 6×104 1×103 0.6

3

H 3×102 1×101

203Hg 2×104 3×103 0.07

131I 2×105 1×104 0.01

85Krf 3×106 7×105 ~0

24Na 2×103 5×103 5

32P 2×104 9×104

46Sc 4×104 1×103 0.7

113Sn 9×104 3×103 0.5

a

 Radionuclides listed are those in Table 2.2.6 and are expected to be the most important

radionuclides distributed as exempt concentrations.

b

 1 Ci = 37 kBq. c

 Value listed in Schedule A of 10 CFR 30.70 for materials in liquid or solid form, except as

noted.

d

 Value is 100 times the limit on effluent concentration in water in Table 2, Column 2 of

Appendix B of revised 10 CFR 20 (NRC, 56 FR 23360), except as noted (see Section 2.2.5).

e

 The dose rate relative to 60Co for each radionuclide is calculated as the ratio of the exempt

concentration times its gamma-ray dose constant (from Table 2.1.2) to that for 60Co. The

relative dose rate values will vary for different materials, configurations and shielding.

f

 Exempt concentration and recalculated value are for gaseous form in units of Ci/mL.
2–48

Table 2.2.13 Summary of Potential External Radiation Doses From Use of Products or

Materials Containing Exempt Concentrations of Byproduct Material a

Exposure Scenario

Individual Annual Effective

Dose Equivalent

(mrem)

Collective Effective

Dose Equivalentb

(person-rem)

Routine use <1c

 0.03e

90d

Distribution and transport <0.5f 1g

Disposal

 Landfills

 Incinerators

 Recyclingj

<0.4h

0.02h

 2i

 0.008i

Accidents and misuse 0.1k

a

 External dose estimates are intended to represent credible upper bounds for products or

materials most commonly distributed under exemption at present time, and are due to external

exposure, except as noted; 1 mrem = 0.01 mSv; 1 rem = 0.01 Sv.

b Refer to text for discussion of time period for collective dose calculations.

c

 Upper bound estimate of dose to workers during wet or dry processing and assembly of

irradiated silicon semiconductor materials into electronic components, due primarily to external

exposure, or to individuals located near contaminated steel containing 60Co. Estimate is based

on the reported concentrations of byproduct material in particular products or materials, rather

than the maximum exempt concentrations, but upper bound should be conservative (see

Section 2.2.4.1).

d

 Dose for 2.25 million irradiated topaz gemstones distributed in a year integrated over 10 years

of use, each gemstone containing average reported concentrations of photon-emitting

radionuclides given in Table 2.2.1. Other known uses of products or materials containing

exempt concentrations of byproduct material should increase collective dose by no more than

factor of 2 (see Section 2.2.4.3).

e

 Dose to individuals wearing large irradiated topaz gemstone containing maximum reported

concentrations of photon-emitting radionuclides given in Table 2.2.1. 

f

 Upper bound estimate of dose applies to individual truck driver who is assumed to transport

half of 2.25 million irradiated topaz gemstones distributed per year (see Section 2.2.4.4).

g

 Dose from annual distribution of 2.25 million irradiated topaz gemstones. Collective dose from

distribution and transport of other products or materials containing exempt concentrations of

byproduct material should be considerably less (see Section 2.2.4.4).

h

 Dose to waste collectors, based on the assumed annual disposals of byproduct materials in

landfills or incinerators (see Section 2.2.4.5).

i

 Dose from assumed annual disposals of byproduct materials in landfills or incinerators (see

Section 2.2.4.5).

j

 Recycling of 60Co contained in steel is a credible scenario, but doses from exposure to

recycled materials would be substantially less than doses from exposure to contaminated

materials prior to recycling (see Section 2.2.4.5.3).

k

 Dose for single occurrence of accident or misuse. Estimate applies to inadvertent ingestion of

50-mL ampule of water containing exempt concentration of 3

H. Dose estimates for other

credible scenarios for accidents and misuse are considerably less (see Section 2.2.4.6).
2–49

2.3 Timepieces, Hands, and Dials

2.3.1 Introduction

In 10 CFR Part 30.15(a)(1), timepieces (i.e., watches and clocks) or hands or dials containing

tritium (3

H) or 147Pm are exempted from licensing requirements for byproduct material, provided

that the following quantities of radioactivity or radiation levels are not exceeded:

(1) 930 megabecquerel (MBq) (25 millicurie (mCi)) of 3

H per timepiece;

(2) 190 MBq (5 mCi) of 3

H per hand;

(3) 560 MBq (15 mCi) of 3

H per dial, including bezels;

(4) 3.7 MBq (100 Ci) of 147Pm per watch or 7.4 MBq (200 Ci) of 147Pm per any

other timepiece;

(5) 0.74 MBq (20 Ci) of 147Pm per watch hand or 1.5 MBq (40 Ci) of 147Pm per

other timepiece hand;

(6) 2 MBq (60 Ci) of 147Pm per watch dial or 4 MBq (120 Ci) of 147Pm per other

timepiece dial, including bezels; and

(7) absorbed dose rates from hands and dials containing 147Pm, when measured

through 50 mg/cm2

 of absorber, shall not exceed:

(ii) 1 microgray (Gy)/h (0.1 mrad/h) at 10 cm from any surface of a

wristwatch,

(ii) 1 Gy/h (0.1 mrad/h) at 1 cm from any surface of a pocket watch, and

(iii) 2 Gy/h (0.2 mrad/h) at 10 cm from any surface of any other timepiece.

This exemption is separate from the class exemption in 10 CFR 30.19 for self-luminous

products, which is discussed in Section 2.14 of this report. Prior to 1998, only 3

H in the form of

paint and 147Pm in the form of paint have been used on timepieces, hands, and dials under

10 CFR 30.15(a)(1). In response to a petition for rulemaking to allow the use of 3

H gas in

sealed glass tubes in timepieces under this exemption (Keating, 1993), the Nuclear Regulatory

Commission amended its regulations (NRC, 63 FR 32969). Therefore, this new use is

evaluated.

The exemption for 3

H in timepieces was proposed on July 2, 1960 (25 FR 6302), and issued as

a final rule on December 13, 1960 (25 FR 12730). This exemption originally included a

requirement that the 3

H be bound in a non-water-soluble and non-labile form,” but this

provision was deleted when 10 CFR Part 32, which specifies requirements for manufacture of

certain items containing byproduct material, was first issued on June 26, 1965 (30 FR 8192).

The exemption for 147Pm in timepieces was proposed on May 20, 1964 (29 FR 6562), and was

issued as a final rule on October 6, 1967 (32 FR 13920). Some of the limits on activity or

radiation level in the final rule are lower than the values originally proposed, in order to conform

to international standards for radioluminous timepieces that had been developed.

2.3.2 Description of Items

Tritium and 147Pm are incorporated in timepieces in a polymer paint that contains a phosphor

(e.g., ZnS) and is applied to hands, dials, and bezels of wristwatches, pocket watches, and

alarm clocks. The 3

H becomes part of the paint and the 147Pm is mixed into the paint either as a
2–50

highly insoluble oxide or in ceramic microspheres. Also, it is possible to seal 3

H gas in

glass-like tubes that are coated with a phosphor. Beta-particle emission by either radionuclide

excites the phosphor as the particles are stopped in the paint. Visible light is produced by

scintillation of the phosphor crystals.

The useful life of a timepiece depends on many factors, including the length of time over which

the hands and dials of the timepiece remain visible. Factors affecting luminosity of the

3

H-bearing paints are the radioactive half-life of the 3H, the stability half-life of the ZnS

phosphor, and the 3

H release half-life. The International Atomic Energy Agency (IAEA) (IAEA,

SS 23) suggests that timepieces containing tritiated paints have a useful life of 10 years. 

Moghissi et al. (NUREG/CP–0001) contend that a 10-year useful life is too long but offer no

specific value of their own. Lacking definitive values for useful lives of timepieces, this

assessment assumes a useful life of ten years.

Potential health hazards associated with use of 3

H and 147Pm in timepieces are associated with

radiations emanating from these materials and with material that might escape from the

timepieces. The weak beta particles emitted by 3

H will be absorbed completely in a timepiece,

but 3

H will escape from paints and emanate from timepieces because of exchange with

atmospheric hydrogen or because of radiolytic decomposition of the paint components. Escape

of 147Pm from paints and its subsequent emanation from timepieces is unlikely, unless a

timepiece is damaged in a way that affects the integrity and containment of the paint. However,

the beta particles emitted by 147Pm, though not able to penetrate timepiece casings, are

sufficiently energetic to produce, when stopped in the timepiece components, bremsstrahlung

that will penetrate the casings.

The quantities of 3

H or 147Pm applied to timepieces, hands, and dials vary significantly,

depending on the design of a particular item. McDowell-Boyer and O’Donnell

(NUREG/CR–0216), using available licensing data from the 1969 to 1976 period, estimated the

average wristwatch to contain 74 MBq (2 mCi) of 3

H or 1.7 MBq (45 Ci) of 147Pm, the average

pocket watch to contain 19 MBq (0.5 mCi) of 3

H, and the average clock to contain 19 MBq

(0.5 mCi) of 3

H or 1.7 MBq (45 Ci) of 147Pm. They also estimated an annual distribution of

8.4 million timepieces that contain 3

H and 2 million that contain 147Pm. Timepieces containing

3

H consisted of 6 million (71%) wristwatches, 1.8 million (22%) clocks, and 0.6 million (7%)

pocket watches. Timepieces containing 147Pm consisted of 1 million (50%) wristwatches, 1

million (50%) clocks, and no pocket watches.

Between 1970 and 1989, available licensing data (Nuclear Regulatory Commission (NRC)

Unpublished Reports, Janney, 1990) indicate that approximately 1.4×107

 GBq (3.9×108 mCi) of

3

H and 1.2×106 GBq (3.3×107 mCi) of 147Pm were incorporated into timepieces. Thus, the

average annual distributions of 3

H and 147Pm were approximately 7×105 GBq (1.9×107 mCi) and

6×104

 GBq (1.6×106 mCi), respectively. The available data do not allow an estimate of the

number of timepieces, dials, and hands distributed or of the average 3

H or 147Pm content of the

items.

More recent, but only partial, licensing data indicate that at least 5.6×105

 GBq (1.5×107 mCi) of

3

H and 2.3×103 GBq (6.1×104 mCi) of 147Pm were distributed in timepieces, hands, and dials

during the 1989-1993 time period. Thus, the average annual distributions of 3

H and 147Pm were

approximately 1.1×105

 GBq (3.0×106 mCi) and 4.5×102 GBq (1.2×104 mCi), respectively. 

These data also indicate that at least 8 million timepieces (1.6 million/yr) containing 3

H and less
2–51

than 1 million timepieces (<0.2 million/yr) containing 147Pm were distributed during the period. 

An accurate breakdown by type of timepiece is unavailable, but the data do indicate that the 3

H

content of a timepiece ranges between 56 and 630 MBq (1.5 and 17 mCi), with an average

content of 63 MBq (1.7 mCi), and that the average 147Pm content is approximately 3.7 MBq

(100 Ci).

The available data indicate a decrease in annual distributions of timepieces containing 3

H or

147Pm from the levels observed by McDowell-Boyer and O’Donnell (NUREG/CR–0216), from

about 8.4 million to 1.6 million timepieces containing 3

H and from about 2 million to 0.2 million

containing 147Pm. Little change is apparent in the average radionuclide contents of the

timepieces, which are well below the exemption limit for timepieces containing 3

H and near the

limit for timepieces containing 147Pm.

2.3.3 Summary of Previous Assessments

2.3.3.1 Timepieces Containing 3

H

Many assessments have been conducted of the potential radiological impacts on the public

from timepieces containing 3

H paint. These assessments include those performed to justify the

exemption for such timepieces and those performed to evaluate the consequences of the

exemption. Fairly comprehensive assessments have been performed by McDowell-Boyer and

O’Donnell (NUREG/CR–0216) and Buckley et al. (NUREG/CR–1775), while Moghissi et al.

(NUREG/CP–0001) cite a host of assessments based on measured 3

H levels in people,

primarily workers who apply 3

H paint, who had come into contact with timepieces.

In the first Federal Register notice from 1960 cited above, the Atomic Energy Commission

(AEC) concluded that, under the conditions of the proposed rule, the exempt use of timepieces

containing 3

H would be safe and would be expected to result in a reduction of radiation

exposure to the population. (Tritium would replace radium in timepieces.) The determination of

safety is based on the belief that the quantities of 3

H in timepieces would not present an undue

hazard to the user or other members of the public. The basis for this conclusion is summarized

as follows:

Since 3

H-activated phosphors would be contained in an insoluble paint that is firmly

bound to the face of a timepiece, release rates of 3

H from timepieces should be low. 

However, quantitative estimates of releases were not given.

Levels of external radiation from 3

H in timepieces would be negligible, because the

maximum range of beta particles emitted in 3

H decay is much less than the thickness of

a watch crystal or the insensitive layer of the skin and most of the low-energy

bremsstrahlung produced by the stopping of beta particles within the watch case or

paint would be absorbed in that medium.

If 3

H in insoluble or soluble form were inhaled during normal handling or as a result of an

accident or fire, a substantial fraction of the allowable inventory of 3

H in a timepiece of

930 MBq (25 mCi) would have to be inhaled by an individual for the resulting doses to

the lungs or whole body to exceed existing limits for radiation workers. Furthermore,
2–52

inhalation of such large quantities of 3

H from a single timepiece, either during normal

handling or an accident, is extremely unlikely.

If 2 million timepieces per year were sold, each timepiece contained the maximum

exempt quantity of 3

H of 930 MBq (25 mCi), and all the 3H were released to the

environment yearly, the addition of 3

H to the environment would be only 2% of the

natural production rate by cosmic rays and the resulting annual dose equivalent to

average individuals would be less than two ten-millionths of the total annual dose

equivalent from all natural background radiation of 1.5 millisieverts (mSv) (150 mrem).

McDowell-Boyer and O’Donnell (NUREG/CR–0216) performed a systematic assessment of the

distribution, use, maintenance, and disposal of timepieces and of accidents that might involve

timepieces. Buckley et al. (NUREG/CR–1775) added to the work of McDowell-Boyer and

O’Donnell. Basically, these studies indicated individual whole-body dose equivalents on the

order of 0.01 mSv/yr (1 mrem/yr) to persons involved in all phases of timepiece life. Accidental

exposures opened the possibility that a few people could receive dose equivalents as high as

0.50 mSv (50 mrem). Collective dose equivalents were estimated to be in the 20 person-Sv/yr

(2000 person-rem/yr) range.

McDowell-Boyer and O’Donnell (NUREG/CR–0216) also attempted to quantify rates at which

3

H escapes from timepieces containing tritiated paints. They cite several studies that found 3H

evolution rates to be between 0.037 and 14 Bq/min (1 and 370 pCi/min) and to average about

1.1 Bq/min (30 pCi/min, or approximately 1 ppm/h). The work of McMillan (NUREG/CP–0001)

provided evidence that the leak rate of 3 ppm/h for 3

H emanating from timepieces containing 3H

paints is in the range of 1-3 ppm/h.

The work performed by and referenced in Moghissi et al. (NUREG/CP–0001) indicates that

persons who use timepieces that contain tritiated paints, as opposed to persons who apply the

paints, could receive whole-body dose equivalents on the order of 0.01 mSv/yr (1 mrem/yr). 

These indications are based on measured 3

H levels in the bodies or in urine samples of

persons who used such timepieces.

2.3.3.2 Timepieces Containing 147Pm

The Federal Register notices cited above include the results of an analysis of radiological

impacts on the public from use of timepieces containing 147Pm at the limits for exemption. The

dose estimates obtained by the AEC are summarized as follows:

During normal use of timepieces, beta particles from 147Pm decay do not penetrate

through the watch glass or case, so the dose from external exposure to beta particles

will be zero.

A small amount of bremsstrahlung produced by stopping of beta particles from 147Pm

decay would penetrate the covering of timepieces and the epidermis. However,

available data on radiation levels from timepieces indicate that annual dose equivalents

to an individual continuously wearing a watch containing the maximum exempt activity of

147Pm would probably be less than 4 mSv (<400 mrem) to a small area of skin on the

wrist and less than 0.01 mSv (<1 mrem) to the gonads. These doses are small fractions

of recommended limits for members of the public.
2–53

McDowell-Boyer and O’Donnell (NUREG/CR–0216) and Buckley et al. (NUREG/CR–1775) also

assessed distribution, use, repair, and disposal of timepieces containing 147Pm in paints. They

found that most individuals should receive only tenths of a microsievert (tenths of a millirem) per

year of normal exposure to timepieces containing 147Pm. Even under accident conditions,

individual dose equivalents were on the order of 0.01 mSv (1 mrem). The annual collective

dose equivalent associated with the above conditions was estimated to be 3.9 person-Sv

(390 person-rem).

2.3.4 Current Assessment for Timepieces Containing 3

H

Table 2.3.1 presents the results of the current assessment of potential radiation doses due to

an annual distribution of 10 million 3

H-containing timepieces (7.1 million wristwatches,

2.2 million clocks, and 0.7 million pocket watches). This distribution value is higher than

indicated by current licensing data but is representative of historic values. Because it is not

clear that the differences in the 3

H contents of wristwatches, clocks, and pocket watches

reported by McDowell-Boyer and O’Donnell (NUREG/CR–0216) are still true, each timepiece is

assumed to contain 74 MBq (2 mCi). A leak rate of 1 ppm/h, or 74 Bq/h (0.002 Ci/h), has

been assumed for average conditions and 3 ppm/h, or 220 Bq/h (0.006 Ci/h), has been

assumed for maximum exposure. The useful lifetime of a timepiece is assumed to be 10 years. 

The results obtained for timepieces containing 74 MBq (2 mCi) of 3

H can be scaled linearly to

reflect the potential consequences of distributing timepieces containing the exempt quantity of

3

H, 930 MBq (25 mCi).

The dose estimates presented in the following assessment are based on exposure conditions

(scenarios) developed from the conditions used by McDowell-Boyer and O’Donnell

(NUREG/CR–0216) and Buckley et al. (NUREG/CR–1775). These scenarios describe typical

conditions under which members of the public may interact with timepieces during distribution,

use, and disposal and allow development of reasonable accident scenarios using a consistent

set of assumptions. Scenarios were developed for (1) distribution workers and members of the

public who might be exposed during product distribution, (2) persons who wear or otherwise

use timepieces, (3) persons who are exposed to timepieces worn or used by others, (4) watch

repairmen, (5) storage of obsolete timepieces in the home, (6) disposal in landfills or by

incineration, and (7) a fire in a warehouse or vehicle that contains a large quantity of

timepieces.

Also considered are timepieces containing 930 MBq (25 mCi) of 3

H gas contained in glass

tubes. The typical 3

H release rate for these timepieces is less than 9.2 Bq/h (<0.25 nCi/h),

which corresponds to a release rate of less than 10 ppb/h, adapted from McDowell-Boyer and

O’Donnell (NUREG/CR–0215). Even though the activity is higher (930 MBq (25 mCi) versus

74 MBq (2 mCi)), the hypothetical doses from timepieces containing 3

H in glass tubes will be

less than those for 3

H in paint due to the assumed lower release rate (10 ppb/h versus

1 - 3 ppm/h). 

2.3.4.1 Distribution

The annual distribution of 10 million 3

H-containing timepieces is assumed to consist of 7.1

million wristwatches, 2.2 million clocks, and 0.7 million pocket watches (see Table 2.3.2). Each
2–54

of 10 manufacturers is assumed to distribute 1 million timepieces (71% wristwatches,

22% clocks, and 7% pocket watches) per year as follows:

all timepieces from a manufacturer are loaded into a small express-delivery truck and

transported to a parcel delivery center;

the parcel delivery center handles 1 million timepieces as follows:

– 60,000 timepieces are loaded into 2 large local-delivery trucks (30,000 each) for

transport to two wholesalers, each of whom loads 300 timepieces into each of

100 large local-delivery trucks for transport to 100 small retail stores;

– 120,000 timepieces are loaded into 2 large local-delivery trucks (60,000 each) for

transport to two chain warehouses, each of whom loads 6,000 timepieces into

each of 10 large local-delivery trucks for transport to 10 large retail stores;

– 20,000 timepieces are loaded into 2 large local-delivery trucks (10,000 each) for

transport to two wholesalers, each of whom loads 1,000 timepieces into each of

10 small local-delivery trucks for delivery to individual customers; and 

– 800,000 timepieces are loaded into a large regional-delivery truck for transport to

truck terminal 1; 

truck terminal 1 handles 800,000 timepieces as follows:

– 90,000 timepieces are loaded into 3 large local-delivery trucks (30,000 each) for

transport to three wholesalers, each of whom loads 300 timepieces into each of

100 large local-delivery trucks for transport to 100 small retail stores;

– 180,000 timepieces are loaded into 3 large local-delivery trucks (60,000 each) for

transport to three chain warehouses, each of whom loads 6,000 timepieces into

each of 10 large local-delivery trucks for transport to 10 large retail stores; 

– 30,000 timepieces are loaded into 3 large local-delivery trucks (10,000 each) for

transport to three wholesalers, each of whom loads 1,000 timepieces into each

of 10 small local-delivery trucks for delivery to individual customers, and

– 500,000 timepieces are loaded into a large regional-delivery truck for transport to

truck terminal 2; 

truck terminal 2 handles 500,000 timepieces as follows: 

– 90,000 timepieces are loaded into 3 large local-delivery trucks (30,000 each) for

transport to three wholesalers, each of whom loads 300 timepieces into each of

100 large local-delivery trucks for transport to 100 small retail stores; 

– 180,000 timepieces are loaded into 3 large local-delivery trucks (60,000 each) for

transport to three chain warehouses, each of whom loads 6,000 timepieces into

each of 10 large local-delivery trucks for transport to 10 large retail stores;
2–55

– 30,000 timepieces are loaded into 3 large local-delivery trucks (10,000 each) for

transport to three wholesalers, each of whom loads 1,000 timepieces into each

of 10 small local-delivery trucks for delivery to individual customers; and 

– 200,000 timepieces are loaded into a large regional-delivery truck for transport to

truck terminal 2; and 

truck terminal 3 handles 200,000 timepieces as follows: 

– 60,000 timepieces are loaded into 2 large local-delivery trucks (30,000 each) for

transport to two wholesalers, each of whom loads 300 timepieces into each of

100 large local-delivery trucks for transport to 100 small retail stores;

– 120,000 timepieces are loaded into 2 large local-delivery trucks (60,000 each) for

transport to two chain warehouses, each of whom loads 6,000 timepieces into

each of 10 large local-delivery trucks for transport to 10 large retail stores; and 

– 20,000 timepieces are loaded into two 2 local-delivery trucks (10,000 each) for

transport to two wholesalers, each of whom loads 1,000 timepieces into each of

10 small local-delivery trucks for delivery to individual customers.

The exposure conditions and calculational methods given in Appendix A.3.3 were used to

calculate individual and collective effective dose equivalents (EDEs) for each step in the model. 

The results of the calculations are presented in Table 2.3.2. The highest calculated individual

EDE was approximately 0.09 mSv (9 mrem) to the drivers of large regional delivery trucks that

deliver timepieces from the parcel delivery center to truck terminal 1. The total annual collective

EDE for distribution was about 7 person-Sv (700 person-rem), almost entirely due to exposures

at retail establishments. 

Two assumptions used in the above calculations have a significant effect on the dose

estimates. First, the use of tractor-trailer rigs for regional deliveries would totally remove

exposures to regional-delivery drivers, the most exposed individuals. However, the driver of the

small express-delivery truck who transports 1 million timepieces per year from a manufacturer

to a parcel delivery center could receive an EDE of about 0.04 mSv (4 mrem). Second, the

assumption that one driver transports all timepieces shipped from one origin facility to a

destination facility could be overly conservative. Doses to truck drivers would be reduced in

direct proportion to the number of drivers involved. For example, if two drivers moved

timepieces from the parcel delivery center to truck terminal 1, the dose to each driver would be

one-half the dose to the maximum driver.

2.3.4.2 Routine Use

Timepieces are used in all environments frequented by humans. Two modes of exposure can

occur during routine use of wristwatches containing tritiated paint: (1) exposure to airborne

releases of 3

H from the wristwatches and (2) exposure due to skin contact with the case of the

wristwatch. The latter mode of exposure applies only to wearers of wristwatches; the first mode

applies to wearers of wristwatches and persons in the vicinity of wearers (e.g., coworkers and

other family members).
1

 Multiply the 3H content of a watch (74 MBq (2 mCi)), the leak rate of 3H from a watch

(1 ppm/h).

2

 Multiply the daily rate of 3H leakage from the watch by exposure time of 16 h/d and the fraction

of 3

H released from the watch that is absorbed through the skin (0.02).

3

 Multiply the intake rate of 3H through the skin (24 Bq/day (6.4×104 Ci/day)), the number of

days per year (365 days/yr), and the dose conversion factor for HTO absorbed through the skin

(1.8×103

 mSv-cm2/Bq)(6.7×109 mrem/Ci); divide by the exposed skin area (10 cm2).

4

 Multiply the dose equivalent averaged over 10 cm2 and the fraction of total skin in contact with

the wristwatch (10 cm2

/1.8 m2), where 10 cm2 is the approximate area of the skin in contact with

the watch and 1.8 m2

 is the approximate area of the skin of the whole body.

5

 Multiply the daily intake of HTO through the skin (24 Bq/day (6.4×104 Ci/day)), the number

of days/yr (365 days/yr), and the dose conversion factor for either absorption through the skin

or ingestion of 3

H (1.7×1011Sv/Bq (6.4×105 rem/Ci)).

2–56

This section discusses individual and collective doses to wearers from skin contact with the

wristwatches; doses to wearers, coworkers, and other family members due to airborne releases

during routine use; and doses due to storage of old watches in homes.

2.3.4.2.1 Skin Contact With Wristwatch Cases

To estimate the potential radiation doses due to skin absorption of 3

H from a wristwatch initially

containing 74 MBq (2 mCi) of 3

H in paint, the procedure described in Section 2.14.4.2.1 was

used. First, determine the 3

H leakage from a watch, 74 Bq/h (0.002 Ci/h).1 Second,

determine the intake of tritiated water vapor (HTO) through the skin in contact with the case of

the watch, 24 Bq/day (6.4×104 Ci/day).2

 Third, determine the annual dose equivalent to the

skin in contact with the case, 2 mSv (200 mrem)3

 when averaged over an area of 10 cm2. 

Fourth, determine the average annual dose equivalent to the skin of the whole body from the

distributed wristwatch source, 0.001 mSv (0.1 mrem).4

 Fifth, determine the contribution of this

skin dose equivalent to the annual EDE, 1×105

 mSv (0.001 mrem), by multiplying the skin dose

equivalent and the organ weighting factor for skin of the whole body (0.01). Sixth, determine

the annual EDE to the internal organs of the body from the absorption of HTO through the skin

in contact with the case of the watch, 2×104

 mSv (0.02 mrem).5

In summary, the annual dose equivalent to skin is estimated to be 2 mSv (200 mrem) when

averaged over an area of 10 cm2

 in contact with the wristwatch, the skin dose due to the

distributed wristwatch source of 3

H makes a negligible contribution to the annual EDE, and the

total annual EDE to a wearer from skin absorption of 3

H in contact with the case of a 3H

containing watch is estimated to be 2×104

 mSv (0.02 mrem). The above discussion applies to

an individual who wears a wristwatch 16 h/day for 365 days/yr. The collective EDE from use of

7.1 million wristwatches during the first year of use could be 1 person-Sv (100 person-rem). 

The total collective EDE over a 10-year useful life of 7.1 million watches is estimated to be

8 person-Sv (800 person-rem).
2–57

The above dose estimates would change (increase) by a factor of 12.5 for timepieces

containing the exempt limit of 3H in paint. For timepieces containing 930 MBq (25 mCi) of 3H

gas, the above dose estimates would change (decrease) by a factor of 0.12 because of the

lower leak rate.

2.3.4.2.2 Airborne Releases from Wristwatch

Because watches may be worn in a variety of different ways during routine use, the following

four scenarios were chosen to indicate the potential dose from airborne releases of 3

H from

timepieces containing 3

H in paint.

Scenario I. A watch wearer spends 12 hours at home each day (4380 h/yr) and exposes three

other family members to airborne releases of 3

H from the wristwatch. The home has an

enclosed volume of 450 m3

 and a ventilation rate of 1 volume change per hour. The average

concentration of HTO in the air of the home over a 12-hour period during the first year is

approximately 0.16 Bq/m3

 (4.4 pCi/m3) and the breathing rate of the individuals is 0.9 m3/h. 

Thus, the annual EDE to the wearer and to other family members could be 2×105

 mSv

(0.002 mrem), assuming the other family members are exposed over the same 12 h/day as the

wearer. The collective EDE to the family could be about 7×108

 person-Sv (7×106 person-rem)

for the first year of use and 5×107

 person-Sv (5×105 person-rem) over 10 years of use.

Scenario II. A clock is kept in the home for 24 h/day (8760 h/yr) and exposes a family of four to

airborne releases of 3

H. The home has an enclosed volume of 450 m3 and a ventilation rate of

1 volume change per hour. The average concentration of HTO in the air of the home during the

first year is approximately 0.16 Bq/m3

 (4.4 pCi/m3) and the breathing rate of the individuals is

0.9 m3

/h. Thus, the annual EDEs to family members could be 3×105mSv (0.003 mrem) from

spending 20 h/day (7300 h/yr) at home. For family members, the collective dose is

1×107

 person-Sv (1×105 person-rem) for the first year of use and 9×107 person-Sv

(9×105

 person-rem) over 10 years of use.

Scenario III. A watch wearer works 8 h/day (2000 h/yr) in an office or shop and exposes two

coworkers to airborne release of 3

H from the watch. The office or shop has an enclosed

volume of 34 m3

 and a ventilation rate of 1 volume change per hour. The average

concentration of HTO in the air of the office or shop over an 8-hour period is approximately

2.2 Bq/m3

 (0.059 nCi/m3) and the breathing rate of the individuals is 1.2 m3/h. Thus, the annual

EDE to the wearer and two coworkers could be 1×104

 mSv (0.01 mrem), assuming the

coworkers are exposed over the same 8 h/day as the wearer. The collective EDE to the office

staff could be about 4×107

 person-Sv (4×105 person-rem) for the first year of use and

3×106

 person-Sv (3×104 person-rem) over 10 years of use.

Scenario IV. A clock is kept continuously in an office or a shop. Three workers spend 8 h/day

(2000 h/yr) in the office or shop. The office or shop has an enclosed volume of 34 m3

 and a

ventilation rate of 1 volume change per hour. The average concentration of HTO in the air of

the office or shop during the first year is approximately 2.2 Bq/m3

 (0.059 nCi/m3) and the

breathing rate of the individuals is 1.2 m3

/h. Thus, the annual EDE to the wearer and two

coworkers could be 1×104

 mSv (0.01 mrem). The collective EDE to the office staff could be

about 4×107

 person-Sv (4×105 person-rem) for the first year of use and 3×106 person-Sv

(3×104

 person-rem) over 10 years of use.
2–58

To estimate the total collective EDE from airborne releases of 3

H during routine use of the

self-luminous watches, it is assumed that all (7.8 million) of the watches are used in Scenario I,

50% (3.9 million) of the watches are also used in Scenario III, 50% (1.1 million) of the clocks

are also used in Scenario II, and that the remaining 50% of clocks are those used under

Scenario IV. Thus, the total collective EDE over 10 years from use in Scenarios I through IV

would be about 20 person-Sv (2000 person-rem).

2.3.4.2.3 Total Individual and Collective Doses

For a 16-hour-per-day wearer of a self-luminous wristwatch containing 74 MBq (2 mCi), the

annual dose equivalent to skin from routine use could be 2 mSv (200 mrem) when averaged

over an area of 10 cm2

 in contact with the wristwatch.

The annual individual EDE to such a wearer from routine use could be 4×104

 mSv (0.04 mrem)

with half being from absorption of 3

H through the skin in contact with the watch and the other

half from airborne releases of 3

H at work and home. The individual dose to coworkers and

other family members is significantly less.

The total collective dose equivalent over a ten-year useful life to both wearers and other

members of the public from routine use of 1 year’s distribution of 10 million timepieces each

containing 74 MBq (2 mCi) of 3

H could be 30 person-Sv (3000 person-rem). Of this total,

10 person-Sv (1000 person-rem) is due to exposure to wearers via absorption of 3

H through the

skin in contact with the case of the watch. The remaining 20 person-Sv (2000 person-rem) is

due to airborne releases from the watch while at work or at home.

2.3.4.3 Watch Repair

Timepiece repairmen may repair, adjust or replace batteries in the watches. It is unlikely that

3

H will be released catastrophically from the timepieces during repair; however, such a release

is modeled in Section 2.3.4.5 for the 3

H in glass tubes.

For a repairman at a jewelry store, potential doses were estimated using the following

scenarios: (1) the shop had an enclosed volume of 34 m3

 and a ventilation rate of 1 volume

change per hour, and the repairman was exposed to airborne leakage of 3

H from a timepiece

for 1 day (8 hours) between the time the timepiece was received at the shop and returned to the

owner, (2) the average time for repair, adjustment and battery replacement was 10 minutes,

and the repairman was exposed during this time to airborne releases of 3

H into a small

hemispherical air space with a radius of 1.5 meters, a volume of 7 m3

, and a ventilation rate of

1 volume change per hour, and (3) during repair, 20% of the 3

H escaping from the timepiece

was absorbed through a skin area of 3 cm2

 on the ends of the repairman’s fingers. From

discussions with a local watch repair shop, the average time to repair a watch was estimated to

be 45 minutes but could take up to 3 hours for an automatic watch. Change a battery takes

about 1 minute but could take up to 10 minutes for a waterproof watch. Since most watch

maintenance is battery replacement, an average exposure time of 10 minutes was assumed. 

Based on these assumptions, a breathing rate of 1.2 m3

/h for light activity, and the repair of 100

timepieces per year, the EDE to the repairman from all exposure pathways could be

5×105

 mSv (0.005 mrem), and the dose equivalent to the skin could be less than 1×105 mSv

(<0.001 mrem) when averaged over an area of 3 cm2

 in contact with the timepieces.
2–59

If a timepiece is serviced every 2 years, mainly due to battery replacement, the collective EDE

to repairmen could be 0.02 person-Sv (2 person-rem) over the 10-year effective lifetime of the

10 million timepieces distributed annually.

2.3.4.4 Disposal

Under normal circumstances, timepieces would be disposed of as ordinary, non combustible

household trash. The following assessment assumes discard of 420 TBq (11 kCi) of 3

H in

10 million 10-year-old timepieces in 1 year.

Using the assumptions of the generic disposal methodology (see Appendix A.2) for disposal of

420 TBq (11 kCi) of 3

H, the highest calculated individual EDE is 0.002 mSv (0.2 mrem), to a

waste collector at a municipal incinerator assuming 20% of the watches are incinerated. All

other doses would be less. The total collective EDE to all workers and potentially exposed

members of the public could be about 0.1 person-Sv (10 person-rem).

2.3.4.5 Accidents and Misuse

To bound the potential consequences of accidents or misuses involving timepieces containing

tritiated paints, the following things are considered: (1) a fire during transport of a large

shipment of 400 timepieces and (2) accidental ingestion of 10% of the tritiated paint. For a

shipment of 400 timepieces containing 30 GBq (0.8 Ci) of 3

H using the general modeling of

Appendix A.1, the EDE associated with a transportation fire could be 6×104

 mSv (0.06 mrem).

Someone accidently ingesting 10% of the paint contained in a timepiece would intake 7.4 MBq

(200 Ci) of 3

H. Such an intake would produce an EDE of about 0.1 mSv (10 mrem). Ingestion

of 10% of the exemption limit of 930 MBq (25 mCi) would produce an EDE of about 1 mSv

(100 mrem).

In the case of accidents for timepieces containing 3

H in glass tubes, the following was

considered: (1) a catastrophic release from crushing of a single watch in a repair shop, (2) an

accident involving the crushing of a single watch in a home, and (3) a shipping accident in a

storeroom or cargo-handling area involving the crushing of a shipment of 200 watches. The 3

H

contained in the watches is assumed to be 99% HT and 1% HTO. Based on these

assumptions and the generic accident methodology in Appendix A.1, the potential radiation

doses from the crushing of self-luminous watches containing 3

H can be summarized as follows:

For a watch repairman, the individual EDE from crushing a single watch containing

930 MBq (25 mCi) of 3

H could be 0.02 mSv (2 mrem) at a small repair shop or

0.008 mSv (0.8 mrem) at a large repair shop.

For a person at home, the individual EDE from crushing a single watch containing

930 MBq (25 mCi) of 3

H could be 5×104 mSv (0.05 mrem).

For a worker in a storeroom or cargo-handling area, the individual EDE from crushing

200 watches containing a total of 185 GBq (5 Ci) of 3

H could be 0.05 mSv (5 mrem).

In the case of misuse, this analysis considers the exposure to a 5-year-old child who plays with

a self-luminous watch as a glow-in-the-dark” toy at night while going to sleep during one year. 
6

 The dose conversion factors for effective dose equivalent due to ingestion of 3H or absorption

of 3

H through the skin are the same numerically.

2–60

It is assumed that (1) the watch is a 10-year-old watch containing 560 MBq (15 mCi) of 3

H,

(2) the child handles the watch for 10 min/day, (3) the child absorbs 2% of the 3

H released from

the watch through a skin area of 10 cm2

 while handling the watch, and (4) the child sleeps in a

closed bedroom with the watch for 12 h/day. It is further assumed that (1) the bedroom has an

enclosed volume of 27 m3

 and a ventilation rate of 1 air change per hour (see Appendix A.1),

(2) the child’s breathing rate is 0.24 m3

/h while sleeping (ICRP 66), (3) the dose conversion

factors for inhalation and ingestion6

 are about twice those for an adult (ICRP 67; ICRP 71), and

(4) the total surface area of the child's skin is approximately 0.8 m2

 (ICRP 23). Based on these

assumptions, the potential radiation doses to the 5-year-old child can be summarized as

follows:

The dose equivalent to the skin of the 5-year-old child due to absorption of 3

H from the

watch could be 0.001 mSv (0.1 mrem) over a skin area of 10 cm2

 in contact with the

watch.

The EDEs would be less than 1×105

 mSv (<0.001 mrem) due to absorption of 3H

through the skin in contact with the watch and 1×105

 mSv (0.001 mrem), due to

airborne releases of 3

H from the watch.

2.3.5 Current Assessment for Timepieces Containing 147Pm

Table 2.3.3 presents the results of the current assessment of potential radiation doses due to

an annual distribution of 1 million 147Pm-containing timepieces. The dose estimates presented

in the following assessment use exposure conditions (scenarios) developed from the conditions

used by McDowell-Boyer and O’Donnell (NUREG/CR–0216). These scenarios describe typical

conditions under which members of the public may interact with timepieces during distribution,

use, and disposal and allow development of reasonable accident scenarios using a consistent

set of assumptions. Scenarios were developed for (1) distribution workers and members of the

public who might be exposed during product distribution, (2) persons who wear or otherwise

use timepieces, (3) persons who are exposed to timepieces worn or used by others, (4) watch

repairmen, (5) storage of obsolete timepieces in the home, (6) disposal in landfills or by

incineration, and (7) a fire in a warehouse or vehicle that contains a large quantity of

timepieces.

This assessment is based on an annual distribution of 1 million timepieces that contain 147Pm,

0.5 million wristwatches, 0.5 million clocks, and no pocket watches. The assumed 147Pm

content of each timepiece is 3.7 MBq (100 Ci); some of the later licensing data indicate that

the 147Pm contents of wristwatches and clocks are approximately the same. 

In all but the accident scenarios, the only credible mode of exposure is external irradiation by

bremsstrahlung produced by the stopping in timepiece components of the beta particles emitted

during decay of 147Pm.
2–61

2.3.5.1 Distribution

The annual distribution of 1 million 147Pm-containing timepieces is assumed to consist of

0.5 million wristwatches, 0.5 million clocks, and no pocket watches (see Table 2.3.4). One

manufacturer is assumed to distribute all the timepieces as follows:

all timepieces from a manufacturer are loaded into a small express-delivery truck and

transported to a parcel delivery center;

the parcel delivery center handles 1 million timepieces as follows: 

– 60,000 timepieces are loaded into 2 large local-delivery trucks (30,000 each) for

transport to two wholesalers, each of whom loads 300 timepieces into each of

100 large local-delivery trucks for transport to 100 small retail stores;

– 120,000 timepieces are loaded into 2 large local-delivery trucks (60,000 each) for

transport to two chain warehouses, each of whom loads 6,000 timepieces into

each of 10 large local-delivery trucks for transport to 10 large retail stores;

– 20,000 timepieces are loaded into 2 large local-delivery trucks (10,000 each) for

transport to two wholesalers (catalog centers), each of whom loads 1,000

timepieces into each of 10 small local-delivery trucks for delivery to individual

customers; and 

– 800,000 timepieces are loaded into a large regional-delivery truck for transport to

truck terminal 1; 

truck terminal 1 handles 800,000 timepieces as follows: 

– 90,000 timepieces are loaded into 3 large local-delivery trucks (30,000 each) for

transport to three wholesalers, each of whom loads 300 timepieces into each of

100 large local-delivery trucks for transport to 100 small retail stores;

– 180,000 timepieces are loaded into 3 large local-delivery trucks (60,000 each) for

transport to three chain warehouses, each of whom loads 6,000 timepieces into

each of 10 large local-delivery trucks for transport to 10 large retail stores;

– 30,000 timepieces are loaded into 3 large local-delivery trucks (10,000 each) for

transport to three wholesalers, each of whom loads 1,000 timepieces into each

of 10 small local-delivery trucks for delivery to individual customers; and 

– 500,000 timepieces are loaded into a large regional-delivery truck for transport to

truck terminal 2; 

truck terminal 2 handles 500,000 timepieces as follows:

– 90,000 timepieces are loaded into 3 large local-delivery trucks (30,000 each) for

transport to three wholesalers, each of whom loads 300 timepieces into each of

100 large local-delivery trucks for transport to 100 small retail stores;
2–62

– 180,000 timepieces are loaded into 3 large local-delivery trucks (60,000 each) for

transport to three chain warehouses, each of whom loads 6,000 timepieces into

each of 10 large local-delivery trucks for transport to 10 large retail stores;

– 30,000 timepieces are loaded into 3 large local-delivery trucks (10,000 each) for

transport to three wholesalers, each of whom loads 1,000 timepieces into each

of 10 small local-delivery trucks for delivery to individual customers; and 

– 200,000 timepieces are loaded into a large regional-delivery truck for transport to

truck terminal 3; and 

truck terminal 3 handles 200,000 timepieces as follows: 

– 60,000 timepieces are loaded into 2 large local-delivery trucks (30,000 each) for

transport to two wholesalers, each of whom loads 300 timepieces into each of

100 large local-delivery trucks for transport to 100 small retail stores;

– 120,000 timepieces are loaded into 2 large local-delivery trucks (60,000 each) for

transport to two chain warehouses, each of whom loads 6,000 timepieces into

each of 10 large local-delivery trucks for transport to 10 large retail stores; and 

– 20,000 timepieces are loaded into 2 large local-delivery trucks (10,000 each) for

transport to two wholesalers, each of whom loads 1,000 timepieces into each of

10 small local-delivery trucks for delivery to individual customers.

The exposure conditions and calculational methods given in Appendix A.3.3 were used to

calculate hypothetical individual and collective EDEs for each step in the model. The results of

the calculations are presented in Table 2.3.4. The highest calculated individual EDE was

approximately 0.009 mSv (0.9 mrem), to a worker in the parcel delivery center. The total

collective EDE for distribution was about 0.02 person-Sv (2 person-rem), almost entirely due to

exposures at retail establishments. Because of the number of cartons carried in each truck, the

average truck driver exposure conditions were used in the calculations. (See Appendix A.3.3.)

The assumptions used in the above calculations have a significant effect on the dose

estimates. First, the use of different-sized trucks for transport would lower the dose estimates

in nearly all cases, and second, the assumption that one driver transports all timepieces from

one manufacturer to one parcel delivery center could be overly conservative. Doses to the

truck driver and the center workers would be reduced in direct proportion to the number of

facilities involved.

2.3.5.2 Routine Use

Since timepieces may be used in a variety of ways, the following three scenarios were chosen

to indicate potential doses to users of 147Pm-containing timepieces.

Scenario I. A watch user wears the watch on the outside of the wrist for 16 h/day (5840 h/yr). 

This person keeps the watch-bearing arm at the side, at an effective distance of 42 cm from the

body, for 4330 h/yr; near the head, at an effective distance of 63 cm from the total body, for
2–63

470 h/yr; and near the stomach, at an effective distance of 21 cm, for 1040 h/yr. This

orientation allows the arm to act as a 5-cm thick absorber. Nonwearers exposed to the

timepiece include 3 other family members, who are 3 meters from the timepiece for 5840 h/yr,

and an average of 20 office workers and passersby, who are 6 meters from the watch for

2000 h/yr. Dose calculations were performed using CONDOS (Computer Codes, O’Donnell et

al., 1975) assuming a point source with a steel (iron) watch casing and a 0.3 cm glass lens. As

discussed in Appendix A.4, a factor of 15 reduction has been applied because of the over

estimation by CONDOS of bremsstrahlung reduction of low energies.

Annual EDEs could hypothetically be 0.002 mSv (0.2 mrem) to the wearer, 3×105

 mSv

(0.003 mrem) to other family members, and less than 1×105

 mSv (<0.001 mrem) to other

persons. The collective EDE could be about 2×106

 person-Sv (2×104 person-rem) for the first

year of use and 2×105

 person-Sv (2×103 person-rem) over 10 years of use.

Scenario II. To bound the potential exposure to timepieces, a watch user wears the watch on

the inside of the wrist for 16 h/day (5840 h/yr). This person keeps the watch-bearing arm at the

side, at an effective distance of 37 cm from the body, for 4330 h/yr; near the head, at an

effective distance of 53 cm from the total body, for 470 h/yr; and near the stomach, at an

effective distance of 16 cm, for 1040 h/yr. In this orientation the arm provides no shielding to

the wearer, but it does provide shielding for nonwearers. As in scenario 1, nonwearers exposed

to the timepiece include 3 other family members, who are 3 meters from the timepiece for

5840 h/yr, and an average of 20 office workers and passersby, who are 6 meters from the

watch for 2000 h/yr.

Annual EDEs could potentially be 0.004 mSv (0.4 mrem) to the wearer, 1×105

 mSv

(0.001 mrem) to other family members, and less than 1×105

 mSv (<0.001 mrem) to other

persons. The collective EDE could be about 7×108

 person-Sv (7×106 person-rem) for the first

year of use and 5×107

 person-Sv (5×105 person-rem) over 10 years of use.

Scenario III. A clock is kept in the home or office. In the home, four family members are

exposed at an average distance of 3 meters for 12 h/day (4380 h/yr). In an office, the most

exposed person is located 3 meters from the clock for 2000 h/yr and 100 other people are an

average distance of 6 meters away for 100 h/yr.

Annual EDEs could be 2×105

 mSv (0.002 mrem) to home users and less than 1×105 mSv

(<0.001 mrem) to the office work and other persons. The collective EDE could be about

8×108

 person-Sv (8×106 person-rem) for the first year of use and 6×107 person-Sv

(6×105

 person-rem) over 10 years of use.

To estimate the total collective EDE from timepieces over 10 years of use, it is assumed that

250,000 watches are worn on the outside of the arm and 250,000 are worn on the inside of the

arm. Thus the collective EDE from wearing 0.5 million wristwatches could be 5 person-Sv

(500 person-rem) over a 10-year useful life. Likewise, if 250,000 clocks are used in homes and

250,000 are used in business settings, the collective EDE could be 0.2 person-Sv

(20 person-rem).
2–64

2.3.5.3 Watch Repair

Timepiece repairmen may adjust or replace batteries in the watches. Other repair operations

are unlikely because of economic factors.

For a repairman at a jewelry store, potential doses were estimated by assuming that the

repairman was exposed to a timepiece for 1 day (8 hours), at an average distance of 3 meters

from the timepiece, between the time the timepiece was received at the shop and returned to

the owner and at an average distance 30 cm from the timepiece while handling it for 10 minutes

to adjust and replace the battery. Based on these assumptions and the repair of 100

timepieces per year, the EDE to the repairman from all exposure pathways could be

7×105

 mSv/yr (0.007 mrem/yr).

If a timepiece is serviced every 2 years, mainly to have a battery replaced, the collective EDE to

repairmen could be 0.001 person-Sv (0.1 person-rem) over the 10-year effective lifetime of the

1 million timepieces distributed annually.

2.3.5.4 Disposal

Under normal circumstances, timepieces would be disposed of as ordinary, noncombustible

household trash. The following assessment assumes discard of 0.25 TBq (6.7 Ci) of 147Pm in

1 million 10-year-old timepieces in 1 year.

Using the assumptions of the generic disposal methodology (see Appendix A.2) for disposal of

0.25 TBq (6.7 Ci) of 147Pm, the highest calculated individual EDE is 2×105

 mSv (0.002 mrem),

to a waste collector at a municipal incinerator, assuming 20% of watches are incinerated. For

all other individuals the dose is less than 1×105

 mSv (<0.001 mrem). The total collective EDE

to all workers and potentially exposed members of the public could be about 4×105

 person-Sv

(4×103

 person-rem).

2.3.5.5 Accidents and Misuse

To bound the potential consequences of accidents or misuses involving timepieces containing

147Pm in paints, the following scenarios are considered: (1) a fire during transport of a large

shipment of 400 timepieces and (2) accidental ingestion of 10% of the 147Pm paint. Using the

generic accident methodology for a transportation fire (Appendix A.1), the potential EDE per

kilobecquerel of 147Pm involved is 9×1012 mSv (EDE per microcurie of 147Pm involved is

3.4×1010 rem). Since a shipment of 400 timepieces contains 1.5 GBq (40,000 Ci) of 147Pm,

the EDE associated with a transportation fire could be 1×105

 mSv (0.001 mrem).

Someone accidently ingesting 10% of the paint contained in a timepiece would intake 0.37 MBq

(10 Ci) of 147Pm. Such an intake would produce an EDE of about 0.1 mSv (10 mrem). 

Ingestion of 10% of the exemption limit of 7.4 MBq (200 Ci) for any timepiece, other than a

watch, would produce an EDE of about 0.2 mSv (20 mrem).
2–65

2.3.6 Summary

Tables 2.3.1 and 2.3.3 present the results of the current assessments of potential radiation

doses to the public from timepieces containing 3

H and 147Pm in paint, respectively. For 3H,

these results are based on annual distribution of 10 million timepieces (7.1 million watches,

2.2 million clocks, and 0.7 million pocket watches). Whereas for 147Pm, these results are based

on an annual distribution of 1 million timepieces (0.5 million wristwatches, 0.5 million clocks,

and no pocket watches). The 3

H timepieces are assumed to contain 74 MBq (2 mCi) of 3H per

timepiece and the 147Pm timepieces are assumed to contain 3.7 MBq (100 Ci) of 147Pm per

timepiece. All of the timepieces are assumed to have a useful lifetime of 10 years. The

estimate of the total collective EDE to the public from timepieces containing 3

H is 60 person-Sv

(6000 person-rem) (see Table 2.3.1). For timepieces containing 147Pm, the estimate of the total

collective EDE to the public is 5 person-Sv (500 person-rem) (see Table 2.3.3).
2–66

Table 2.3.1 Summary of Potential Radiation Doses From Timepieces Containing

3

H in Paint

Exposure Scenario

Highest Individual Annual

Effective Dose Equivalent

Ratea

(mrem)

Collective Effective Dose

Equivalenta

(person-rem)

Distribution 9 700

Routine useb

 Skin absorption

 In the home and office

Total

0.02

0.04

0.06

800

2000

3000

Maintenance and repairb 0.005 2

Disposal 0.20 10

Accidents or misusec 10 NAd

a

 Refer to text discussion for time period of collective dose calculation. 1 mrem = 0.01 mSv;

1 person-rem = 0.01 person-Sv.

b

 Dose estimates are based on the assumption that each timepiece contains 74 MBq (2 mCi) of

3

H. To estimate doses at the exemption limit, 930 MBq (25 mCi), multiply these doses by 12.5;

and to estimate doses at the higher leak rate of 3 ppm/h, versus 1 ppm/h average, multiply

these doses by 3. To estimate doses for timepieces containing 930 MBq (25 mCi) of 3

H gas,

multiply these doses by 0.12. Collective doses are based on an annual distribution of 10 million

timepieces.

c

 Based on average activity of 74 MBq (2 mCi) of 3H. To estimate dose at the exemption level

of 930 MBq (25 mCi), multiply by 12.5.

d

 Not applicable.
2–67

Table 2.3.2 Summary of Model and Potential Individual and Collective Effective Dose

Equivalents for Distribution of Timepieces Containing 3

H in Paint

Step Representation

Pieces per

Facility

Number

of

Facilities

 Individual

Effective

Annual

Dose

Equivalent

(mrem)a

Collective

Effective

Dose

Equivalent

(person-rem)b

To parcel

 delivery

 center

Express delivery,

 small truck

1,000,000 10 4 0.04

At parcel

 delivery

 center

Medium

 warehouse

1,000,000 10 2 0.3

To regional

 truck center 1

Regional

delivery,

 large truck

800,000 10 9 0.09

At regional

 truck center 1

Large warehouse 800,000 10 1 0.06

To regional

 truck center 2

Regional

delivery,

 large truck

500,000 10 5 0.05

At regional

 truck center 2

Large warehouse 500,000 10 1 0.04

To regional

 truck center 3

Regional

delivery,

 large truck

200,000 10 2 0.02

At regional

 truck center 3

Large warehouse 200,000 10 0.4 0.02

To catalog

 center

Regional

delivery,

 large truck

10,000 100 0.08 0.008

At catalog

 center

Medium

 warehouse

10,000 100 0.02 0.03

To customers Local delivery,

 small truck

1,000 1,000 0.01 0.01
2–68

Table 2.3.2 Summary of Model and Potential Individual and Collective Effective Dose

Equivalents for Distribution of Timepieces Containing 3

H in Paint (continued)

Step Representation

Pieces

per

Facility

Number

of

Facilities

 Individual 

Effective

Annual

Dose

Equivalent

(mrem)a

Collective

Effective

Dose

Equivalent 

(person-rem)b

To wholesaler Local delivery,

 large truck

30,000 100 0.2 0.02

At wholesaler Medium

 warehouse

30,000 100 0.06 0.09

To small store Local delivery,

 small truck

300 10,000 0.002 0.02

At small store Small store 300 10,000 0.04 100

To chain

 warehouse

Local delivery,

 large truck

60,000 100 0.5 0.05

At chain

 warehouse

Medium

 warehouse

60,000 100 0.1 0.2

To large store Local delivery,

 large truck

6,000 1,000 0.05 0.05

At large store Large store 6,000 1,000 0.2 600

Total 700

a

 1 mrem = 0.01 mSv.

b

 1 person-rem = 0.01 person-Sv. Refer to text for time period of collection dose calculators.
2–69

Table 2.3.3 Summary of Potential Radiation Doses From Timepieces Containing

147Pm in Painta

Exposure Scenario

Highest Individual Annual

Effective Dose Equivalent Rate

(mrem)b

Collective Effective Dose

Equivalent

(person-rem)b

Distribution 1 2

Routine usea 0.4 500

Maintenance and repaira 0.007 0.1

Disposal 0.002 0.004

Accidents or misuse 10 NAc

a

 Dose estimates are based on the assumption that each timepiece contains 3.7 MBq (100 Ci)

of 147Pm. To estimate the dose at the exemption level of 7.4 MBq (200 Ci) for any timepiece,

other than watch, multiply doses by 2.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv. Refer to text for the period of collective

dose calculations.

c

 Not applicable.
2–70

Table 2.3.4 Summary of Model and Potential Individual and Collective Effective Dose

Equivalents for Distribution of Timepieces Containing 147Pm in Paint

Step Representation

Pieces per

Facility

Number

of

Facilitie

s

Individual

Annual

Effective

Dose

Equivalent

(mrem)a

Collective

Effective

Dose

Equivalent

(person-rem)b

To parcel

 delivery

 center

Express

 delivery, small

 truck

1,000,000 1 0.2 3×104

At parcel

 delivery

 center

Medium

 warehouse

1,000,000 1 0.9 0.006

To regional

 truck center 1

Regional

 delivery, large 

 truck

800,000 1 0.05 6×105

At regional

 truck center 1

Large

 warehouse

800,000 1 0.2 0.001

To regional

 truck center 2

Regional

 delivery, large

 truck

500,000 1 0.03 4×105

At regional

 truck center 2

Large

 warehouse

500,000 1 0.1 7×104

To regional

 truck center 3

Regional

 delivery, large

 truck

200,000 1 0.01 1×105

At regional

 truck center 3

Large

 warehouse

200,000 1 0.04 3×104

To catalog

 center

Regional

 delivery, large

 truck

10,000 10 0.004 5×105

At catalog

 center

Medium

 warehouse

10,000 10 0.009 6×104

To customers Local delivery,

 small truck

1,000 100 5×104 6×105
2–71

Table 2.3.4 Summary of Model and Potential Individual and Collective Effective Dose

Equivalents for Distribution of Timepieces Containing 147Pm in Paint (continued)

Step Representation

Pieces per

Facility

Number

of

Facilities

Individual

Annual

Effective

Dose

Equivalent

(mrem)a

Collective

Effective

Dose

Equivalent

(person-rem)b

To wholesaler Local delivery,

 large truck

30,000 10 0.01 1×104

To small store Local delivery,

 small truck 300

1,000 1×104 1×104

At small store Small store 300 1,000 5×103 1

To chain

 warehouse

Local delivery,

 large truck

60,000 10 0.03 3×104

At chain

 warehouse

Medium

 warehouse

60,000 10 0.06 0.003

To large store Local delivery,

 large truck

 6,000 100 0.003 3×104

At large store Large store 6,000 100 0.08 1

Total 2

a

 1 mrem = 0.01 mSv.

b

 1 person-rem = 0.01 person-Sv. Refer to text for time period of collection dose calculations.

2–73

2.4 Automobile Lock Illuminators

2.4.1 Introduction

In 10 CFR Part 30.15(a)(2), persons who receive, possess, use, transfer, own, or acquire

automobile lock illuminators are exempted from licensing requirements for byproduct material,

provided that such illuminators (1) do not contain more than 555 megabecquerel (MBq)

(15 millicurie (mCi)) of tritium (3

H) or 74 MBq (2 mCi) of 147Pm per illuminator and (2) the

absorbed dose rate at 1 cm from any surface of an illuminator containing 147Pm (when

measured through 50 mg/cm2

 absorber) does not exceed 0.01milligray (mGy)/h (1 mrad/h). 

The exemption first appeared as a notice of receipt of petition for use of 3

H in automobile lock

illuminators on April 26, 1961 (26 FR 3571), and was issued final on November 7, 1961

(26 FR 10472). Also on November 7, 1961 (26 FR 10487), a second notice was issued

proposing the requirements on manufacturers and importers, with the final ruling on March 31,

1962 (27 FR 3123). Later, the exemption was amended to add provisions for 147Pm. This

exemption was proposed on May 20, 1964 (29 FR 6562), and finalized on March 13, 1965

(30 FR 3374).

The information in the Federal Register notices on potential radiological impacts on the public

from use of lock illuminators containing 3

H is discussed in Section 2.4.3.

2.4.2 Description of Items

Self-luminous paint containing 3

H or 147Pm maybe used in automobile lock illuminators so the

locks could be seen easily in the dark. Attempts to determine the current technology of lock

illumination were futile, but with the development of more sophisticated forms of illumination

(e.g., fiber optics), radioactive self-luminous paints containing either 3

H or 147Pm apparently are

not being used. It is believed that automobile lock illuminators containing 3

H or 147Pm have

never been manufactured for commercial use. 

2.4.3 Summary of Previous Analyses and Assessments

The Federal Register notices cited in Section 2.4.1 contain the only previously published

information on radiological impacts on the public from use and disposal of automobile lock

illuminators. Radioactive self-luminous paints in lock illuminators authorized for use under the

exemption would contain either 3

H or 147Pm. In evaluating the dose from exposure to 3H in lock

illuminators, a potential release was assumed to occur as a result of burning prior to automobile

salvage (26 FR 10472). Thus, the maximum hypothetical dose to individuals near open field

burning of vehicles prior to salvage was estimated. Collective doses, however, were not

estimated. Individual doses were estimated to be 0.025 millisievert (mSv) (2.5 mrem) to a

maximally exposed individual near 50 burning automobiles that each contained three lock

illuminators.

For lock illuminators containing 147Pm, the analyses by the Atomic Energy Commission

(29 FR 6562) included only external exposure from bremsstrahlung to occupants of

automobiles. Collective doses again were not estimated. Individual doses were estimated to

be 0.01 mSv/yr (1.0 mrem/yr) to the gonads of a maximally exposed individual occupying the
2–74

front seat of an automobile for long periods of time (e.g., taxi driver). This was based on the

dose limit of 0.01 mGy/h (1 mrad/h) at 1 cm specified in the exemption.

2.4.4 Present Exemption Analysis

Although it appears that self-luminous paints containing 3

H were never used in lock illuminators,

such uses are allowed under this exemption and could occur in the future. In this assessment,

hypothetical doses are estimated for distribution of automobiles, routine use (private and

commercial), disposal as solid waste (landfill, incineration and recycle), accidents involving a

fire, and potential misuse. Collective doses were not determined as this product is not believed

to be currently manufactured or in use. However, for purposes of modeling individual doses

during transport and distribution, and disposal, it is assumed that 1 million lock illuminators per

year are distributed with radioactive self-luminous paint containing 555 MBq (15 mCi) of 3

H or 1

million automobile lock illuminators containing 74 MBq (2 mCi) of 147Pm. These specific

quantities of 3

H and 147Pm are the limits for the exemptions. There are assumed to be two lock

illuminators per automobile. The assessments of routine exposure to 3

H during distribution and

transport and during routine use assume that leakage from lock illuminators occur into occupied

areas (i.e., the cab of a truck, showroom of a dealership, or interior of an automobile). This

assumption is conservative in all cases.

2.4.4.1 Distribution and Transport

The manufacture, installation, importation, and distribution of lock illuminators are not included

in this exemption (30 FR 3374). The transport of vehicles via truck transport from the

manufacturer to the retailer and retail sales are, however, evaluated in this assessment. Based

on industry information, the maximum number of automobiles per truck trailer is 12. Each

retailer receives 100 automobiles per year (at 10 per shipment; 10 shipments per year). The

doses from transport are estimated using the generic distribution methodology in Appendix A.3.

Table 2.4.1 includes the individual effective dose equivalents (EDEs) from the distribution of

1 million automobiles to 10,000 retailers from both 3

H and 147Pm. Because of the nature of

automobile distribution (i.e., virtually no handling of the product itself), only exposure to a lock

illuminator while the automobile is within the showroom of a dealership during retailing is

assumed. Ten automobiles are assumed to be on display at any one time. The scenario

chosen from Appendix A.3 for retail sales was for a dealership showroom assumed to be the

same as a medium warehouse. The resulting doses for retail sales are likely conservative

since the assumptions used in Appendix A.3 include more contact with the product than

routinely occurs at an automobile dealership. Additional exposure scenarios and dose

estimates are described in the following paragraphs.

2.4.4.1.1 Tritium

The doses from transport are estimated using the generic distribution methodology in

Appendix A.3. The doses from both transport and retail sales are estimated assuming (1) a

quantity of 3

H per lock illuminator of 555 MBq (15 mCi), (2) a leakage rate of 1 ppm/h, which is

the same as that from watches containing luminous paint (NUREG/CR–0216), (3) a breathing

rate of 1 m3

/h, (4) shipments by semi-truck occurring 10 times per year to each of 10,000
2–75

retailers, and (5) two lock illuminators per car. It is assumed that the 3

H is uniformly distributed

within a volume of air.

Using the average dose factor in Table A.3.2, the annual EDE from transport is estimated to be

1×105

 mSv (0.001 mrem) for all individuals involved in transportation and distribution. The

individual annual EDE from retail sales is less than 1×105

 mSv (<0.001 mrem).

2.4.4.1.2 Promethium-147

The doses from both transport and retail sales are estimated assuming (1) a quantity of 147Pm

per lock illuminator of 74 MBq (2 mCi), (2) shipments occurring 10 times per year to each

retailer, and (3) two lock illuminators per car. Using the average dose factor in Table A.3.2, the

annual EDE from transport is estimated to be 8×105

 mSv (0.008 mrem) for all individuals

involved in transportation and distribution. The annual EDE from retail sales is less than

6×105

 mSv (<0.006 mrem).

2.4.4.2 Routine Use

This section presents hypothetical estimates of dose from private and commercial use of

automobiles with lock illuminators containing either 3

H or 147Pm. It is assumed that exposure

occurred to both a driver and passengers while traveling to and from work and while using an

automobile to perform other normal activities during the day.

Automobiles are also employed for commercial use (e.g., taxicabs). Information gathered

indicates that taxicab companies generally buy used cars and keep them for 1 to 2 years before

replacing them (Phone call, U. G. Turner, Manager, Yellow Cab Co., Knoxville, TN, September

1996). The used cars are originally used as private use automobiles (or part of a motor pool

fleet) before being used as taxicabs. It is assumed that automobiles are used as private

transport for 5 years prior to use as a taxicab. As with private use, commercial use also

involves drivers and passengers but for different amounts of time.

2.4.4.2.1 Tritium

Hypothetical doses to an individual resulting from inhalation of 3

H from automobile lock

illuminators were estimated assuming (1) the quantity of 3

H per illuminator is 555 MBq (15 mCi)

(the limit for the exemption), (2) two lock illuminators per car, used on the inside of the vehicle,

(3) a leakage rate of 1 ppm/h (equivalent to that from luminous paint watches;

NUREG/CR–0216), (4) a ventilation rate in an automobile of 5 air volume changes per hour, (5)

an enclosure volume of 6.2 m3

, and (6) a breathing rate of 1 m3/h. The 3H leaking into the

automobile interior is assumed to be uniformly distributed within that volume of air (i.e., driver

and passengers are exposed to equal amounts of 3

H). Table 2.4.2 presents the estimated

individual EDEs. 

For private use, the maximum annual EDE is to an individual during the first year of use and is

estimated to be 5×104

 mSv (0.05 mrem). This is based on 80 min/day (490 h/yr) of time spent

inside an automobile (EPA/600/P-95/002Fa). 

For commercial use, the 3

H in the lock illuminator would have decayed to 415 MBq (11mCi) in

the 5 years the automobile was used as a private vehicle before being used for commercial
2–76

purposes. The maximum annual EDE is to an individual during the first year of commercial use

and is estimated to be 1×103 mSv/yr (0.1 mrem/yr), based on 2000 h/yr driving (e.g., taxi

driver).

2.4.4.2.2 Promethium-147

Hypothetical doses to an individual resulting from external exposure were estimated assuming

(1) the quantity of 147Pm of 74 MBq (2 mCi)—the limit in the exemption, (2) two lock illuminators

per car, and (3) a distance between an illuminator and an exposed individual of 90 cm in the

front seat and 180 cm to the back seat of an automobile. Based on the maximum dose rate

allowed in the exemption (0.01 mGy/h (1 mrad/h) at 1 cm), the inverse square law was used to

determine the EDE at different distances. Table 2.4.3 summarizes the individual EDEs from

147Pm in lock illuminators. The assumptions for 147Pm are the same as for 3

H except that

external exposure is the primary pathway of concern. 

For private use, the maximum annual EDE is to an individual during the first year of use and is

estimated to be 0.001 mSv (0.1 mrem).

For commercial use, the 147Pm in the lock illuminator would have decayed to 19 MBq (0.53 mCi)

in the 5 years the automobile was used as a private vehicle before being used for commercial

purposes. The maximum EDE is during the first year of commercial use and is estimated to be

0.001 mSv/yr (0.1 mrem/yr), based on 2000 h/yr of driving (e.g., taxi driver).

2.4.4.3 Disposal

Generic assumptions and dose-to-source ratios from Appendix A.2 were used to estimate

individual EDEs from the disposal of lock illuminators containing 3

H or 147Pm. These doses are

summarized in Tables 2.4.4 and 2.4.5. It is assumed that the potential fate of lock illuminators

could involve landfill disposal or incineration. Recycling is also considered. According to

Appendix A.2, there are 3500 active landfills and about 150 incinerators. It is assumed that

80% of the lock illuminators are sent to landfills and 20% to incinerators. Alternatively, it is

assumed that 100% of the lock illuminators are recycled. As discussed in Appendix A.2.3.1.5, a

factor of 10 reduction has been applied for the ingestion and inhalation of pathways to account

for the solid form of the lock illuminators and the reduced dispersibility.

2.4.4.3.1 Landfills

In landfill disposal, the main groups of individuals considered for dose assessment are

collectors, operators, off-site residents, and future on-site residents. The hypothetical exposure

to off-site residents involves resuspension of soil during operation activities and atmospheric

dispersion off-site, and drinking water from an off-site municipal well. Future on-site residents

are assumed to be living on a former landfill site.

For 3

H, the maximum annual EDE to an individual from landfill disposal is to a collector and is

estimated to be 2×105

 mSv/yr (0.002 mrem/yr). Drinking water from a municipal well off-site

after a landfill is closed represents an estimated dose less than 1×105

 mSv (<0.001 mrem).

For lock illuminators containing 147Pm, the maximum annual EDE is to the collector and is

estimated to be 1×105

 mSv/yr (0.001 mrem/yr).
2–77

2.4.4.3.2 Incineration

For disposal involving incineration, the two main groups of exposed individuals are incineration

workers and off-site residents exposed to atmospheric releases during operation. The

maximum EDE to an individual from either radionuclide is to a collector. For 3

H, the maximum

hypothetical individual EDE to a collector is estimated to be 1×104

 mSv/yr (0.01 mrem/yr). For

147Pm, the maximum hypothetical annual EDE to a collector is estimated to be 7×105

 mSv

(0.007 mrem).

2.4.4.3.3 Recyle

Since lock illuminators are an integral part of the automobile, recycling is possible. Assuming

the annual recycle of 1,000,000 automobiles containing one lock illuminator at the exemption

level, decayed for 10 years, and using the dose factors from Table A.2.15, the annual EDE to

the slag worker is estimated to be 0.002 mSv (0.2 mrem) for 3

H and 0.002 mSv (0.2 mrem) for

147Pm.

2.4.4.4 Accidents

Of the accident scenarios discussed in the generic accident methodology in Appendix A.1, the

two that are most applicable to the current assessment involve a transport fire and a warehouse

fire. Inhalation is the primary radiological exposure pathway during a fire. A release fraction of

1 (or 100%) is used for 3

H (i.e., gas), and the firefighter is assumed to wear a respirator

providing a protection factor of 1000. A transport fire is assumed to involve one shipment of 10

automobiles, each of which contains one lock illuminator. A warehouse fire is assumed to

involve a manufacturing or distribution facility containing (e.g., for storage) 1000 lock

illuminators.

2.4.4.4.1 Tritium

The EDE to a maximally exposed individual as a result of a transport fire involving 10 lock

illuminators, each containing 555 MBq (15 mCi) of 3

H, is estimated to be 1×104 mSv

(0.01 mrem). For a warehouse fire involving 1000 lock illuminators, the EDE to a maximally

exposed individual (i.e., firefighter) is about 0.002 mSv (0.2 mrem).

2.4.4.4.2 Promethium-147

A transport fire involving 10 lock illuminators containing 74 MBq (2 mCi) of 147Pm could result in

an EDE of 7×106

 mSv (7×104 mrem) to a firefighter, and 8×105 mSv (0.008 mrem) for

cleanup. For a warehouse fire involving 1,000 lock illuminators, the EDE is estimated to be

1×104

 mSv (0.01 mrem) to a firefighter and 8×104 mSv (0.08 mrem) for clean-up.

2.4.4.5 Misuse

It is unlikely that lock illuminators will be misused. If, however, a lock illuminator is removed

from the automobile and worn as costume jewelry, hypothetical exposure can be calculated. An

individual is assumed to wear a 10-year-old automobile lock illuminator as a piece of costume

jewelry for 520 h/yr (NRC, 49 FR 18308). The jewelry is assumed to have the same surface

area as a watch, or 10 cm2

.
2–78

2.4.4.5.1 Tritium

The skin dose and the EDE from 3

H absorbed by the skin were estimated using the

assumptions in Section 2.14.4.2 on skin absorption of 3

H. The original 555 MBq (15 mCi) of 3H

would decay to about 315 MBq (8.5 mCi) in 10 years. A leakage rate of 1 ppm/h was assumed

(equivalent to that from luminous paint watches; NUREG/CR–0216). The estimated dose

assumes 520 h/yr exposure. The annual dose equivalent from tritiated water vapor (HTO) to

the part of the skin in contact with the jewelry is estimated to be about 0.6 mSv/yr (60 mrem/yr). 

The average annual dose equivalent to the skin of the whole body (assuming 10 cm2

 surface

area for the jewelry and 1.8 m2

 for the whole body) is 3×104 mSv/yr (0.03 mrem/yr). The

contribution of the skin dose to the annual EDE is less than 1×105

 mSv (<0.001 mrem), using a

weighting factor for the skin of 0.01. The annual EDE to the internal organs from HTO

absorbed through the skin is 6×104

 mSv/yr (0.06 mrem/yr).

2.4.4.5.2 Promethium-147

It is assumed that the 74 MBq (2 mCi) of 147Pm yields the original dose limit in the exemption of

0.01 mGy/h (1 mrad/h) at 1 cm and that the original amount of 74 MBq (2 mCi) of 147Pm decays

to 4 kBq (0.14 mCi) after 10 years. Using the inverse square law, the maximum EDE from

external exposure at a depth of 10 cm (considered representative of internal organs) would be

0.004 mSv/yr (0.4 mrem/yr) from wearing, 520 h/yr, a piece of jewelry made from a 10-year-old

lock illuminator. This estimate should be conservative since it does not take into account the

considerable attention of the very low-energy photons in transport through 10 cm of tissue. 

The dose to a small area of skin is estimated to be 0.4 mSv/yr (40 mrem/yr), assuming a

nominal 1 cm distance between the 147Pm source in the lock illuminator and the skin. Assuming

a 10 cm2

 exposed area and a skin weighting factor of 0.01, the contribution of this skin dose to

the annual EDE is less than 1×105

 mSv (<0.001 mrem).

2.4.5 Summary 

The present evaluation assesses the hypothetical radiological impacts from 1 million

automobiles with a lock illuminator containing 3

H or 147Pm. Doses involve routine exposure to

the public from distribution, routine use, and disposal. Also included are the dose estimates for

nonroutine exposure from postulated accidents and potential misuse. The results are

summarized in Tables 2.4.6 and 2.4.7.

The highest individual exposures under routine conditions are to drivers during commercial use

(e.g., taxi) for both 3

H and 147Pm. For 3H and 147Pm, the highest individual dose equivalent is

estimated to be 0.001 mSv/yr (0.1 mrem/yr). 
2–79

Table 2.4.1 Hypothetical Doses From Distribution of Automobiles With Lock Illuminators

Containing 3

H or 147Pm a

Exposure Scenario

Maximum Individual Effective Dose Equivalent

(mrem/yr)b

3

H

Transport

 Driver (semi-truck)

 Persons along route

<0.001 

NAd

Distributionc

 Worker <0.001 

147Pm

Transport

 Driver (semi-truck)

 Persons along route

<0.008 

Distributionc

 Worker <0.006 

a

 Assumes there is an average of 10 full-size automobiles per semi-truck trailer during transport

with each retailer receiving 100 automobiles per year (at 10 per shipment; 10 shipments per

year). Each automobile is assumed to have two lock illuminators each containing either

555 MBq (15 mCi) of 3

H or 74 MBq (2 mCi) of 147Pm. Assumes a leakage rate of 1 ppm/h for

3

H.

b

 1 mrem = 0.01 mSv. 

c

 Assumes an automobile dealership showroom as equivalent to a medium warehouse (see

Appendix A.3). Doses may be conservative since the assumptions for a warehouse include

several individuals in closer contact with the product during loading, unloading, and storage

than is likely encountered during sales.

d

 Not applicable.
2–80

Table 2.4.2 Assumptions and Estimated Doses From Use of Automobiles

With Lock Illuminators Containing 3

H a

Exposure Scenariob

Individual Annual Effective Dose

Equivalent

 (mrem)c

Private

 Driver and

 passengers

0.05

Commerciald

 Driver

 Passenger

 (rear seat)

0.1

0.02

a

 Two lock illuminators each assumed to contain 555 MBq (15 mCi) of 3H, with a leakage rate of

1 ppm, uniformly dispersed within the volume of the automobile.

b

 The exposure time for driver and passengers during private use is 490 h/yr. For commercial

use, the driver is exposed for 2000 h/yr while a passenger rides for 250 h/yr.

c

 1 mrem = 0.01 mSv.

d

 Allow 5 years of radioactive decay before automobile is used commercially.
2–81

Table 2.4.3 Assumptions and Estimated Doses From Use of Automobiles

With Lock Illuminators Containing 147Pm a

Exposure Scenariob

Individual Annual Effective Dose

Equivalent

 (mrem)c

Private

 Driver and

 1 passenger

 (at 90 cm)

0.1

Commerciald

 Driver

 Passenger

 (at 180 cm)

0.1

0.004

a Two lock illuminators each assumed to contain 74 MBq (2 mCi) of 147Pm.

b

 The exposure time for driver and passengers during private use is 490 h/yr. For commercial

use, the driver is exposed for 2000 h/yr while a passenger rides 250 h/yr.

c

 1 mrem = 0.01 mSv.

d Allow 5 years of radioactive decay before automobile is used commercially.
2–82

Table 2.4.4 Assumptions and Doses From Exposure to Disposal of

Automobile Lock Illuminators Containing 3

H a, b

Disposal Scenario

Individual Annual Effective

 Dose Equivalent

(mrem)c

Landfill

 Collector

 Operator

 Off-site resident (air) and (groundwater)

 Future on-site resident

0.002

<0.001 

<0.001 

<0.001 

Incineration

 Collector

 Operator

 Off-site resident

0.01

<0.001 

<0.001 

Recycled

Slag Worker

0.2

a

 Assumptions based on the disposal of 1 million lock illuminators per year containing 555 MBq

(15 mCi) of 3

H, 80% by landfill disposal and 20% by incineration. Allow 10 years of radioactive

decay before disposal.

b

 Estimated using the dose-to-source ratios found in Appendix A.2.

c

 1 mrem = 0.01 mSv.

d

 Conservative assumption of 100% recycle of one million automobiles.
2–83

Table 2.4.5 Assumptions and Doses From Exposure to Disposal of

Automobile Lock Illuminators Containing 147Pm a, b

Disposal Scenario

Individual Annual Effective

 Dose Equivalent

(mrem)c

Landfill

 Collector

 Operator

 Off-site resident (air)

 Future on-site resident

0.001

<0.001

<0.001

<0.001

Incineration

 Collector

 Worker

 Off-site resident

0.007

<0.001

<0.001

Recycle d

Slag Worker

0.2

a

 Assumptions based on the disposal of 1 million lock illuminators per year containing 74 MBq

(2 mCi) of 147Pm, 80% by landfill disposal and 20% by incineration. Allow 10 years of

radioactive decay before disposal.

b

 Estimated using the dose-to-source ratios found in Appendix A.2.

c

 1 mrem = 0.01 mSv.

d

 Conservative assumption of 100% recycle of one million automobiles.
2–84

Table 2.4.6 Hypothetical Radiation Doses From Automobile Lock

Illuminators Containing 3

H

Exposure Scenario

Individual Annual Effective

 Dose Equivalent

 (mrem)a

Distributionb

 Transport

 Retail sales

<0.001

<0.001

Routine Use

 Private

 Commercial

0.05

0.1

Disposalc

 Landfill

 Incineration

 Recycle

0.002

0.01

0.2

Accidentsd

 Transport fire

 Warehouse fire

0.01

0.2

Misusee 0.06

a

 1 mrem = 0.01 mSv.

b

 Assumptions based on the generic distribution methodology (Appendix A.3). Retail sales for

automobiles based on the numbers for a medium warehouse as described in Appendix A.3.

c

 Assumptions based on the generic disposal methodology (Appendix A.2). The highest

individual dose equivalent is listed and is to the collector for both landfill disposal and

incineration.

d

 Assumptions based on the generic accident methodology (Appendix A.1). Transport

accidents assume 10 lock illuminators and warehouse accidents involve 1000 lock illuminators.

e

 Assuming an individual wears a 10-year-old lock illuminator as costume jewelry for 520 h/yr. 

The dose is the total effective dose equivalent to the skin and to the internal organs from

absorption of 3

H by the skin.
2–85

Table 2.4.7 Hypothetical Radiation Doses From Automobile Lock

Illuminators Containing 147Pm

Exposure Scenario

Individual Annual Effective 

Dose Equivalent

 (mrem)a

Distributionb

 Transport

 Retail sales

0.008

0.006

Routine Use

 Private

 Commercial

 0.1

 0.1

Disposalc

 Landfill

 Incineration 

 Recycle

 0.001

 0.007

0.2

Accidentsd

 Transport fire

 Warehouse fire

<0.001

 0.08

Misusee 0.4

a

 1 mrem = 0.01 mSv.

b

 Assumptions based on the generic distribution methodology (Appendix A.3). Retail sales for

automobiles based on the numbers for a medium warehouse as described in Appendix A.3.

c

 Assumptions based on the generic disposal methodology (Appendix A. 2). The highest

individual dose equivalent is listed and is to the collector for both landfill disposal and

incineration.

d

 Assumptions based on the generic accident methodology (Appendix A. 1). Transport

accidents assume 10 lock illuminators and warehouse accidents involve 1000 lock illuminators.

e

 Assuming an individual wears a 10-year-old lock illuminator as costume jewelry for 520 h/yr

(NRC, 49 FR 18308). Dose is from external exposure only.

2–87

2.5 Balances of Precision

2.5.1 Introduction

In 10 CFR Part 30.15(a)(3), persons who receive, possess, use, transfer, own, or acquire

balances of precision containing tritium (3

H) are exempted from licensing requirements for

byproduct material, provided that the balances of precision do not contain more than

37 megabecquerel (MBq) (1millicurie (mCi)) of 3

H per balance. This exemption was proposed

on April 8, 1964 (29 FR 4918), and issued as a final rule on August 8, 1964 (29 FR 11445). 

The information in the Federal Register notices on potential radiological impacts on the public

from use of balances of precision containing 3

H is discussed in Section 2.5.3.

2.5.2 Description of Items

The 3

H was to be used as an antistatic device and was to be applied to each of two unexposed

points on the metal parts of each balance. The Federal Register notices cited in Section 2.5.1

do not specifically state how the 3

H is applied to the balances of precision. It is assumed that

the 3

H is contained in self-luminous paint. According to Setra Systems, Inc., and Mettler

Toledo, major manufacturers of precision balances, 3

H is not currently being used on balances

of precision (Phone call, G. Zenoni, Safety and Quality, Setra Systems, Inc., Acton, MA, July

1996; phone call, Sales Support, Mettler Toledo, Hightstown, NJ, June 1996).

2.5.3 Summary of Previous Analyses and Assessments

The Federal Register notices cited above contain the only previously published information on

radiological impacts on the public from use of balances of precision containing 3

H. However,

these notices do not provide information about specific radiological doses from distribution, use,

or disposal. The notice of proposed rulemaking (29 FR 4918) noted that it would be required

that the tritium be so applied as to preclude direct physical contact with it by the users.” It was

also stated that the 3

H was to be applied in such a manner that it would not be released or

removed from the part under normal conditions of use of the balance.”

It is also stated in the proposed rulemaking that, even in the highly unlikely event of ingesting

37 MBq (1 mCi) of 3

H, a person would receive a total radiation dose of only 0.0018 sievert (Sv)

(0.18 rem)” and was compared to the 5 mSv/yr (0.5 rem/yr) dose limit recommended by the

Federal Radiation Council (FRC) (25 FR 4402) and the International Commission on

Radiological Protection (ICRP) (ICRP 2) for individuals members of the public.

2.5.4 Present Exemption Analysis

Although it appears that 3

H is not presently being used in balances of precision, such uses are

allowed under this exemption and could occur in the future. In this hypothetical assessment, it

is assumed that the 3

H is applied as self-luminous paint. Doses are estimated for distribution,

routine use, disposal as solid waste (landfill or incineration), postulated accidents involving fire,

and potential misuse. Collective doses are not determined as this product is not believed to be

currently manufactured or in use. However, for purposes of modeling individual doses during

distribution and transport and for disposal, it is assumed that 10,000 balances per year are

distributed with radioactive self-luminous paint containing 37 MBq (1 mCi) of 3

H, which is the
2–88

limit for the exemption. These 10,000 balances are assumed to be installed in 2,000 new

laboratories (i.e., 5 balances per laboratory).

2.5.4.1 Distribution and Transport

It was not possible to determine specific distribution information on precision balances. Thus,

for 10,000 balances distributed annually, it is assumed that 5 balances are sent to each of

2,000 laboratories. For this assessment, one manufacturer is assumed to distribute 10,000

balances per year. It is assumed that all of the balances are shipped to a medium-sized

warehouse before being sent by air transport to each of the laboratories. The resulting doses

are estimated using the generic distribution methodology and dose factors in Appendix A.3.

Doses to an individual resulting from inhalation of 3

H were estimated assuming (1) a quantity of

3

H per balance of 37 MBq (1 mCi), (2) a leakage rate of 1 ppm/h, which is the same as that

from watches containing luminous paint (NUREG/CR–0216), (3) a breathing rate of 1 m3

/h, and

(4) shipments of five precision balances occurring one time each year to each of 2,000

laboratories. It is assumed that the 3

H is uniformly distributed within a volume of air. The

specific volumes of air assumed for trucks and warehouses are discussed in Appendix A.3. 

Based on assumptions in the generic distribution methodology, the 3

H is more concentrated in

smaller sized facilities (i.e., small trucks, medium warehouses, small retail stores), thus yielding

higher individual doses. Table 2.5.1 includes the individual and collective effective dose

equivalents (EDEs) from distribution.

The highest individual annual EDE during ground transport would be to the driver of a small

truck during regional delivery and is estimated to be less than 1×105

 mSv (<0.001 mrem). The

individual EDE during air transport is also estimated to be less than 1×105

 mSv (<0.001 mrem). 

2.5.4.2 Routine Use

This section includes dose estimates from normal use of balances of precision in a laboratory. 

Doses to an individual resulting from inhalation of 3

H in balances of precision were estimated

assuming (1) a quantity of 3

H per balance of 37 MBq (1 mCi), (2) a leakage rate of 1 ppm/h,

(3) an air ventilation rate in a laboratory of 6 changes per hour, (4) an average air volume in the

laboratory of 180 m3

, (5) a breathing rate of 1.2 m3/h, (6) individual worker exposed for

2000 h/yr, and (7) five balances per laboratory. The 3

H leaking into the laboratory interior is

assumed to be uniformly distributed within that volume of air. The assumed ventilation rate and

laboratory air volume are taken from the generic modeling in Appendix A.1. Individual EDEs

are included in Table 2.5.1. During routine use, the maximum individual EDE is during the first

year of exposure and is estimated to be 1×105

 mSv/yr (0.001 mrem/yr).

2.5.4.3 Disposal

Generic assumptions and dose-to-source ratios from Appendix A.2 were used to estimate

individual and collective EDEs from disposal of balances of precision containing 3

H. These

doses are summarized in Table 2.5.2. It is assumed that the potential fate of balances could

involve landfill disposal or incineration. Assuming that 80% of the products go to active landfills

and 20% to incinerators, a total of 8000 balances of precision would be disposed of in landfills

annually and 2000 would be incinerated. As discussed in Appendix A.2.3.1.5, a factor of 10

reduction has been applied for the ingestion and inhalation pathways to account for the solid
2–89

form of the balances and the reduced dispersibility. A nominal 25-year decay has been

assumed resulting in an activity of 9 MBq (0.24 mCi) per balance.

2.5.4.3.1 Landfills

In landfill disposal, the main groups of individuals considered for dose assessment are

collectors, operators, off-site residents, and future on-site residents. The exposure to off-site

residents involves resuspension of soil during operation and atmospheric dispersion off-site,

and drinking water from an off-site municipal well after landfill closure.

The estimated annual EDEs from landfill disposal to the waste collector, landfill operator,

off-site members of the public, and future on-site residence are all less than 1×105

 mSv

(<0.001 mrem).

2.5.4.3.2 Incineration

During incineration, the main groups of individuals considered for dose assessment are

collectors, workers, and off-site residents. The exposure to off-site residents involves

atmospheric dispersion during operation. The highest annual EDE to an individual during

incineration would be to a collector and is estimated to be less than 1×105

 mSv (<0.001 mrem).

2.5.4.4 Accidents

Of the accident scenarios discussed in the generic accident methodology in Appendix A.1, the

two that are most applicable to the present assessment involve a transportation fire and a

warehouse fire. Inhalation is the primary radiological exposure pathway during a fire. A release

fraction of 1 (or 100%) is used for 3

H (i.e., gas), and the firefighter is assumed to wear

protective clothing and a respirator providing a protection factor of 1000.

A transportation fire is assumed to involve one shipment of five precision balances. Using

factors discussed in Appendix A.1, the EDE to a firefighter would be less than 1×105

 mSv

(<0.001 mrem). For a warehouse fire, it is assumed that all of the 10,000 balances are stored

inside a warehouse during a fire. The individual EDE to a firefighter is estimated to be

0.002 mSv (0.2 mrem).

2.5.4.5 Misuse

A reasonable, misuse scenario is difficult to envision since the 3

H is affixed to a metal surface

and is unlikely to pose either an inhalation or ingestion pathway. However, for an unlikely

scenario regarding misuse, doses are estimated for an individual who removes a 25-year-old

balance and wearing its parts as costume jewelry. An individual is assumed to wear a piece of

costume jewelry for 520 h/yr (NRC, 49 FR 18308). The jewelry is assumed to have the same

surface area as a watch, or 10 cm2

.

The skin dose and the EDE from 3

H absorbed by the skin were estimated using the

assumptions in Section 2.14.4.2 on skin absorption of 3

H. The original 37 MBq (1 mCi) of 3H

would decay to about 9 MBq (0.24 mCi) in 25 years. A leakage rate of 1 ppm/h was assumed

(equivalent to that from luminous paint watches; NUREG/CR–0216). The estimated dose

assumes 520 h/yr exposure. The annual dose equivalent from tritiated water vapor (HTO) to
2–90

the part of the skin in contact with the jewelry is estimated to be about 0.02 mSv/yr (2 mrem/yr). 

The average annual dose equivalent to the skin of the whole body (assuming a 10 cm2 surface

area for the jewelry and 1.8 m2

 for the whole body) is 1×105 mSv (0.001 mrem). The

contribution of the skin dose to the annual EDE is estimated to be less than 1×105

 mSv

(<0.001 mrem), using a weighting factor for the skin of 0.01. The annual EDE to the internal

organs from HTO absorbed through the skin is 2×105

 mSv/yr (0.002 mrem/yr). The total EDE

is, therefore, about 2×105

 mSv/yr (0.002 mrem/yr).

It was presented in the proposed rulemaking (29 FR 4918) that even in the highly unlikely event

of ingesting 37 MBq (1 mCi) of 3

H, a person would receive a total radiation dose of only

1.8 mSv (0.18 rem) compared to the 5 mSv/yr (0.5 rem/yr) dose limit recommended by the

FRC (FRC, 25 FR 4402) and the ICRP (ICRP 2) for individual members of the public. Utilizing

the ingestion dose conversion factors given in Table 2.1-2, the estimated individual EDE is

0.64 mSv (64 mrem), from ingestion of 37 MBq (1 mCi) of 3

H.

2.5.5 Summary 

Tritium is not being used in balances of precision. Even so, this hypothetical assessment of

radiological impacts from balances of precision containing 3

H evaluates potential exposure to

members of the public from distribution, routine use, and disposal. Also, dose estimates for

postulated accidents and potential misuse are included. Each balance initially contains 37 MBq

(1 mCi) of 3

H. The results are summarized in Table 2.5.3.

The highest individual exposures are to workers during normal use. The estimated annual EDE

is 1×105

 mSv (0.001 mrem). 
2–91

Table 2.5.1 Estimated Doses From Distribution and Use of Balances of Precision

Containing 3

H

Exposure Scenario

Individual Annual Effective Dose Equivalent

(mrem)a

DISTRIBUTIONb

To warehouse

 Small truck (regional) <0.001

At warehouse

 Medium <0.001

To laboratory

 Air terminal 

 Airplane 

<0.001

<0.001

ROUTINE USE

Laboratory workerc 0.001

a

 1 mrem = 0.01 mSv.

b

 Effective dose equivalents for distribution are based on the generic distribution methodology

described in Appendix A.3. It is assumed that during distribution balances are delivered from

the manufacturer to a warehouse before being shipped by air to each of 2,000 laboratories (five

balances per laboratory). 

c

 Assumes five balances per laboratory. 
2–92

Table 2.5.2 Estimated Doses From Disposal of Balances of Precision

Containing 3

H a, b

Disposal Scenario

Individual Annual Effective 

Dose Equivalent

 (mrem)c

Landfill

 Collector

 Worker

 Off-site resident

 Future on-site resident

<0.001

<0.001

<0.001

<0.001

Incineration

 Collector

 Worker

 Off-site resident

<0.001

<0.001

<0.001

a

 Assumptions based on Appendix A.2; there are approximately 3,500 active landfills and

150 incinerators.

b

 Assumptions based on the disposal of 10,000 balances of precision per year containing 3H

(allowing 25 years of radioactive decay before disposal).

c

 1 mrem = 0.01 mSv.
2–93

Table 2.5.3 Summary of Radiation Doses From 10,000 Balances of Precision

Containing 3

H 

Exposure Scenario

Individual

Annual Effective Dose Equivalent

(mrem)

Distributionb <0.001

Routine usec 0.001

Disposald

 Landfill

 Incineration

<0.001

<0.001

Accidentse

 Transportation fire

 Warehouse fire

<0.001

0.2

Misusef 0.002

a

 1 mrem = 0.01 mSv.

b

 Assumptions based on the generic distribution methodology (Appendix A.3).

c

 Highest individual dose is for the first year of use with five balances per laboratory. 

d

 Assumptions based on the generic disposal methodology (Appendix A.2). Includes 25 years

of radioactive decay before disposal. 

e

 Assumptions based on the generic accident methodology (Appendix A.1). During the

transportation fire, the firefighter is assumed to be exposed to one shipment of five balances. 

During the warehouse fire, the firefighter is assumed to be exposed to 10,000 balances stored

in a warehouse.

f

 Assumes an individual wears 25-year-old parts from a balance as costume jewelry for

520 h/yr.

2–95

2.6 Automobile Shift Quadrants

2.6.1 Introduction

In 10 CFR Part 30.15(a)(4), persons who receive, possess, use, transfer, own, or acquire

automobile shift quadrants are exempted from licensing requirements for byproduct material,

provided that such shift quadrants do not contain more than 930 megabecquerel (MBq)

(25 millicurie (mCi)) of tritium (3

H). The exemption was proposed on September 17, 1965

(30 FR 11923), and was issued final on April 2, 1966 (31 FR 5315). The information in the

Federal Register notices on potential radiological impacts on the public from use of shift

quadrants containing 3

H is discussed in Section 2.6.3.

2.6.2 Description of Items

Self-luminous paint containing 3

H may be used in automobile shift quadrants so that they could

be read easily in the dark. However, with the development of more sophisticated forms of

illumination (e.g., fiber optics), radioactive self-luminous paints containing 3

H are apparently not

being used (Phone call, United Equipment Accessories, Inc., Waverly, IA, November 1994). It

is believed that automobile shift quadrants containing 3

H are not being manufactured, or have

ever been manufactured, for commercial use. 

2.6.3 Summary of Previous Analyses and Assessments

The Federal Register notices cited in Section 2.6.1 do not provide information about specific

radiological doses to members of the public from use or disposal of shift quadrants containing

3

H. It was emphasized in 30 FR 11923 that this exemption does not involve any new

considerations beyond those involved in 3

H contained in timepieces, lock illuminators, or

balances of precision. The Atomic Energy Commission’s decision to grant the petition was

based on the following:

Requirements for manufacture assumed that the 3

H would not be released under the

most severe conditions likely to be encountered in routine use.

Annual release of 3

H to the environment would be a small fraction of the natural

production rate of 3

H, which provides only a very small annual dose to members of the

public.

Burning of automobiles containing these shift quadrants prior to salvage would probably

result in doses to maximally exposed individuals that are only a small fraction of the

recommended annual limit for members of the public.

2.6.4 Present Exemption Analysis

Although it appears that self-luminous paints containing 3

H are not presently being used in shift

quadrants, such uses are allowed under this exemption and could occur in the future. In this

assessment, hypothetical doses are estimated for distribution of automobiles, routine use

(private and commercial), disposal as solid waste (landfill, incineration and recycle), accidents

involving a fire, and potential misuse. Collective doses were not determined as this product is
2–96

not believed to be currently manufactured or in use. However, for purposes of modeling the

transportation, distribution and disposal doses, it is assumed that 1 million shift quadrants per

year are distributed with radioactive self-luminous paint containing the exemption limit of

930 MBq (25 mCi) of 3

H. This specific quantity of 3H is the limit for the exemption. There is

assumed to be one shift quadrant per automobile.

2.6.4.1 Distribution and Transport

The manufacture, installation, importation, and distribution of shift quadrants are not included in

this exemption (30 FR 3374). The transport of vehicles via truck from the manufacturer to the

retailer and retail sales are evaluated in this assessment. The maximum number of

automobiles per truck trailer is 12 (Phone call, Rice Oldsmobile Dealership, Knoxville, TN,

October 1996). 

For this assessment, it is assumed that a retailer would receive 100 automobiles per year (at 10

per shipment; 10 shipments per year). The doses from transport were estimated using the

generic distribution methodology in Appendix A.3. 

Table 2.6.1 includes the individual effective dose equivalents (EDEs) from the distribution of

1 million automobiles to 10,000 retailers. Because of the nature of automobile distribution

(i.e., virtually no handling of the product itself), it is assumed that the only exposure to a shift

quadrant occurs while the automobile is within the showroom of a dealership during retailing. 

The scenario chosen from Appendix A.3 for retail sales was for a dealership showroom, which

is assumed to be the same as a medium warehouse. The resulting doses for retail sales are

likely to be conservative since the assumptions used in Appendix A.3 include more contact with

the product than routinely occurs at an automobile dealership.

The doses from transport are estimated using the generic distribution methodology in

Appendix A.3. The doses from both transport and retail sales are estimated assuming (1) a

quantity of 3

H per shift quadrants of 930 MBq (25 mCi), (2) a leakage rate of 1 ppm/h, which is

the same as that from watches containing luminous paint (NUREG/CR–0216), (3) a breathing

rate of 1 m3

/h, (4) shipments by semi-truck occurring 10 times per year to each retailer, (5) one

shift quadrant per car, and (6) each driver makes 10 shipments. It is also assumed that the 3

H

is uniformly distributed within a volume of air.

Using the average dose factor in Table A.3.2, the EDE from transport is estimated to be

1×105

 millisievert (mSv) (0.001 mrem) for an express delivery via semi-truck. The EDE from

retail sales is estimated to be 3×105

 mSv (0.003 mrem) assuming 10 cars on display at any

one time.

2.6.4.2 Routine Use

This section presents estimates of dose from private and commercial use of automobiles with

shift quadrants containing 3

H. Of the 1 million automobiles, it is assumed that 80% of them

(800,000) would be employed for private use. In estimating doses over the lifetime of the

product, it is assumed that the average private automobile would be used for 10 years. This is

based on an average automobile accumulating 120,000 miles at 12,000 miles/yr. 

Consideration was given to both the driver and any passengers traveling to and from work as

well as during other normal use. Table 2.6.2 includes the individual EDEs. 
2–97

The manufacturing and installation of shift quadrants containing tritium are subjected to

stringent prototype testing pursuant to the requirements of 10 CFR 32.40, which includes drop

and vibration testing followed by immersion testing. With this level of design and manufacturing

control, direct transfer of tritium to the user under the exemption is not likely. Therefore,

ingestion intake of tritium is not considered in the dose modeling for routine use. 

To assess commercial use (i.e., taxicab), information gathered indicates that taxicab companies

generally buy used cars and keep them for 1 to 2 years before replacing them (Phone call,

U. G. Turner, Manager, Yellow Cab Co., Knoxville, TN, September 1996). The used cars are

originally used as private use automobiles (or part of a motor pool fleet) before being used as

taxicabs. It is assumed that an automobile is used as private transport for 5 years, then a

maximum of 5 more years as a taxicab. As with private use, commercial use also involves

drivers and passengers but for different amounts of time.

Doses to an individual resulting from inhalation of 3

H in automobile shift quadrants were

estimated assuming (1) the quantity of 3

H per shift quadrant is 930 MBq (25 mCi) (the limit for

the exemption), (2) one shift quadrant per car, (3) a leakage rate of 1 ppm/h (equivalent to that

from luminous paint watches; NUREG/CR–0216), (4) an air ventilation rate in an automobile of

5 volume changes per hour, (5) an enclosure volume of 6.2 m3

, and (6) a breathing rate of

1 m3

/h. The 3H leaking into the automobile interior is assumed to be uniformly distributed within

that volume of air (i.e., driver and passengers are exposed to equal amounts of 3

H).

For private use, the potential maximum annual EDE is to an individual during the first year of

use and is estimated to be 4×104

 mSv (0.04 mrem). This is based on 250 h/yr of driving or

riding to and from work and 365 h/yr for other automobile use, and is based on 80 min/day

(490 h/yr) of time spent inside an automobile (EPA/600/P-95/002Fa).

For commercial use, the 3

H in the shift quadrants would have decayed to 705 MBq (19 mCi)

during the 5 years the automobile was used as a private vehicle before being used for

commercial purposes. The maximum annual EDE is to an individual during the first year of

commercial use and is estimated to be 0.001 mSv (0.1 mrem). 

2.6.4.3 Disposal

Generic assumptions and dose-to-source ratios from Appendix A.2 were used to estimate

individual EDEs from the disposal of shift quadrants containing 3

H. These doses are

summarized in Table 2.6.3. It is assumed that the potential fate of shift quadrants could involve

landfill disposal, incineration, or recycle. It is assumed that 80% of the shift quadrants would be

sent to landfills and 20% to incinerators. As discussed in Appendix A.2.3.1.5, a factor of 10

reduction has been applied for the ingestion and inhalation pathways to account for the solid

form of the shift quadrants and the reduced dispersibility. Alternatively, to conservatively bound

the potential recycle dose, it is assumed that 100% of the shift quadrants would be sent for

recycle.

2.6.4.3.1 Landfills

In landfill disposal, the main groups of individuals considered for dose assessment are

collectors, operators, off-site residents, and future on-site residents. The exposure to off-site

residents involves resuspension of soil during operation activities and atmospheric dispersion
2–98

off-site, and drinking water from an off-site municipal well. Future on-site residents are

assumed to be living on a former landfill site.

The potential maximum annual EDE to an individual from landfill disposal is to a collector and is

estimated to be 4×105

 mSv (0.004 mrem). Drinking water from a municipal well off-site after a

landfill is closed could potentially result in an individual annual EDE of about 1×105

 mSv

(0.001 mrem). 

2.6.4.3.2 Incineration

For disposal involving incineration, the two main groups of exposed individuals would be

incineration workers and off-site residents exposed to atmospheric releases during operation. 

The potential maximum EDE to an individual from either radionuclide is to a collector. The

potential maximum annual individual EDE to a collector is estimated to be 2×104

 mSv

(0.02 mrem).

2.6.4.3.3 Recycle

Since automobile shift quadrants are an integral part of the automobile, recycle is highly

probable. Assuming the annual recycle of 1,000,000 automobiles containing one shift quadrant

each at the exemption limit, decayed for 10 years, and using the dose factors from

Table A.2.15, the dose to the slag worker is estimated to be 0.003 mSv/yr (0.3 mrem/yr).

2.6.4.4 Accidents

Of the accident scenarios discussed in the generic accident methodology in Appendix A.1, the

two that are most applicable to the present assessment involve a transportation fire and a

warehouse fire. Inhalation would be the primary radiological exposure pathway during a fire. A

release fraction of 1 (or 100%) is used for 3

H (i.e., gas), and the firefighter is assumed to wear

protective clothing and a respirator providing a protection factor of 1000. A transportation fire is

assumed to involve one shipment of 10 automobiles, each of which contains one shift quadrant. 

A warehouse fire is assumed to involve a manufacturing or distribution facility containing (e.g.,

for storage) 1000 shift quadrants.

The EDE to a maximally exposed individual as a result of a transportation fire involving 10 shift

quadrants, each containing 930 MBq (25 mCi) of 3

H, is estimated to be 2×104 mSv

(0.02 mrem). For a warehouse fire involving 1000 shift quadrants, the EDE to a maximally

exposed individual is about 0.004 mSv (0.4 mrem).

2.6.4.5 Misuse

For an unlikely scenario regarding misuse, doses are estimated for an individual removing a

10-year-old automobile shift quadrant and wearing it as costume jewelry. An individual is

assumed to wear the shift quadrant for 520 h/yr (NRC, 49 FR 18308). The shift quadrant is

assumed to have the same surface area as a watch, or 10 cm2

.

The skin dose and the EDE from 3

H absorbed by the skin is estimated by using the

assumptions in Section 2.14.4.2 on skin absorption of 3

H. The original 930 MBq (25 mCi) of 3H

would decay to about 530 MBq (14 mCi) in 10 years. A leakage rate of 1 ppm/h is assumed
2–99

(equivalent to that from luminous paint watches; NUREG/CR–0216). The estimated dose

assumes an exposure period of 520 h/yr (NRC, 49 FR 18308). The annual dose equivalent

from tritiated water vapor (HTO) to the part of the skin in contact with the shift quadrant is

estimated to be about 0.1 mSv (100 mrem). The average annual dose equivalent to the skin of

the whole body (assuming 10 cm2

 surface area for the shift quadrant and 1.8 m2 for the whole

body) is 6×104

 mSv (0.06 mrem). The contribution of the skin dose to the annual EDE is

estimated to be less than 1×105

 mSv (<0.001 mrem), using a weighting factor for skin of 0.01. 

The annual EDE to the internal organs from HTO absorbed through the skin is 0.001 mSv

(0.1 mrem). The total annual EDE to the wearer is the sum of the contribution of the skin dose

to the annual EDE less than 1×105

 mSv (<0.001 mrem)) and the annual EDE to internal organs

from HTO absorbed through the skin (0.001 mSv (0.1 mrem)). The total annual EDE is,

therefore, about 0.001 mSv (0.1 mrem).

2.6.5 Summary 

The present evaluation assesses radiological impacts from 1 million automobiles with a shift

quadrant containing 3

H. Doses involve routine exposure to members of the public from

distribution, routine use, and disposal. Also included are dose estimates for nonroutine

exposure from postulated accidents and potential misuse. The results of this assessment are

based on 1 million shift quadrants containing 3

H. The results are summarized in Table 2.6.4.

The highest potential individual exposures are to drivers during commercial use (e.g., taxicab). 

The highest potential individual dose equivalent is estimated to be 0.001 mSv/yr (0.1 mrem/yr).
2–100

Table 2.6.1 Estimated Doses From Distribution of 1 Million Automobiles With Shift

Quadrants Containing 3

H a, b

Exposure Scenario

Individual Annual Effective 

Dose Equivalent

(mrem)c

Transport

 Driver (semi-truck)

 Persons along route

0.001

NAe

Distributiond

 Worker 0.003

a

 Assumes there is an average of 10 full-size automobiles per semi-truck trailer during

transport. Each driver makes 10 shipments. Each automobile is assumed to have one shift

quadrant.

b

 Each shift quadrants containing 930 MBq (25 mCi) of 3H. Assumes a leakage rate of 1 ppm/h. 

Effective dose equivalents for distribution is based on the generic distribution methodology

described in Appendix A.3.

c

 1 mrem = 0.01 mSv.

d

 Assumes an automobile dealership showroom as equivalent to a medium warehouse (see

Appendix A.3). Doses may be conservative since the assumptions for a warehouse include

several individuals in closer contact with the product during loading, unloading, and storage

than is likely to be encountered during sales.

e

 Not applicable.
2–101

Table 2.6.2 Assumptions and Estimated Doses From Use of 1 Million Automobiles

With Shift Quadrants Containing 3

H a

Exposure Scenariob

Individual Annual Effective Dose

Equivalent

 (mrem)c

Private

 Driver and

 3 passengers

0.04

Commerciald

 Driver

 Passenger

 (rear seat)

0.1

0.02

a

 Each shift quadrant assumed to contain 930 MBq (25 mCi) 3H, with a leakage rate of 1 ppm,

uniformly dispersed within the volume of the automobile.

b

 The exposure time for drivers and passengers during private use is 490 h/yr. For commercial

use, the driver is exposed for 2,000 h/yr while a passenger rides 250 h/yr.

c

 1 mrem = 0.01 mSv.

d

 Allow 5 years of radioactive decay before automobile is used commercially.
2–102

Table 2.6.3 Assumptions and Doses From Exposure to Disposal of 1 Million 

Automobile Shift Quadrants Containing 3

H a, b

Disposal Scenario

Individual Annual Effective

 Dose Equivalent

(mrem)c

Landfill

 Collector

 Worker

 Off-site resident

 - Air

 - Groundwater

 Future on-site resident

0.004

<0.001

<0.001

0.001

<0.001

Incineration

 Collector

 Worker

 Off-site resident

0.02

<0.001

<0.001

Recycled

 Slag Worker

0.3

a

 Assumptions based on the disposal of 1 million shift quadrants per year containing 930 MBq

(25 mCi) of 3

H, 80% by landfill disposal and 20% by inceration.

b

 Allowing for 10 years of radioactive decay before disposal.

c

 1 mrem = 0.01 mSv.

d

 Conservative assumption of 100% recycle of one million automobiles. Estimated using the

dose-to-source ratios found in Appendix A.2. Doses for recycle would be in lieu of doses due to

landfill and incineration.
2–103

Table 2.6.4 Potential Radiation Doses From 1 Million Automobile Shift Quadrants

Containing 3

H

Exposure Scenario

Individual Annual Effective Dose

Equivalent

 (mrem)a

Distributionb

 Transport

 Retail sales

<0.001

0.003

Routine Use

 Private

 Commercial

0.04

 0.1

Disposalc

 Landfill

 Incineration

 Recycle

0.004

0.02

0.3

Accidentsd

 Transportation fire

 Warehouse fire

0.02

0.4

Misusee 0.1

a

 1 mrem = 0.01 mSv.

b

 Assumptions based on the generic distribution methodology (Appendix A.3). Retail sales for

automobiles based on the numbers for a medium warehouse, as described in Appendix A.3.

c

 Assumptions based on the generic disposal methodology (Appendix A.2). Only the highest

doses are listed. Doses from recycle would be in lieu of doses due to landfill and incineration.

d

 Assumptions based on the generic accident methodology (Appendix A.1). Transportation

accidents assume 10 shift quadrants and warehouse accidents involve 1,000 shift quadrants.

e

 Assuming an individual wears a 10-year-old shift quadrant as costume jewelry for 520 h/yr

(NRC, 49 FR 18308). The dose is the total dose equivalent to the skin and to the internal

organs from absorption of 3

H by the skin.

2–105

2.7 Marine Compasses and Navigational Instruments

2.7.1 Introduction

In 10 CFR Part 30.15(a)(5), marine compasses and other marine navigational instruments

containing tritium (3

H) gas are exempted from licensing requirements for byproduct material,

provided that the quantity of 3

H does not exceed 28 gigabecquerel (GBq) (750 millicurie (mCi))

in a compass and 9.3 GBq (250 mCi) in another navigational instrument. The exemption for

marine compasses was proposed on September 17, 1965 (30 FR 11923), and was issued as a

final rule on April 2, 1966 (31 FR 5315). The exemption for other navigational instruments was

proposed on September 14, 1966 (31 FR 12023), and was issued as a final rule on January 24,

1967 (32 FR 785). This second rulemaking specified that the 3

H be in gaseous form.

2.7.2 Description of Items

Marine compasses and other navigational instruments are intended for use in marine vessels. 

Although detailed descriptions of tritium-containing instruments are unavailable, such

instruments should be similar to those that do not contain 3

H. These devices, which usually are

large and panel mounted, typically consist of a rigid case with a transparent face that protects

the working components of the device. Tritium gas is contained in quartz or glass tubes that

are affixed to the readout components of the device. The maximum amounts of 3

H allowed are

28 GBq (750 mCi) in an exempt marine compass and 9.3 GBq (250 mCi) in other exempt

navigational instruments. One manufacturer reports using much less than the maximum

exempt quantities.

No recent distribution and use information specific to marine compasses and other navigational

instruments that contain 3

H is available. Apparently, domestic manufacture and import of these

devices has ceased. This appearance is supported by telephone conversations with

representatives of the largest domestic suppliers of marine instruments. Only one domestic

distributor reported recent distribution of a few thousand marine compasses containing much

less 3

H than is allowed by the exemption. However, based on the small quantity of 3H in each

compass, it is likely, though not certain, that these compasses are exempt self-luminous

products similar to the handheld compasses often used by recreational sailors.

Lacking recent distribution data, estimates of potential consequences of this exemption are

made using reported quantities of 3

H distributed in these devices from 1980 through 1989. 

During that period, Nuclear Regulatory Commission records indicate that the quantity of 3

H

distributed annually ranged between about 1,850 and 3,330 GBq (50 and 90 Ci). These

quantities of 3

H would correspond to an annual manufacture of 67 to 120 compasses or 200 to

360 navigational instruments containing the maximum allowable quantities of 3

H. Although not

believed to be currently manufactured, a reasonable estimate of potential distribution is 100

marine compasses, each containing 28 GBq (750 mCi) of 3

H, plus 200 other marine

navigational instruments, each containing 9.3 GBq (250 mCi) of 3

H.

2.7.3 Summary of Previous Assessments

The Federal Register notices from 1965 and 1966 cited above do not provide specific

information on analyses of radiological impacts on the public from the distribution, use, and
2–106

disposal of marine compasses and other navigational instruments containing 3

H. However, the

first notice concerning marine compasses from 1965 emphasized that this exemption does not

involve any new considerations beyond those involved in the exemptions for timepieces,

automobile lock illuminators, and balances of precision containing 3

H (see Sections 2.3, 2.4,

and 2.5), except for the small additional amounts of 3

H that would eventually be released to the

environment. Thus, the decision by the Atomic Energy Commission (AEC) to grant the petition

for exemption was apparently based on the following factors:

Requirements for manufacture of marine compasses ensure that 3

H will not be released

in significant amounts under the most severe conditions likely to be encountered in

normal use and handling.

Eventual annual releases of 3H to the environment will be a small fraction of the natural

production rate of 3

H, and naturally occurring 3H results in very small annual dose

equivalents to average individuals in the public.

The only known published information on radiological impacts on the public from use and

disposal of other marine navigational instruments containing 3

H is found in the Federal Register

notice from September 1966 cited above. On the basis of the following arguments, the AEC

concluded there does not appear to be any significant hazard associated with the possession

and use of these instruments.

So long as 3

H gas is confined in the glass capsule of an instrument, 3H is not available

for uptake into the body.

The low-energy beta particles emitted by 3

H are absorbed by the walls of the glass

capsules; therefore, there is no external radiation hazard (31 FR 12024).

In the event of severe damage to an instrument, the glass capsule might be broken and

3

H gas dispersed into the air. However, instruments normally are placed in locations

subject to considerable ventilation, and less than 0.1% of 3

H gas inhaled into the lungs

is retained in the body. For example, if 9 GBq (250 mCi) of 3

H gas, which is the limit for

the exemption, were released into a volume of 10 m3

 with a ventilation rate of 10 air

changes per hour, the dose equivalent to an individual would not exceed 0.02 millisievert

(mSv) (2 mrem), or about 2% of the annual dose equivalent from exposure to all

sources of natural background radiation.

If an annual production of 100,000 navigational instruments is assumed, which probably

is unrealistically high, and each instrument contained 9 GBq (250 mCi) of 3

H gas, which

is the limit for the exemption, the total annual release of 3

H to the environment

eventually could reach 930 TBq (25,000 Ci). However, this amount of 3

H is much less

than the annual natural production rate of 0.3 EBq (8 MCi), and the annual dose

equivalent to an average individual from naturally occurring 3

H is only about 0.03 Sv

(3 rem).
2–107

2.7.4 Present Exemption Analysis

2.7.4.1 General Information

Even though it is not believed that marine compasses containing tritium are currently being

manufactured, the present analysis is based on the hypothetical manufacture, distribution, use,

and disposal of 100 compasses, each containing 28 GBq (750 mCi) of 3

H, and 200 navigational

instruments, each containing 9.3 GBq (250 mCi). Thus, the annual distribution of 3

H is taken to

be 4.6 TBq (125 Ci). Consistent with the analysis in other sections where products are not

currently produced, collective doses are not evaluated.

The only credible exposure mechanisms are inhalation and absorption through the skin of 3

H

that has escaped from the devices. The rate of 3

H escape from the devices is taken to be

10 ppb/h (see Section 2.14.4). Thus, the initial rate of 3

H release is 280 Bq/h (7.5 nCi/h) from

compasses and 93 Bq/h (2.5 nCi/h) from other instruments. Because of catalytic interactions

with instrument housing materials, 3

H is assumed to be released as tritiated water vapor (HTO).

2.7.4.2 Distribution

Distribution of marine compasses and other marine navigational instruments includes transport

and associated handling of the devices during movement from a manufacturer or initial

distributor to a facility at which the devices are installed in a marine vessel. Given the small

numbers of devices that may be distributed during a year, one licensed facility (initial distributor)

is assumed to ship 20 compasses and 40 other navigational instruments to each of five facilities

that install the devices into marine vessels. All devices are transported via one small

express-delivery truck to one parcel delivery center (large warehouse). From there, five large

regional-delivery trucks each transport 20 compasses and 40 other instruments to one of five

regional truck terminals (medium warehouses). From each of these terminals, a small localdelivery truck carries the devices to an installer’s facility.

The generic distribution methodology, adjusted to reflect the quantities of 3

H present in each

facility and a leak rate of 10 ppb/h, was used to estimate effective dose equivalents (EDEs) to

distribution workers and the affected members of the public. The highest potential individual

EDE, 3×105

 mSv (0.003 mrem), would be to the driver of the small express-delivery truck that

transports devices between the initial distributor and the first parcel delivery terminal.

Workers at each installation facility are assumed to be exposed to 3

H leaking from the

instruments while being in a building containing the instruments and while installing instruments

in the pilot houses of 20 marine vessels.

Workers are assumed to be in the storage/work building for 2000 h/yr. The storage/work

building has a volume of 640 m3

, has an air ventilation rate of 5 volumes per hour, and contains,

on average, 10 compasses and 20 other instruments. Each worker could receive an EDE of

9×105

 mSv (0.009 mrem).

A crew of three instrument specialists is assumed to fit the pilot houses of 20 marine vessels

per year (10 per crew). A crew is assumed to spend 80 hours in each pilot house, or 800 h/yr,

with a breathing rate of 1.2 m3

/h. A pilot house has a volume of 45 m3, has an air ventilation
2–108

rate of 5 volume changes per hour during installation activities, and contains one

tritium-containing compass and two other tritium-containing navigational instruments. Each

worker could potentially receive an EDE of 5×105

 mSv (0.005 mrem).

2.7.4.3 Routine Use

The most exposed group of persons during use of marine compasses and navigational

instruments should be the navigation crew of a marine vessel. This crew mans the pilot house

of the vessel. Potential EDEs to these crew members are estimated by assuming that: (1) one

compass and two navigational instruments are present in a ship’s pilot house, (2) 3

H emanates

from these instruments as HTO at a rate of 10 ppb/h, (3) the volume of the pilot house is 45 m3

,

(4) the air ventilation rate is 2 volume changes per hour, and (5) the individual is in the pilot

house for 2000 h/yr. Under these conditions, an individual could receive an EDE of about

3×10-4 mSv (0.03 mrem) during the first year of use. 

2.7.4.4 Accidents and Misuse

The potentially most serious and credible accident involving a marine compass or navigational

instrument is breakage in the pilot house. To illustrate the potential radiation doses associated

with such an event, it is assumed there is immediate dispersal of 28 GBq (750 mCi) 3

H into a

45 m3

 pilot house with an air ventilation rate of 2 volume changes per hour. In this case, only

about 1% (0.28 GBq (7.5 mCi)) of the 3

H should be in the form of tritiated water vapor (see

Appendix A.1, Section A.1.7). The maximum possible EDE could be about 0.1 mSv (10 mrem)

if breakage occurs during the first year.

It is difficult to visualize a credible misuse situation for marine compasses and navigational

instruments.

2.7.4.5 Disposal

Obsolete marine compasses and navigational instruments and their parts may be refurbished

for reuse or may be discarded as trash. Discarded instruments and parts likely would be

deposited in a landfill. However, disposal of tritium-containing parts via incineration is possible.

After 20 years, the original inventory (4.6 TBq (125 Ci)) of 3

H will have decayed to about

1.5 TBq (40 Ci). As discussed in Appendix A.2.3.1.5, a factor of 10 reduction has been applied

to the ingestion and inhalation pathways to account for the reduced dispersibility of the 3

H in

glass capsules.

Using the generic disposal methodology for release and dispersal of 3

H during landfill and

incinerator operations, the maximum individual EDE is estimated to be less than 1×105

 mSv

(<0.001 mrem).

2.7.5 Summary 

Table 2.7.1 presents the results of the present analysis of the radiological impacts on the public

from the distribution, use, and disposal of marine compasses and navigational instruments that

contain gaseous 3

H. For distribution and installation, highest potential individual EDEs were
2–109

estimated to be 3×105

 mSv (0.003 mrem) and 9×105 mSv (0.009 mrem), respectively. For

routine use, the maximum individual EDE was estimated to be 3×104

 mSv (0.03 mrem). For

disposal, the maximum individual EDE was estimated to be less than 1×105

 mSv

(<0.001 mrem). For an accident involving breakage of a compass, the maximum EDE was

estimated to 0.1 mSv (10 mrem).

Based on the information published by the AEC and the present analysis, it is concluded that

the radiological impacts on the public from use and disposal of marine compasses and other

marine navigational instruments containing 3

H are very small. This conclusion is supported by

the calculated EDEs and the fact that conservative exposure assumptions were used unless

clear evidence to the contrary was available. This conservatism may be especially true for the

assumption that 3

H gas escaping from the devices is completely oxidized, except for the

accident scenario, and for the various ventilation rates used for vehicles, buildings, and pilot

houses.
2–110

Table 2.7.1 Potential Annual Radiation Doses From Marine Compasses and

Navigational Instruments

Exposure Pathway

Individual Annual Effective Dose

Equivalent

(mrem)a

Distribution 0.003

Installation 0.009

Routine use 0.03

Accident involving

 breakage

10

Disposal <0.001

 a

 1 mrem = 0.01 mSv.
2–111

2.8 Thermostat Dials and Pointers Containing Tritium

2.8.1 Introduction

In 10 CFR Part 30.15(a)(6), persons who receive, possess, use, transfer, own, or acquire

thermostat dials and pointers containing tritium (3

H) are exempted from licensing requirements

for byproduct material, provided that the thermostat dials and pointers do not contain more than

930 megabecquerel (MBq) (25 millicurie (mCi)) of 3

H per thermostat. This exemption was

proposed on September 17, 1965 (30 FR 11923), and issued as a final rule on April 2, 1966

(31 FR 5315). The information in the Federal Register notices on potential radiological impacts

on the public from use of thermostat dials and pointers containing 3

H is discussed in Section

2.8.3.

2.8.2 Description of Items

Self-luminous paint containing 3

H was proposed to be applied to thermostat dials and pointers

so they could be read easily in the dark. However, according to Honeywell, Inc., a major

manufacturer of thermostat housing, tritiated paint is not currently being used on thermostat

dials and pointers, primarily because electronic displays are now available for illumination

purposes (Phone call, J. Phillips, Liaison Engineer, Honeywell, Inc., Minneapolis, MN, January

1995). Neither are gaseous 3

H light sources used for thermostat dials or pointers. 

2.8.3 Summary of Previous Analyses and Assessments

The Federal Register notices cited in Section 2.8.1 contain the only previously published

information on radiological impacts on the public from use of thermostat dials and pointers

containing 3

H. However, these notices do not provide information about specific radiological

doses from distribution, installation, use, or disposal. The notice of proposed rulemaking

pointed out that detailed safety analyses of the use of 3

H were published in conjunction with

notices of proposed rulemaking to exempt timepieces (25 FR 6302), automobile lock

illuminators (26 FR 3571, 10472), and precision balances (29 FR 4918), and to generally

license aircraft safety devices (26 FR 8522). The notice of proposed rulemaking also noted that

it would be unlikely that releases of 3

H from these and other consumer products would

compare with the yearly production of several millions of curies from cosmic rays, an amount

that accounts for less than one-hundred-thousandth of the total dose rate from all natural

sources of radiation.”

2.8.4 Present Exemption Analysis

There is no indication that self-luminous paints containing 3

H are currently manufactured or in

wide-scale use in thermostat dials and pointers. Thus, in this assessment, hypothetical doses

are estimated for distribution, installation and service, routine use, disposal as solid waste

(landfill or incineration), accidents involving a fire, and potential misuse. For purposes of

modeling the transportation, distribution, and disposal doses, it is assumed that 10,000

thermostats per year are distributed with radioactive self-luminous paint containing 930 MBq

(25 mCi) of 3

H, which is the limit for the exemption. These 10,000 thermostats are assumed to

be installed in 5,000 homes (i.e., 2 thermostats per home). If gaseous 3

H light sources were to

be similarly used, the potential doses would be considerably less; quantitative estimates are not
2–112

developed here. Collective doses were not determined as this product is not believed to be

currently manufactured or in use.

2.8.4.1 Distribution and Transport

Based on distribution information provided by Honeywell, 1% of all thermostats are assumed to

be sent to retail stores, 16% to heating and air-conditioning unit manufacturers, and 83% to

wholesalers (Phone call, M. Schlener, Sales, Honeywell, Inc., Minneapolis, MN, December

1996). Thus, for 10,000 thermostats distributed annually, the following scenarios are assumed:

(1) 100 thermostats are sent directly to retail stores, (2) 1,600 are sent to heating and airconditioning unit manufacturers, and (3) the remaining 8,300 are distributed to wholesalers who

sell directly to installation companies.

For this assessment, one manufacturer is assumed to distribute 10,000 thermostats per year,

and the resulting doses are estimated using the generic distribution methodology in Appendix

A.3. Doses to an individual resulting from inhalation of 3

H were estimated assuming (1) a

quantity of 3

H per thermostat of 930 MBq (25 mCi), (2) a leakage rate of 1 ppm/h, which is the

same as that from watches containing luminous paint (NUREG/CR–0216), (3) a breathing rate

of 1 m3

/h, (4) shipments occurring 5 times per year to retail stores and 10 times a year to both

heating and air-conditioning unit manufacturers and wholesalers, and (5) 10 thermostats per

box. It is assumed 3

H is uniformly distributed within a volume of air. The specific volumes of air

assumed for trucks, retail stores, and warehouses are discussed in Appendix A.3.

Based on assumptions made in the generic distribution methodology in Appendix A.3, the 3

H is

more concentrated in smaller sized facilities (i.e., small trucks, medium warehouses, small retail

stores), thus yielding higher individual doses. Table 2.8.1 includes the individual effective dose

equivalents (EDEs) from three distribution scenarios. The exposure scenarios and dose

estimates are described in the following paragraphs:

Scenario I. This scenario involves distribution of 100 thermostats directly from one

manufacturer to two retail stores (i.e., 50 thermostats per store per year). Delivery is assumed

to be express (i.e., nonstop) in either a large or a small truck. Each retail store receives five

shipments each year, each containing one box. Truck drivers and clerks within each store

would be exposed to all 50 of the thermostats, while members of the public are assumed to be

exposed to one box of 10 thermostats assumed to be in stock at all times.

Scenario II. This scenario involves distribution of 1600 of the thermostats directly from one

manufacturer to two heating and air-conditioning unit manufacturers (i.e., 800 thermostats per

company). Delivery is assumed to occur 10 times each year and by express delivery in either a

large or a small truck. Each company receives a total of 80 thermostats (eight boxes) during

each shipment. Truck drivers and workers would be exposed to 800 thermostats each year.

Scenario III. In this scenario, the manufacturer sends 8,300 of the thermostats to wholesalers

who, in turn, distribute the thermostats to installation companies. For easy calculation, it is

assumed that 800 thermostats are sent to each of 10 wholesalers who, in turn, distribute 200

thermostats to each of four installation companies. The individual dose equivalents estimated

during distribution from manufacturer to wholesaler are the same as those in Scenario II. 

However, distribution from a wholesaler to each of four installation companies involves regional

delivery in a small truck. 
2–113

The highest potential individual annual EDE is estimated to be 8×104

 millisievert (mSv)

(0.08 mrem) and occurs to a worker at a small retail store.

2.8.4.2 Installation, Servicing and Routine Use

This section includes dose estimates from the installation, servicing, and routine use of

thermostats in private residences. Doses to an individual resulting from inhalation of 3

H in

thermostat dials and pointers installed and used in a private residence were estimated

assuming (1) a quantity of 3

H per thermostat of 930 MBq (25 mCi), (2) a leakage rate of

1 ppm/h, the same as that from watches containing luminous paint (NUREG/CR–0216), (3) a

ventilation rate in a residence of 1 air volume per hour, (4) an average air volume in the

residence of 450 m3

, (5) a breathing rate of 1 m3/h, and (6) two thermostats per home. Using

statistics on the mean number of minutes per day that an individual spends indoors at home

(EPA/600/P–95/002Fa), it is estimated that individual spends an average of about 6100 h/yr

(70%) indoors in a residence. The 3

H leaking into the home interior is assumed to be uniformly

distributed within that volume of air. Individual EDEs are summarized in Table 2.8.2.

2.8.4.2.1 Installation and Service

In this assessment, a maximum number of 50 thermostats is assumed to be handled by each

installer or serviceman per year. Both installation and one-time service require 30 minutes per

thermostat. During installation, it is assumed there is no leakage of 3

H into the home prior to

installation (i.e., the thermostats are new). The potential individual annual EDE to an installer

exposed to 50 thermostats is less than 1×105

 mSv (<0.001 mrem), based on an installation

time of 30 minutes per thermostat (Phone call, P. Murphy, R&M Climate Control, Knoxville, TN,

December 1996).

It is unlikely a thermostat would require routine maintenance. However, it is possible a

thermostat might require an adjustment after installation. It is assumed each thermostat is

serviced only once, after the first year of use. As a result, leakage of 3

H from existing

thermostats in the home is for 1 year at the time of service. The potential individual EDE to an

individual servicing 50 thermostats per year is less than 1×105

 mSv (<0.001 mrem). 

 2.8.4.2.2 Routine Use

During routine use, the maximum individual annual EDE is during the first year of exposure and

is estimated to be 6×104

 mSv (0.06 mrem). 

2.8.4.3 Disposal

Generic assumptions and dose-to-source ratios from Appendix A.2 were used to estimate

individual EDEs from disposal of thermostat dials and pointers containing 3

H. It is assumed the

potential fate of thermostat dials and pointers could involve landfill disposal or incineration. 

Recycling would be analogous to incineration, because 3

H would likely become volatile during

metal processing. Assuming that 80% of the products go to active landfills and 20% to

incinerators, a total of 8,000 thermostat dials and pointers would be disposed of in landfills

annually and 2,000 would be incinerated. As discussed in Appendix A.2.3.1.5, a factor-of-10

reduction in the ingestion and inhalation pathways has been assured because of the reduced

dispersibility of 3

H in chemical form for the dials and pointers.
2–114

2.8.4.3.1 Landfills

In landfill disposal, the main groups of individuals considered for dose assessment are

collectors, operators, off-site residents, and future on-site residents. The exposure to off-site

residents involves resuspension of soil during operation and atmospheric dispersion off-site,

and drinking water from an off-site municipal well after landfill closure. 

The highest estimated annual EDE would be to a collector and is less than 1×105

 mSv

(<0.001 mrem). 

2.8.4.3.2 Incineration

For disposal involving incineration, the two main groups of exposed individuals are collectors

and off-site residents exposed to atmospheric releases during operation. The highest annual

EDE to individuals is to collectors and is estimated to be less than 1×105

 mSv (<0.001 mrem).

2.8.4.4 Accidents

Of the accident scenarios discussed in the generic accident methodology in Appendix A.1, the

two most applicable to the present assessment involve a transportation fire and a warehouse

fire. Inhalation is the primary radiological exposure pathway during a fire. A release fraction of

1 (or 100%) is used for 3

H (i.e., gas), and the firefighter is assumed to wear protective clothing

and a respirator providing a protection factor of 1000.

A transportation fire is assumed to involve one shipment of 80 thermostats. Using factors

discussed in Appendix A.1, the EDE to a firefighter would be about 0.002 mSv (0.2 mrem). For

a warehouse fire, 1000 thermostats are assumed to be stored inside a warehouse. The

individual EDE to a firefighter is estimated to be 0.004 mSv (0.4 mrem).

2.8.4.5 Misuse

For an unlikely scenario regarding misuse, doses are estimated for an individual who removes

a 20-year-old thermostat dial or pointer and wears it as costume jewelry for 520 h/yr

(NRC, 49 FR 18308). The thermostat dial or pointer is assumed to have the same surface area

as a watch, 10 cm2

.

The skin dose and the EDE from 3

H absorbed by the skin is estimated by using the

assumptions in Section 2.14.4.2 on skin absorption of 3

H. The original 930 MBq (25 mCi) of 3H

would decay to about 300 MBq (8 mCi) in 20 years. A leakage rate of 1 ppm/h is also assumed

(equivalent to that from luminous paint watches; NUREG/CR–0216). The estimated dose

assumes an exposure period of 520 h/yr. The annual dose equivalent from tritiated water vapor

(HTO) to the part of the skin in contact with the jewelry is estimated to be about 0.6 mSv

(60 mrem). The potential annual dose equivalent to the skin of the whole body (assuming

10 cm2

 surface area for the thermostat dial or pointer and 1.8 m2 for the whole body) is

3×104

 mSv (0.03 mrem). The contribution of the skin dose to the annual EDE is estimated to

be less than 1×105

 mSv (<0.001 mrem), using a weighting factor for skin of 0.01. The annual

EDE to the internal organs from HTO absorbed through the skin is estimated to be 5×104

 mSv

(0.05 mrem). The potential annual EDE to the wearer is the sum of the contribution of the skin

dose to the annual EDE (less than 1×105

 mSv (<0.001 mrem)) and the annual EDE to internal
2–115

organs from HTO absorbed through the skin (5×104

 mSv (0.05 mrem)). Therefore, the total

annual EDE would be about 5×104

 mSv (0.05 mrem).

2.8.5 Summary 

The present assessment of radiological impacts from thermostat dials and pointers containing

3

H includes exposure to members of the public from distribution, installation and service, routine

use, and disposal. Also included are the dose estimates for postulated accidents and potential

misuse. The results of this assessment are based on an annual distribution of 10,000

thermostat dials and pointers containing 3

H in radioluminescent paint. Each residence is

assumed to contain 2 dials or pointers. Each thermostat dial or pointer initially contains

930 MBq (25 mCi) of 3

H. The results are summarized in Table 2.8.3.

The highest individual exposures would be to workers inside a small retail store during

distribution and to residents during normal use. Each of these individuals is estimated to

receive an annual EDE of less than 0.001 mSv (<0.1 mrem).
2–116

Table 2.8.1 Estimated Doses From Distribution of 10,000 Thermostat

Dials and Pointers Containing 3

H a

Exposure Scenario

Total Number of

Thermostats

Distributed

Annuallyb

Individual Annual

Effective Dose

Equivalent

(mrem)c

TRANSPORT

Driver

 To retailer - large truck

 small truck

 50

 50

0.002

0.003

 To heating unit co. - large truck

 small truck 

800

800

0.02

0.04

 To wholesaler - large truck

 small truck

200

200

0.02

0.04

 To installation co. - large truck

 small truck

200

200

0.03

0.06

DURING STORAGE

Worker

 At retailer - large store

 small store

 50

 50

0.02

0.08

 At heating unit co. - large warehouse

 small warehouse

800

800

0.02

0.02

 At Wholesaler - large warehouse

 medium warehouse

200

200

0.02

0.02

 At installation co. - medium warehouse 

 

200 <0.001

a

 Distribution of 10,000 thermostat dials and pointers, each containing 930 MBq (25 mCi) of 3H. 

Assumes a leakage rate of 1 ppm/h, uniformly dispersed within the volume of the truck,

warehouse, or store. Effective dose equivalents for distribution based on generic distribution

methodology described in Appendix A.3.

b

 It is assumed that during distribution, thermostats are delivered from the manufacturer to the

various outlets approximately 10 times per year in boxes containing 10 thermostats each.

c

 1 mrem = 0.01 mSv.
2–117

Table 2.8.2 Estimated Doses From Installation, Servicing, and Routine Use of 

10,000 Thermostat Dials and Pointers Containing 3

H a

Exposure Scenario

Annual Number of

Thermostats per Person

Maximum Annual

Individual Effective Dose

Equivalent

(mrem)b

Installationc 50 <0.001

Serviced 50 <0.001

Routine usee 2 0.06

a

 Assumptions based on 10,000 thermostat dials and pointers distributed and used each year.

b

 1 mrem = 0.01 mSv.

c

 Individual spends 30 minutes per thermostat for installation. During installation, thermostat is

new (i.e., no prior leakage of 3

H into the air volume of the residence).

d

 Individual spends 30 minutes per thermostat for service. Assumes that servicing occurs after

1 year (i.e., 1 year of leakage into the air volume) and is only required once in the 20-year

lifetime of the thermostat. 

e

 Highest individual dose is for first year of use. Average individual dose equivalent over

20 years is about 4×104

 mSv (0.04 mrem). Occupant spends 6,100 h/yr inside residence. 
2–118

Table 2.8.3 Summary of Radiation Doses From 10,000 Thermostat Dials and Pointers

Containing 3

H a

Exposure Scenario

Individual Annual Effective 

Dose Equivalent

(mrem)b

Distributionc

 Drivers

 Workers (small retail store)

0.06

0.08

Routine Use

 Installation

 Servicing

 Routine use

 <0.001

 <0.001

0.06

Disposald

 Landfill

 Incineration

 <0.001

 <0.001

Accidentse

 Transportation fire

 Warehouse fire

0.2

0.04

Misusef 0.05

a

 Assumptions based on distribution of 10,000 thermostat dials and pointers per year containing

3

H.

b

 1 mrem = 0.01 mSv.

c

 Assumptions based on generic distribution methodology (see Appendix A.3).

d

 Assumptions based on generic disposal methodology (see Appendix A.2). Includes 20 years

of radioactive decay before disposal. 

e

 Assumptions based on generic accident methodology (see Appendix A.1). During a

transportation fire, a firefighter is assumed to be exposed to one shipment of 80 thermostats. 

During a warehouse fire, a firefighter is assumed to be exposed to 1,000 thermostats stored in

a warehouse.

f

 Assuming an individual wears a 20-year-old thermostat dial or pointer as costume jewelry for

520 h/yr (NRC, 49 FR 18308). The dose is the total dose equivalent to the skin and to the

internal organs from absorption of 3

H by the skin.
2–119

2.9 Electron Tubes

2.9.1 Introduction

In 10 CFR 30.15(a)(8), any person who receives, possesses, uses, transfers, owns, or acquires

electron tubes containing byproduct material, including spark gap tubes, power tubes, gas

tubes and glow lamps, receiving tubes, microwave tubes, indicator tubes, pickup tubes,

radiation detection tubes, and any other completely sealed tubes that are designed to conduct

or control electrical currents, are exempted from licensing requirements, provided that the

following quantities of radioactivity and radiation level are not exceeded:

5.6 gigabecquerel (GBq) (150 millicurie (mCi)) of tritium (3

H) per microwave receiver

protector tube or 0.37 GBq (10 mCi) of 3

H per any other electron tube,

0.037 MBq (1 Ci) of 60Co,

0.19 MBq (5 Ci) of 63Ni,

1.1 MBq (30 Ci) of 85Kr,

0.19 MBq (5 Ci) of 137Cs,

1.1 MBq (30 Ci) of 147Pm, and

an absorbed dose rate of 0.01 milligray (mGy)/h (1 mrad/h) at 1 cm from any surface

when measured through 7 mg/cm2

 of absorber.

The exemption for 147Pm in spark gap tubes was proposed on February 7, 1967 (32 FR 2575),

and issued as a final rule on April 26, 1967 (32 FR 6433), but the limit on the absorbed dose at

that time was 0.005 mGy/h (0.5 mrad/h). Except for 3

H in microwave receiver protector tubes,

all of the other present exemptions for electron tubes, including the current limit on absorbed

dose of 0.01 mGy/h (1 mrad/h), were proposed on November 14, 1968 (33 FR 16602), and

issued as a final rule on April 18, 1969 (34 FR 6651). The exemption for 3

H in microwave

receiver protector tubes was proposed on December 25, 1969 (34 FR 20276), and issued as a

final rule on June 6, 1970 (35 FR 8820).

Some information on radiological impacts on the public from electron tubes containing

byproduct material are provided by the Atomic Energy Commission (AEC) studies in the

Federal Register notices cited above and by the reports of Buckley et al. (NUREG/CR–1775)

and the National Council on Radiation Protection and Measurements (NCRP) (NCRP 95).

2.9.2 Description of Items

Electron tubes include mostly glow lamps, indicator lamps, display tubes, voltage regulators,

spark gap tubes, voltage-sensitive switching tubes (e.g., lightning arresters and radar

transmit-receiver switches), and noise generators. These tubes are found in household

appliances and lamps, electronic games, electronic instruments and equipment, electricity

distribution systems, communication equipment, and other electronically powered devices.
2–120

A typical electron tube consists of a glass tube, a filler gas, a radioactive source (which, if it is

3

H or 85Kr, may be mixed with the filler gas), and appropriate electrical components. Byproduct

materials provide pre-ionization in gases to pass an electric current, so that the electronic

equipment reads faster and more reliably, or displays more constant characteristics

(UNSCEAR, 1982).

Indicator lamps are used in electrical appliances such as clothes washers and dryers, stereos,

coffeemakers, and pinball machines (NCRP 95). Such electron tubes typically contain about

740 KBq (20 Ci) of 3

H or 7.4 KBq (0.2 Ci) of 85Kr, and typically have a length of 2.5 cm and a

radius of 0.25 cm. Several hundred million of these tubes with a service life of about 25,000

hours were utilized in appliances during the 1970s (NUREG/CR–1775). 

Voltage regulators and surge arresters are used to provide protection against voltage

transients, which are a particular hazard to solid-state equipment operating on AC power lines

(Marshall, 1973). These devices usually contain less than 37 KBq (<1 Ci) of 60Co, 63Ni, 85Kr

(gas), or 137Cs. The voltage surge arrester is a smaller version of the microwave receiver

protector tube or so-called radar transmit-receiver switch (Vodicka, 1966).

Glow lamps or spark gap tubes are used as starters for compact fluorescent lamps and in

electric blanket thermostats and other specialty devices (NCRP 95). Several million of these

devices are manufactured annually and generally contain from 37 to 370 KBq (1 to 10 Ci) of

63Ni, 85Kr, or 147Pm (NUREG/CR–1775).

2.9.3 Summary of Previous Analyses and Assessments

In establishing the exemption for 147Pm in spark gap tubes, the AEC concluded there does not

appear to be any significant hazard associated with the possession and use of this product. 

The bases for this conclusion are summarized as follows:

External exposure to beta particles emitted by 147Pm decay would be prevented by the

thickness of the glass walls in spark gap tubes, which exceeds 50 mg/cm2

. The small

amount of low-energy bremsstrahlung produced by stopping of the beta particles in

a tube would result in an external dose to individuals from handling of these devices that

is only a fraction of the recommended limit for members of the public

(5 millisieverts (mSv)/yr (500 mrem/yr) at the time).

It is difficult to conceive of any circumstance in which a significant internal exposure

could be received, because spark gap tubes are extremely difficult to break or crush and

the 147Pm is fixed to the glass walls or the surfaces of the electrodes. Even in the very

unlikely event that one-fourth of the 147Pm in a tube were inhaled, the resulting dose

to the skeleton would be within the recommended limit for members of the public.

For an assumed annual distribution of 110 to 190 GBq (3 to 5 Ci) of 147Pm in spark

gap tubes, the quantity of radioactive material involved, the short half-life of 147Pm,

and the nature of the handling, use, and disposal of these tubes ensure that no

significant population exposure or contamination of the environment would occur.
2–121

The Federal Register notice from 1968 cited above discusses radiological impacts on the public

associated with the exemptions for electron tubes containing 3H, except in microwave receiver

protector tubes, 60Co, 63Ni, 85Kr, and 137Cs. The AEC concluded that the use of these electron

tubes containing these quantities of byproduct material will not constitute an unreasonable risk

to public health. This conclusion was based on the following considerations:

Requirements for manufacture of electron tubes ensure that releases of radioactive

material from defective tubes under normal conditions of use and disposal would not

present a significant hazard to the public.

The limits on the amount of radioactivity and level of absorbed dose for each tube

ensure that doses to individuals from normal handling and use of electron tubes would

not exceed more than a few percentage points of the recommended limits for members

of the public.

The second Federal Register notice from 1969 cited above discusses radiological impacts on

the public associated with the exemption for 3

H in microwave receiver protector tubes. On the

basis of the following considerations, the AEC concluded that the use of these tubes will not

constitute an unreasonable risk to the health and safety of the public:

There is no external radiation hazard because all beta particles emitted in 3

H decay are

absorbed by materials in the tube.

As long as the 3

H is absorbed on metallic tabs and confined in the vacuum-tight

envelope of the electron tube, there would be no uptake of 3

H into the body. If the tube

was severely damaged, 3

H might be dispersed into the atmosphere. However, since

these tubes are normally installed in open-air environments, the resulting rapid dilution

of the dispersed 3

H would ensure that the radiation dose that individuals might receive

near a damaged tube would be very small.

A credible condition could be the storage of a severely damaged tube in a small, closed

room. If the highest recorded 3H escape rate of 2.5 ppm/h is assumed, the annual dose

equivalent to the whole body received by the most highly exposed individuals would not

exceed a few percentage points of the recommended limit for members of the public at

the time.

If a fire occurred in a storage depot containing 10 tubes, the maximum dose equivalent

to the whole body of an individual present in the depot for the first hour would, using

conservative assumptions for intakes, be about 4 millisieverts (mSv) (0.4 rem). This

was less than the recommended limit for members of the public for routine exposures at

the time.

Doses that might result from disposal of tubes or from processing of discarded tubes for

scrap are not expected to be significant.

Thus, although the AEC provided little quantitative information, it is evident that accident

scenarios as well as scenarios for routine use of the electron tubes were considered.
2–122

Buckley et al. (NUREG/CR–1775) performed a detailed analysis of radiological impacts on the

public from use and disposal of electron tubes containing byproduct material. The analysis was

based on data on the average annual distribution of electron tubes that contain 3

H, 60Co, 63Ni,

85Kr, and 147Pm and the average activity of these radionuclides in the tubes over the years 1970

to 1978. Hence, the results of their analysis are not directly related to the limits on activity for

the exemptions specified in 10 CFR 30.15(a)(8).

The routine exposure scenarios for electron tubes containing byproduct material considered by

Buckley et al. (NUREG/CR–1775) include: (1) external exposure to users from tubes containing

the photon-emitting radionuclides 60Co, 85Kr, or 137Cs, (2) inhalation exposure to users from

tubes containing 3

H, and (3) internal exposures from disposal of electron tubes in landfills or by

incineration. The exposures during distribution and transport of electron tubes were not

considered, because the collective doses from these activities were assumed to be much less

than doses from routine use.

The single accident scenario considered by Buckley et al. (NUREG/CR–1775) involved the

inhalation exposure to firefighters during a fire in a warehouse storing tubes containing 3

H, 85Kr,

or 147Pm.

The dose estimates obtained by Buckley et al. (NUREG/CR–1775) for the routine use scenarios

are summarized as follows:

For external exposure during normal use, the annual dose equivalent to the whole body

of an individual from tubes containing 60Co would not exceed 0.003 mSv (0.3 mrem),

and the annual collective dose equivalent would not exceed 0.9 person-Sv (90

person-rem). For tubes containing 85Kr, the annual dose equivalent to the whole body of

an individual would not exceed 3×105

 mSv (0.003 mrem), and the annual collective

dose equivalent would not exceed 10 person-Sv (1000 person-rem). For tubes

containing 137Cs, the annual dose equivalent to the whole body would not exceed 0.004

mSv (0.4 mrem) and the annual collective dose equivalent would not exceed 3

person-Sv (300 person-rem). The estimated upper limits on dose are based on

assuming continuous exposure throughout the year at an average distance of 2 meters

from a tube.

For disposal in landfills, doses to individuals were not estimated, and the annual

collective dose equivalent to the whole body would be 1.2 person-Sv (120 person-rem). 

Essentially all of the collective dose equivalent to the whole body would be from disposal

of electron tubes containing 3

H.

For disposal by incineration, doses to individuals were not estimated, and the annual

collective dose equivalent would be 0.007 person-Sv (0.7 person-rem) to the whole body

and 0.011 person-Sv (1.1 person-rem) to the bone. Nearly all of the collective dose to

the whole body and 70% of the collective dose to the bone would be from tubes

containing 3

H, and most of the remaining dose would be from tubes containing 147Pm.

Based on the analyses of Buckley et al. (NUREG/CR–1775) discussed above, the NCRP

(NCRP 95) concluded that the annual collective effective dose equivalent (EDE) from use and

disposal of all electron tubes containing byproduct material would be about 10 person-Sv

(1,000 person-rem).
2–123

The single accident scenario considered by Buckley et al. (NUREG/CR–1775) involved

inhalation exposure by firefighters during a fire in a warehouse storing tubes containing 3H, 85Kr,

or 147Pm. For this accident scenario, they assumed that 10% of the total inventory of

radioactivity in all electron tubes produced annually would be released in the fire, the exposure

to the firefighters lasted for 8 hours, and the firefighters used no respiratory protection during

the fire. The estimated dose equivalents to firefighters were 2 mSv (0.2 rem) to the whole body

for electron tubes containing 3

H, 40 mSv (4 rem) to the lungs for tubes containing 85Kr, and

2 mSv (0.2 rem) to the whole body and 40 mSv (4 rem) to the bone for tubes containing 147Pm. 

This analysis provides overestimates of doses that reasonably could be received in a fire.

Thus, the conclusion is that the existing analyses of radiological impacts on the public from the

use and disposal of electron tubes containing byproduct material are incorrect, and the

following refinements to the existing analyses are needed: 

• First, for exposure during routine use, the estimates of individual dose should provide

doses that would result from electron tubes containing the maximum quantities of

byproduct material allowed by this exemption. In addition, the assumptions used in

estimating external dose should be reevaluated and the individual doses should be

estimated for all radionuclides used in electron tubes. 

• Second, internal dose from normal use of tubes containing 3

H should be reevaluated

and internal dose from breakage of tubes should be considered. 

• Third, the radiation doses from a fire should be reevaluated using more realistic

assumptions about exposure times and the use of protective clothing and respiratory

equipment. 

• Finally, estimates should be developed of (1) individual doses to on- and off-site

members of the public from disposal of electron tubes in landfills and (2) individual

doses from incineration of electron tubes.

2.9.4 Present Exemption Analysis

Table 2.9.1 provides data from Nuclear Regulatory Commission (NRC) records on the annual

distribution of electron tubes containing radionuclides from 1970 to 1986 (NRC, Unpublished

Reports, 1989). It must to be noted that the average activity per tube in Table 2.9.1 is

significantly less than the maximum activity per tube allowed by the exemption (see

Section 2.9.1). Hence, individual doses resulting from single-tube exposures during routine use

and during accidents are assumed to be controlled by the maximum amount of activity per tube. 

However, all other individual doses resulting from multiple-tube exposures and, thereby, all

collective doses are assumed to be controlled by the average activity per tube and by the

number distributed historically (see Table 2.9.1). Individual doses based on both maximum and

average activities per tube for routine exposures are provided, however, because the latter

were developed for use in the estimation of collective dose. It is further assumed that the

useful lifetime of the electron tubes is 10 years (NCRP 95) and the leakage rate of 3

H is

2.5 ppm/h (see Section 2.9.3). The leakage rate of 2.5 ppm/h for 3

H in an electron tube is

consistent with that used in the previous assessment by Buckley et al. (NUREG/CR–1775).
2–124

2.9.4.1 Distribution and Transport

Shipments of consumer products from manufacturers might contain several hundred to several

thousand electron tubes. Hence, it is assumed that (1) a typical shipment might contain 1,000

electron tubes with the average activities per tube listed in Column 4 of Table 2.9.1 and

(2) 84,500 such shipments per year would be required for the average annual distribution of

84.5 million electron tubes listed in Column 3 of Table 2.9.1.

Consumer products containing the electron tubes are assumed to be shipped by the

manufacturer to a nearby distribution center by commercial semi-truck (see Appendix A.3). It is

further assumed that commercial semi-trucks are used to ship the consumer products between

distribution centers, and the electron tubes pass through an average of three distribution

centers before being delivered by commercial semi-truck to a large retail store. In addition,

(1) radiation doses to distribution workers are assumed to be the same as estimated for

workers in a large warehouse, (2) retail store clerks in some departments are assumed to be

exposed to an average of 100 electron tubes during the year, and the (3) leakage rate from

electron tubes containing 3

H is assumed to be 2.5 ppm/h. This is 2.5 times the value of 1

ppm/h used in the development of the generic methodology in Appendix A.3.

Based on the above assumptions and the generic methodology of Appendix A.3, the individual

receiving the largest dose is an electronics or housewares department store clerk, who is

assumed to be exposed to an average of 100 electron tubes containing 60Co during the year. 

The annual EDE to this individual (see Table 2.9.2) is estimated to be 0.004 mSv (0.4 mrem). 

The annual collective dose from distribution and transport (see Table 2.9.2) is estimated to be

about 1 person-Sv (100 person-rem), due almost entirely to exposure to store clerks and

shoppers.

2.9.4.2 Routine Use

Because electron tubes are made in a variety of designs and may be used in a variety of

exposure situations, it is not feasible to attempt a detailed assessment of electron tubes. Thus,

to indicate the potential radiation doses from use of electron tubes for this assessment, the

following exposure scenarios were chosen. In the first scenario, a person is exposed to five

electron tubes in a home-like environment, and in the second scenario, a person is exposed to

one electron tube in a work-like environment.

Scenario I. A person is exposed to five electron tubes in a home-like environment. This person

is assumed to be in the house for 6000 h/yr (EPA/600/P–95/002Fa) at average distances of

1 meter from one tube and 3 meters and 6 meters from two of the other four tubes. These

distances were used to represent average distances from a tube, not necessarily the same

tube, while a person was moving about the house. Each tube was assumed to contain the

average quantity of the radionuclide of interest (see Table 2.9.1), and each radionuclide was

assessed separately (see Table 2.9.3). External dose equivalents from exposure to gamma

rays emitted by radionuclides contained in the tube and to bremsstrahlung produced by

stopping of beta particles in the glass wall of the tube were calculated using CONDOS II

(Computer Codes, O’Donnell et al., 1981) (see Appendix A.3). The contribution from

bremsstrahlung is important in the case of radionuclides such as 85Kr and 147Pm, which decay

primarily by the emission of beta particles. For 63Ni, the bremsstrahlung radiation is of such low

energy that the contribution to the EDE is essentially zero (0). (Refer to Appendix A.4.)
2–125

For tubes containing 3

H gas, the EDE was calculated assuming a leakage rate from the tubes

of 2.5 ppm/h, a building volume of 450 m3

, a building ventilation rate of 1 volume change per

hour, a breathing rate of 0.9 m3

/h, and a dose conversion factor per unit intake from Table 2.1.2

of 2.6×1011 Sv/Bq (9.6×105

 rem/Ci)). The resulting equilibrium air concentration of 3H,

assumed to be present as water vapor in these dose calculations, was 41 MBq/m3

 (1.1 pCi/m3).

Scenario II. A person is exposed to one electron tube in a work-like environment. The tube

was assumed to contain an average quantity of a radionuclide (see Table 2.9.1), and each

radionuclide was assessed separately (see Table 2.9.3). The person is assumed to work in a

small room at 1 meter from the tube for 2000 h/yr and to have a breathing rate of 1.2 m3

/h. For

3

H, the equilibrium air concentration was 180 MBq/m3 (4.9 pCi/m3) based on a small-room

volume of 18 m3

 (small repair shop from Appendix A.1) and the other assumptions of

Scenario I.

Table 2.9.3 gives individual and collective doses that could be received by persons exposed

under two scenarios. The collective EDEs were calculated by assuming that (1) each electron

tube was used an average of 10 years (NCRP 95), (2) three persons were involved in each

multitube exposure (three persons per five tubes) under Scenario I, and (3) one person was

involved in each single-tube exposure (one person per tube) under Scenario II. Collective

EDEs from the annual distribution for each type of tube, half in multi-tube exposures

(Scenario I) and half in single-tube exposures (Scenario II) are presented in Table 2.9.3. The

collective EDE from all types of electron tubes totals approximately 10 person-Sv

(1000 person-rem).

To determine individual doses at the maximum quantities of byproduct materials allowed by this

exemption (see Section 2.9.1), the results of Scenario II, as shown in Table 2.9.3, were used. 

For 3

H, the individual annual dose in a work-like environment at the average quantity per tube of

1.45 MBq (3.9×102

 mCi) is 1×105 mSv (0.001 mrem). The individual annual dose at the

maximum quantity allowed by the exemption of 5.6 GBq (150 mCi) for a microwave receiver

protector tube is 0.05 mSv (5 mrem) (see Table 2.9.4). Also, for 60Co and 137Cs, an individual

annual dose of 0.02 mSv (2 mrem) was determined at the maximum allowed quantities of

0.037 MBq (1 Ci) of 60Co and 0.19 MBq (5 Ci) of 137Cs. For collective doses, it would be

unreasonable to assume that all electron tubes would ever be produced and distributed at the

maximum quantity allowed by an exemption. Collective doses could be higher, however, if a

significant fraction of electron tubes were distributed at or near the limits of this exemption.

In addition to the limits of this exemption on activity per tube, it is also required that the

absorbed dose rate at 1 cm from any surface must be 0.01 mGy/h (1 mrad/h) or less, when

measured through an absorber with a mass thickness of 7 mg/cm2

 (see Section 2.9.1). The

effect of this latter limit is to restrict the amount of material that can be used in small electron

tubes, such as indicator lamps (see Section 2.9.2). For example, the exemption limits of

0.037 MBq (1 Ci) of 60Co and 0.19 MBq (5 Ci) of 137Cs would require that a cylindrical

electron tube or the protective cover over a cylindrical electron tube have a radius of

approximately 2.5 cm. Assuming that the inverse square law is applicable to the radiation field

from the tube and that an absorbed dose of 0.01 mGy (1mrad) is equal to an EDE of 0.01 mSv

(1 mrem), the annual EDE to an individual exposed for 2000 h/yr, at an average distance of

1 meter from an electron tube having an outer radius of 2.5 cm and reading 0.01 mSv/h

(1 mrem/h) at 1 cm from the outer surface of the tube, is approximately [(0.01 mSv/h) × (1 cm +

2.5 cm)2

/(100 cm)2] × (2000 h/yr), or 0.02 mSv/yr (2 mrem/yr). This value agrees to within
2–126

±30% of the values for 60Co and 137Cs in Table 2.9.4, which are based on more exact

calculations for the EDE using CONDOS II (Computer Codes, O’Donnell et al., 1981) (see

Appendix A.3).

2.9.4.3 Disposal

To estimate potential individual and collective doses to members of the public from the disposal

of electron tubes containing byproduct material in landfills and by incineration, the generic

disposal methodology in Appendix A.2 is used. The effective lifetime of the electron tubes is

assumed to be 10 years. The amount of activity in 1 year’s distribution of electron tubes after

10 years of radioactive decay is as follows: 700 TBq (190 Ci) of 3

H, 93 MBq (2.5 mCi) of 60Co,

2.6 GBq (71 mCi) of 63Ni, 410 GBq (11 Ci) of 85Kr, 520 MBq (14 mCi) of 137Cs, and 32 GBq

(850 mCi) of 147Pm. Most of the electron tubes are assumed to remain intact during waste

collection. Therefore, the dose-to-source ratios in Appendix A.2 for inhalation and ingestion by

waste collectors at both landfills and incinerators have been reduced by a factor of 10.

Estimates of individual and collective doses from disposal of electron tubes are presented in

Tables 2.9.5 and 2.9.6, respectively. As noted from Table 2.9.5, the most highly exposed

individuals are waste collectors. For waste collectors at landfills, the annual individual dose

equivalent is estimated to be 3×104

 mSv (0.03 mrem). For waste collectors at incinerators, the

annual individual dose is estimated to be 2×103

 mSv (0.2 mrem). As further noted from

Table 2.9.6, the total collective dose is estimated to be 4×104

 person-Sv (0.04 person-rem) for

landfills and 4×104

 person-Sv (0.04 person-rem) for incinerators. The disposal of electron

tubes containing 3

H and 85Kr accounts for about 70% of the total collective dose from landfills

and 85% of the total collective dose from incinerators.

2.9.4.4 Accidents and Misuse

In this section, the following situations are considered: 

• a repairman who continuously carries small indicator lamps containing average activities

of 10 kBq (0.28 Ci) of 85Kr in the pocket of his coveralls; 

• misuse or accidents involving the crushing of a microwave receiver protector tube

containing 5.6 GBq (150 mCi) of 3

H or a spark gap generator containing 1.1 MBq

(30 Ci) of 147Pm;

• accidents involving a residential fire and five electron tubes containing average activities

of 3

H, 60Co, 63Ni, 85Kr, 137Cs, or 147Pm; and 

• accidents involving a warehouse fire and 10% of the annual production of electron tubes

containing 3

H, 85Kr, and 147Pm.

For a repairman who carries one electron tube containing an average activity of 10 kBq

(0.28 Ci) of 85Kr in his pocket for 2000 h/yr, the estimated dose equivalent is 0.4 mSv/yr

(40 mrem/yr) from external irradiation of skin and the EDE is 6×104

 mSv/yr (0.06 mrem/yr)

from external irradiation of the whole body. The dose equivalent to skin is based on a

calculation for a separation distance of 1 cm between the small electron tube and skin. The

EDE is based on a calculation at a tissue depth of 10 cm, which is considered a reasonable
2–127

approximation for the average depth of the body organs relative to a small source on the

surface of the body. These calculations for 85Kr were made using CONDOS II (Computer

Codes, O’Donnell et al., 1981) (see Appendix A.3), so that bremsstrahlung produced by the

stopping of beta particles in the glass wall of the electron tube was included in the dose

estimates. If the repairman routinely carried an average of five tubes in his pocket, the dose

equivalent from external irradiation of skin could be 2 mSv/yr (200 mrem/yr) and the annual

EDE from external irradiation of the whole body could be 0.003 mSv (0.3 mrem).

For crushing of a microwave receiver protector tube containing 5.6 GBq (150 mCi) of 3

H in a

small volume room of about 18 m3

 (see, for example, the data for a small watch repair shop in

Table A.1.9 of Appendix A.1), the estimated EDE was 10 mSv (1000 rem) from inhalation and

skin absorption of 3

H. This estimate assumes that (1) the ventilation rate in the room is 1

volume change per hour, (2) the individual’s breathing rate is 1.2 m3

/h, and (3) the individual

remains in the room for at least 4 hours after the tube was crushed (see Appendix A.1). For

crushing of a spark gap tube containing 1.1 MBq (30 Ci) of 147Pm, the generic accident

methodology developed in Appendix A.1 for inhalation following the spill of a radioactive

material in the form of a powder was used, but applied here to a small room with a volume of 18

m3

 and a ventilation rate of 1 volume change per hour. Thus, the EDE from inhalation of 147Pm

from the spark gap tube is estimated to be 8×104

 mSv (0.08 mrem), assuming the individual

remained in the small room for 4 hours after the tube was crushed.

For a residential fire involving five electron tubes containing average activities of 3

H, 60Co, 63Ni,

85Kr, 137Cs, or 147Pm (see Table 2.9.1), the individual dose estimates from inhalation and

submersion can be summarized as follows:

The maximum individual EDE to a person trying to escape from a residential fire or a

neighbor trying to rescue a person from such a fire would be 8×105

 mSv (0.008 mrem)

and would occur at a fire involving five tubes containing average activities of 1.4 MBq

(39 Ci) of 3

H.

The maximum individual EDE to a firefighter who wears protective clothing and a

respirator during a residential fire would be less than 1×105

 mSv (<0.001 mrem) and

would occur at a fire involving five tubes containing average activities of 1.4 MBq

(39 Ci) of 3

H.

The maximum individual EDE to a worker who is involved in the cleanup following a fire,

but does not wear protective clothing and a respirator, would be less than 1×105

 mSv

(<0.001 mrem). This would occur at a residential fire involving five tubes containing

average activities of 21 KBq (5.6 Ci) of 147Pm.

For a warehouse fire involving 10% of the activity in the annual production of electron tubes

containing 3

H, 85Kr, and 147Pm (see Table 2.9.1), the individual dose estimates from inhalation

and submersion can be summarized as follows:

At a warehouse fire involving 10% of the activity in the annual production of electron

tubes containing 3

H or 1.2 TBq (33 Ci), the individual EDE to a firefighter who wears

protective clothing and a respirator during the fire is estimated to be 0.005 mSv

(0.5 mrem). 
2–128

At a warehouse fire involving 10% of the activity in the annual production of electron

tubes containing 85Kr or 78 GBq (2.1 Ci), the individual EDE to a firefighter who wears a

respirator during the fire would be 0.009 mSv (0.9 mrem). Normally, the dose

equivalent to the lungs from 85Kr contained within the lungs is about equal to the EDE

from submersion in 85Kr (NCRP 44). However, the dose equivalent to the lungs from

inhalation of 85Kr by a firefighter who wears a respirator is small in comparison to the

EDE to the firefighter from submersion in the 85Kr.

At a warehouse fire involving 10% of the activity in the annual production of electron

tubes containing 147Pm or 44 GBq (1.2 Ci), the individual EDE to a firefighter who wears

a respirator during the fire would be 7×105

 mSv (0.007 mrem). The individual EDE to a

worker who is involved in the cleanup following the fire but does not wear a respirator

would be 5×104

 mSv (0.05 mrem). The dose equivalent to the bone (i.e., the endosteal

cells on the bone surfaces) from inhalation of 147Pm is about twice the above values for

the EDE to a firefighter or worker involved in the cleanup following the fire

(EPA–520/1–88–020).

The latter set of dose estimates for a warehouse fire are found to be at least two orders of

magnitude smaller than the previous dose estimates made by Buckley et al.

(NUREG/CR–1775) (see Section 2.9.3).

2.9.5 Summary 

Table 2.9.7 presents the results of this assessment of potential radiation doses to members of

the public from the distribution, routine use, and disposal of electron tubes containing byproduct

materials. These results are based on an assumed 1 year’s distribution of 84.5 million electron

tubes with average amounts of byproduct material, as listed in Table 2.9.1, and an assumed

useful lifetime of 10 years per tube.

For routine use of electron tubes including distribution and disposal, the annual effective dose

to the most highly exposed individual (a routine user of electron tubes) was estimated to be

about 0.05 mSv (5 mrem). The total collective dose equivalent to the public, nearly all of which

is received by routine users of electron tubes, was estimated to be about 10 person-Sv

(1000 person-rem).

For accidents, the estimated individual EDE was 10 mSv (1 rem) for crushing of a microwave

receiver tube containing the maximum amount of 3

H allowed by this exemption. For both

accidents and misuse of electron tubes containing other radionuclides with activities at or below

the maximum amount allowed by this exemption, the individual EDEs would be significantly

less.
2–129

Table 2.9.1 Average Annual Distribution of Electron Tubes Containing Byproduct

Material During the Period of 1970 to 1986 a

Average Annual Distribution

Radionuclide mCi/yrb tubes/yr Ci/tubeb

3

H

60Co

63Ni

85Kr

137Cs

147Pm

 3.3×105

 9.3

 7.6×101

 2.1×104

 1.7×101

 1.2×104

 8.3×106

 3.2×104

 7.4×104

 7.4×107

 1.7×104

 2.1×106

 39

 0.29

 1.0

 0.28

 1.0

 5.6

a

 NRC, Unpublished Reports, 1989.

b

 1 mCi = 37 MBq; 1 Ci = 37 kBq.
2–130

Table 2.9.2 Potential Radiation Doses From Distribution and Transport of Electron

Tubes Containing Average Quantities of Byproduct Materials a

Radionuclide

Individual Annual Effective

Dose Equivalentb

(mrem)d

Collective Effective

Dose Equivalentc

(person-rem)d

3

H

60Co

85Kr

137Cs

147Pm

<0.001

0.4

<0.001

0.3

<0.001

40

 2

70

8

0.3

a

 Individual doses from distribution and transport result from exposure to multiple tubes and,

therefore, are based on the average activities per tube in Column 4 of Table 2.9.1.

b

 Dose estimates apply to store clerks in some electronics and housewares departments; dose

estimates would be less for store clerks in other departments, truck drivers, distribution

workers, and members of the public exposed along the truck routes or while shopping in retail

stores (see Section 2.9.4.1). 

c

 Collective doses are based on the average annual distribution of electron tubes in Column 3 of

Table 2.9.1. 

d

 1 mrem = 0.01 mSv. 1 person-rem = 0.01 person-Sv.
2–131

Table 2.9.3 Potential Radiation Doses From Routine Use of Electron Tubes

Containing Average Quantities of Byproduct Materials a, b

Radionuclide

Average Quantity

per Electron Tube

(Ci)c

Individual Dose

Over First Year

(mrem)d

Individual Dose

Over 10 Years

(mrem)d

Number of

Exposed Persons

C

O

(

SCENARIO I

3H

60Co

85Kr

137Cs

147Pm

39

0.290.281.05.6

<0.001

20.0042

<0.001

0.004

10

0.03

10

0.003

2.5×106

9.6×1032.2×1075.1×1036.3×105

SCENARIO II

3H

60Co

85Kr

137Cs

147Pm

39

0.29

 0.28

1.05.6

0.0010.50.0010.4

<0.001

0.0130.0084

<0.001

4.2×106

1.6×1043.7×1078.5×1031.1×106

a Collective doses result from multiple tube exposures and, therefore, are based on average activities per tube 

Table 2.9.1.

b Scenario I is for exposure to multiple electron tubes in a home-like environment; Scenario II is for exposure to 

in a work-like environment (see Section 2.9.4.2).

c 1 Ci = 37 kBq.d 1 mrem = 0.01 mSv.e 1 person-rem = 0.01 person-Sv.
2–132

Table 2.9.4 Potential Radiation Doses to Individuals From Routine Use of Electron 

Tubes Containing the Maximum Quantity of a Byproduct Material Allowed 

by Exemption a

Radionuclide Maximum Quantity

per Electron Tubeb

Individual Annual Effective

Dose Equivalentc

(mrem)d

3

H

60Co

85Kr

137Cs

147Pm

150 mCi

1 Ci

30 Ci

5 Ci

30 Ci

5

2

0.1

2

0.001

a

 Individual doses may result from single tube exposures and, therefore, are based on the

maximum activities per tube allowed by the exemption (see Section 2.9.1).

b

 1 mCi = 37 MBq; 1 Ci = 37 kBq. c

 Individual doses over first year of exposure to a single tube containing the maximum allowed

activity based on Scenario II (see Section 2.9.4.2). 

d

 1 mrem = 0.01 mSv.
2–133

Table 2.9.5 Potential Individual Doses From Disposal of Electron Tubes Containing 

Average Quantities of Byproduct Material

Annual Effective Dose Equivalent

(mrem)a

Radionuclide Waste Collectorsb On-Site Workersb

Off-Site Residentsb

(airborne releases)

Off-Site Residentsc

(water releases)

LANDFILL

3H

60Co

63Ni

85Kr

137Cs

147Pm

<0.001

0.004

<0.001

0.020.005

<0.001

<0.001<0.001<0.001<0.001<0.001<0.001

<0.001<0.001<0.001<0.001<0.001<0.001

<0.001 

 Total 0.03 <0.001 <0.001 <0.001 
2–134

Table 2.9.5 Potential Individual Doses From Disposal of Electron Tubes Containing 

Average Quantities of Byproduct Material

(continued)

Annual Effective Dose Equivalent

(mrem)a

Radionuclide Waste Collectorsb On-Site Workersb

Off-Site Residentsb

(airborne releases)

Off-Site Residentsc

(water releases)

INCINERATION

3H

60Co

63Ni

85Kr

137Cs

147Pm

<0.001 0.02<0.001 0.1 0.03<0.001

<0.001<0.001<0.001<0.001<0.001<0.001

<0.001<0.001<0.001<0.001<0.001<0.001

NAe 

 Total 0.2 <0.001 <0.001

a 1 mrem = 0.01 Sv.b Dose estimates are for 1 year’s disposal of electron tubes at landfills and incinerators (see Section 2.9.4.3).c Dose estimates are for 30 years’ disposal corrected for decay during 30 years of disposal (see Appendix A.2).d Dose estimates are for 30 years’ disposal corrected for decay during 30 years of disposal plus an additional 3

following landfill closure (see Appendix A.2).

e Not applicable.
2–135

Table 2.9.6 Potential Collective Doses From Disposal of Electron Tubes Containing 

Average Quantities of Byproduct Material

Collective Effective Dose Equivalent a

(person-rem)b

Radionuclide Waste Collectorsc On-Site Workersc

Off-Site Residentsc

(air transport)

Off-Site Residentsd

(water transport)

Future 

Resid

LANDFILL

3H

60Co

63Ni

85Kr

137Cs

147Pm

<0.001

0.01

<0.001

0.060.02

<0.001

0.0020.009

<0.001

0.020.01

<0.001

<0.001<0.001<0.001<0.001<0.001

<0.001

0.2 <0

<0<0

00

<0

 Total 0.09 0.04 <0.001 0.2 0
2–136

Table 2.9.6 Potential Collective Doses From Disposal of Electron Tubes Containing 

Average Quantities of Byproduct Material

(continued)

Collective Effective Dose Equivalent a

(person-rem)b

Radionuclide Waste Collectorsc On-Site Workersc

Off-Site Residentsc

(air transport)

Off-Site Residentsd

(water transport)

Future 

Resid

INCINERATION

3H

60Co

63Ni

85Kr

137Cs

147Pm

<0.001

0.003

<0.001

0.010.005

<0.001

<0.001<0.001<0.001<0.001<0.001<0.001

0.02

<0.001<0.001<0.001<0.001<0.001

NAf N

 Total 0.02 <0.001 0.02 

a Collective doses are based on disposal of electron tubes equal to 1 year’s distribution following 10 years of ra

each radionuclide in electron tubes (see Section 2.9.4.3). 

b 1 person-rem = 0.01 person-Sv.c Dose estimates for exposure during 1 year’s disposal of electron tubes (see Appendix A.2).d Dose estimates for 1000 years of exposure to 1 year’s disposal of electron tubes (see Appendix A.2).e Dose estimates for 1000 years of exposure to 1 year’s disposal of electron tubes after landfill closure, plus 30 

A.2).

f Not applicable.
2–137

Table 2.9.7 Potential Radiation Doses From Electron Tubes Containing 

Byproduct Material

Exposure Pathway

Individual Annual 

Effective

Dose Equivalent

(mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport

Routine use

Disposal

 Landfills

 Incineration

Accidents and misuse

 Fire

 Carrying in pocket

 Crushing of tubes

0.4c

5d

0.02e

0.2f

0.9g

0.3h

1000i

120

1,300

0.4

0.04

a

 1 mrem = 0.01 mSv. 1 person-rem = 0.01 person-Sv.

b

 Collective doses are based on an annual distribution of 84.5 million electron tubes containing

byproduct materials as listed in Column 3 of Table 2.9.1. Refer to text for time period for

collective dose calculations. 

c

 Dose estimate applies to store clerks in some electronics and housewares departments; dose

estimates would be less for store clerks in other departments, truck drivers, distribution

workers, and members of the public exposed along truck routes or while shopping in retail

stores less (see Section 2.9.4.1).

d

 Dose estimate applies to individuals exposed to electron tubes containing maximum activities

of byproduct materials allowed by this exemption; dose estimates for individuals exposed to

more typical activities of byproduct materials distributed in electron tubes would be significantly

less (see Section 2.9.4.2).

e

 Dose estimate applies to waste collectors at landfills; dose estimates are less for workers at

landfills, off-site members of the public, and future on-site residents at landfills (see

Section 2.9.4.3).

f

 Dose estimate applies to waste collectors at incinerators; dose estimates for workers at

incinerators and off-site members of the public (see Section 2.9.4.3).

g

 Dose estimate applies to firefighters at a warehouse fire involving 10% of the annual

production of electron tubes containing 85Kr; dose estimates would be less for individuals at

warehouse fires involving other radionuclides. Dose estimates are also less for residential fires

involving five tubes containing average activities of either 3

H, 60Co, 63Ni, 85Kr, 137Cs, or 147Pm

(see Section 2.9.4.4).

h

 Dose estimate applies to irradiation of the whole body of a repairman who carries five electron

tubes containing average activities of 85Kr in his pocket for 2,000 h/yr; dose estimate for

irradiation of a small area of skin is 2 mSv (200 mrem) (see Section 2.9.4.4).

i

 Dose estimate applies to the crushing of an electron tube containing maximum amount of 3H

allowed by this exemption; dose estimates for crushing of tubes containing typical activities of

other radionuclides would be significantly less (see Section 2.9.4.4).

2–139

2.10 Ionizing Radiation Measurement Instruments Containing,

for Purposes of Internal Calibration or Standardization,

One or More Sources of Byproduct Material

2.10.1 Introduction

In 10 CFR 30.15(a)(9), the receipt, possession, use, transfer, ownership, or acquisition of

ionizing radiation measuring instruments containing, for purposes of internal calibration or

standardization, one or more sources of byproduct material are exempted from licensing

requirements, provided that (1) each source contains no more than one exempt quantity set

forth in 10 CFR 30.71, Schedule B, and (2) each instrument contains no more than 10 exempt

quantities. Excepted from this exemption are persons who apply byproduct material to, or

persons who incorporate byproduct material into, such instruments, or persons who initially

transfer for sale or distribution such instruments containing byproduct material. An instrument’s

source(s) may contain either one type or different types of radionuclides. An individual exempt

quantity may be composed of fractional parts of one or more of the exempt quantities in

10 CFR 30.71, Schedule B, provided that the sum of such fractions shall not exceed unity. For

purposes of this exemption, 2 kilobecquerel (KBq) (0.05 microcurie (Ci)) of 241Am is

considered an exempt quantity under 10 CFR 30.71, Schedule B.

The exemption for ionizing radiation measuring instruments, except for instruments containing

241Am, was first established on April 22, 1970 (35 FR 6426), in conjunction with the

establishment of exempt quantities of byproduct material (see Section 2.13). An amendment to

permit installation of multiple sources in such instruments was proposed on February 25, 1981

(46 FR 14019), and issued as a final rule on May 13, 1981 (46 FR 26471). An amendment to

include the exempt quantity of 241Am given above was proposed on July 9, 1981 (46 FR 35522),

and issued as a final rule on September 23, 1981 (46 FR 46875).

Receipt, possession, use, transfer, or acquisition of byproduct material used for calibration,

standardization, or as a reference for instruments is addressed in two additional sections of the

regulations. The use of individual sources containing exempt quantities of byproduct material

(not contained within instruments) set forth in 10 CFR 30.71, Schedule B, is allowed under

10 CFR 30.18 (see Section 2.13). In addition, a general license is issued in 10 CFR 31.5 to

commercial and industrial firms and research, educational and medical institutions, individuals

in the conduct of their business, and Federal, State, or local government agencies to acquire,

receive, possess, use, or transfer byproduct material contained in devices designed and

manufactured for a number of specific purposes, including measuring radiation or producing

light. Certain ionizing radiation measuring instruments (e.g., thermoluminescence dosimeter

readers and liquid scintillation counters) use internal calibration and reference sources

distributed for use under this general license.

There are no limits in 10 CFR 31.5 on the amount of byproduct material that can be used in

devices, but an applicant for a specific license to manufacture or initially transfer devices for

use under 10 CFR 31.5 must demonstrate that the devices will meet certain safety

requirements contained in 10 CFR 32.51. Section 4.5 provides a separate analysis, which

includes evaluation of proposed changes in Nuclear Regulatory Commission (NRC) regulations

to exempt certain radionuclides and quantities for calibration and reference sources, which

would include some devices now used under the general license in 10 CFR 31.5.
2–140

2.10.2 Description of Items

Sources that qualify for exemption from licensing requirements under 10 CFR 30.15(a)(9) are

described in reports submitted to the NRC by instrument manufacturer licensees as part of their

responsibilities under 10 CFR 32.16 and 10 CFR 32.20. Table 2.10.1 summarizes the

quantities of specific radionuclides distributed as internal radiation monitoring instrumentation

calibration or standardization sources reported by licensees from 1985 to 1995. These reports

describe source assemblies fabricated using 9 different radionuclides. The average unit

quantity of byproduct material and fractional part of Schedule B quantity per unit, also shown in

the table, were derived from the annual quantities and numbers of units. No information is

available about the number of instruments or sources of byproduct material in use, but

estimates can be derived by assuming that annual distribution represents the replacement rate

for instruments taken out of service for disposal. The number of instruments in use is the

product of useful life (average time from distribution to disposal) and replacement rate. 

Estimates of useful life for estimating the number of instruments and sources in use are

provided in Section 2.10.4.2.

The form of the radioactive source is discussed in only a few reports. Two manufacturers

describe sources consisting of 137Cs adsorbed onto ion exchange resin beads; another, sources

fabricated from the salt of 137Cs dispersed in ceramic. A fourth licensee manufactures sources

from 241Am electroplated onto metal rods. The source assemblies themselves are typically

constructed by depositing the byproduct material into a recess in a metal rod, screw, or plate,

then permanently securing the material in place by crimping, welding, or gluing the source into

the recess. These assemblies, in turn, are housed inside the instrument case or within some

integral component of larger systems and are normally not accessible by the user. One

manufacturer, however, indicates the source assemblies may be sold separately. One may

assume this is generally true for sources whose byproduct material has a relatively short halflife, or for instruments that may house a variable number of sources. 

An exception to the source fabrication designs described above is found in a recent license

amendment request to allow incorporation of small amounts of 241Am as an impurity in

scintillation detector crystals, which are coupled optically with photomultiplier tubes and

hermetically sealed as complete subassemblies inside metal containers. No reports were found

of the numbers of such devices or the total quantities of 241Am used in this way, although the

manufacturer estimated in a telephone interview that about six such devices, each containing

370 to 740 Bq (0.01 to 0.02 Ci) of 241Am, are distributed to exempt users annually. The

estimated annual distribution of 241Am is 11 KBq (0.3 Ci) in 15 devices, about equally divided

between scintillation detectors and other types of instruments, and is included in Table 2.10.2. 

This estimate includes summary data from the NRC for 1970 to 1989 (NRC, Unpublished

Reports, Janney, 1990) in addition to the information from telephone interviews with

manufacturers. (No quantities of 241Am were included in manufacturers’ reports to the NRC

from 1985 to 1995 (NRC, Licensee Reports, Material Transfer Reports, 1985-1995), but

interviews indicate that some 241Am-containing devices described in these reports are currently

being distributed to exempt users under 10 CFR 30.15(a)(9) or to general licensees for use

under 10 CFR 31.8.)

Instruments in which sources distributed under 10 CFR 30.15(a)(9) are used may be classified

as handheld survey instruments, benchtop devices for laboratory use, or area/process monitors
2–141

that operate unattended in isolated locations for long periods of time. The total quantity of

byproduct material and the number of items distributed for use in each type of instrument from

1985 to 1995 (NRC, Licensee Reports, Material Transfer Reports, 1985-1995) were determined

from reports submitted to the NRC by instrument manufacturer licensees (supplemented by

telephone interviews when additional information was required). The final estimates are shown

in Table 2.10.2.

2.10.3 Summary of Previous Analyses and Assessments

No estimates were found relating to hazards from distribution and transport, use, accidents and

misuse, or disposal of radiation measuring instruments containing byproduct material for

purposes of internal calibration or standardization.

2.10.4 Present Exemption Analysis

This section provides an assessment of the radiation doses to individuals and the public from

routine use, distribution and transport, disposal, and accidents associated with the use of

ionizing radiation measuring instruments containing one or more sources of byproduct material

for purposes of internal calibration or standardization. The safety of such sources is enhanced

not only by inaccessibility to the user implied by their definition in 10 CFR 30.15(a)(9), but also

by secondary containment in ceramic or epoxy matrixes (and, in one case, within the crystal

matrix of the detector material itself) and often by further sealing the source within

subassemblies, which are themselves mounted within the radiation measuring instrument. 

There is no ingestion or inhalation concern for the reported radionuclides during either shipping

or normal use; the principal exposure pathway is external irradiation of the whole body. 

Assumptions about the number of sources and total quantities of byproduct materials

distributed annually are taken from Table 2.10.2.

2.10.4.1 Distribution and Transport

Radioactive sources are distributed in comparatively small quantities under 10 CFR 30.15(a)(9),

usually in sophisticated specialty equipment, and are therefore considered for purposes of this

analysis to be fabricated on demand and shipped directly to the user without intermediate

storage in a warehouse facility. Individual and collective doses from distribution and transport

were estimated using the methodology described in Appendix A.3 and are summarized in

Table 2.10.3. Distribution is assumed to involve five steps: 

express delivery (small truck) from the manufacturer to a nearby airport;

processing at the airport freight terminal and loading on the outbound plane;

air transport by plane;

unloading the plane and processing at the receiving airport freight terminal; and 

local delivery (small truck, within 400 km of the receiving airport) to the user. 
7

 A conventional value of 1760 hours is often used for the typical time spent at work annually

and accounts for vacation, sick leave, etc. This value, taken as an initial value for estimating

time of exposure to byproduct material contained in ionizing radiation measuring instruments

used by radiation protection technicians and technologists, is modified upward by overtime and

is reduced by time spent on administrative duties (e.g., preparing reports). An annual exposure

time of 2000 hours may be somewhat conservative, but conservatism is considered to be

warranted in order to account for a significant amount of variability in the time estimate.

2–142

Individual dose estimates are derived using the greatest annual quantity for each radionuclide

reported by a manufacturer in licensee reports to the NRC (NRC, Licensee Reports, Material

Transfer Reports, 1985-1995). A single driver is assumed to transport all items in a small truck

from a given manufacturer to the same outbound air terminal. Shipments are further assumed

to be equally distributed to 10 regional airports, where equal quantities are directed from each

receiving airport to two users by separate drivers, each of whom makes all the deliveries to a

particular user. Individuals receiving the highest effective dose equivalent (EDE),

0.004 millisievert (mSv) (0.4 mrem), in this scenario are the express-truck drivers who deliver

instruments containing byproduct material from the manufacturer’s facility to the outbound

freight terminal at the nearby airport. The total annual population dose from distribution of

byproduct material under 10 CFR 30.15(a)(9) is 8×10-5 person-Sv (0.008 person-rem). 

Exposure from 137Cs is the predominate contributor to both individual and collective doses from

distribution of these instruments.

2.10.4.2 Routine Use

Doses are estimated from handheld instruments, benchtop instruments, and area/process

monitors based on assumptions about how each of these devices is used, maintained, and

stored. Individual EDEs consider the likely sources and quantities of byproduct material in each

instrument, whereas population doses use estimates of the total quantities of byproduct

material in use. No empirical estimates of quantities in use are available, so these values are

estimated as the product of average source or instrument lifetime and annual distribution

quantity shown in Table 2.10.2. The mean useful lifetime for handheld instruments and for

area/process monitors is assumed to be 5 years, and, for benchtop instruments, 10 years, after

which the instrument and source are replaced. 

Handheld instruments are used primarily by technicians, educators, researchers, and students. 

Radiological control technicians are assumed to use most of the radiation monitoring

instruments distributed annually. About 25% of instruments are estimated to be in use at a

given time, with the remainder either available in an instrument pool or out of service for

calibration and maintenance. The average distance from the source to the whole body of a

single user is approximately 0.5 m and, to his hands, 0.1 m. Instruments not in use are

assumed to be stored at an average distance of 2 m in a room that averages two occupants. 

Only one source is assumed to be provided with each instrument. The time of exposure to both

the instrument user and to the occupants of a room in which instruments are stored is assumed

to be 2000 hours annually7

. An annual dose, approximately 0.1 mSv (10 mrem) EDE and

2 mSv (200 mrem) to hands, is received by a radiological control technician using a radiation

monitoring instrument containing 33 KBq (0.9 Ci) of 60Co (the ratio of total annual quantity of

60Co and total annual items distributed, in Table 2.10.2). A more likely scenario is exposure to

the user of an instrument containing 44 KBq (1.2 Ci) of 137Cs as an internal calibration or
2–143

reference source. This gives a dose of about 0.04 mSv (4 mrem) EDE and 1.0 mSv

(100 mrem) to the hands. The total annual dose to all users of handheld radiation monitoring

instruments is estimated from the assumed 5-year mean lifetime, the annual distribution shown

in Table 2.10.2, and the assumptions used above to be 0.09 person-Sv (9 person-rem) EDE

and 2 Sv (200 rem), almost exclusively from 137Cs, which accounts for most of the byproduct

material used in these devices. The corresponding annual incidental dose to persons who work

in areas where handheld instruments are stored is 0.03 person-Sv (3 person-rem). The total

collective EDE (for both users and others who work around handheld instruments) is

0.1 person-Sv (10 person-rem).

Benchtop instruments are presumed to be used in a laboratory by a dedicated technician who

spends about half of the time (1000 h/yr) with the trunk of the whole body at 0.5 m and hands at

an average 0.2 m from the internal radioactive sources of a single instrument. Multiple sources

might be used in these instruments to verify appropriate response to several different

radionuclides or to the same radionuclide in several ranges. The distribution of 55Fe

predominated between 1985 to 1995 (NRC, Licensee Reports, Material Transfer Reports,

1985-1995) for use in these devices, but the greater exposure potential is from 137Cs. This

analysis assumes an instrument contains three sources of 137Cs at 100%, 33%, and 10% of the

limiting Schedule-B quantity for determining the dose to this technician. An additional

laboratory occupancy of two people who spend about half their time at other workstations

located an average distance of 3 m from the radioactive sources is also assumed. The

technician’s annual dose is approximately 0.2 mSv (20 mrem) EDE and 1 mSv (100 mrem) to

the hands. The total dose to all technicians using benchtop instruments, estimated from the

assumed 10-year mean lifetime, the annual distribution shown in Table 2.10.2, and the

assumptions used above, is 0.009 person-Sv (0.9 person-rem) collective EDE, all from 137Cs. 

The total annual dose to other laboratory occupants is approximately 5×10-4 person-Sv

(0.05 person-rem), so that the total collective EDE (for both technicians and other laboratory

occupants) is 0.01 person-Sv (1 person-rem).

Although a practical application is not known, a maximum theoretical dose to laboratory

technicians of about 2 mSv (200 mrem) could be received from an instrument containing the

allowed ten exempt 137Cs sources (37 MBq (100 Ci)).

Instruments used as area radiation monitors or for monitoring radiation levels in process

streams normally operate without human intervention for long periods, except for occasional

routine maintenance and calibration. Multiple sources might be used in these instruments to

verify appropriate response to several different radionuclides or to the same radionuclide in

several ranges. They may be located either in very isolated areas or in areas with incidental

occupancy, and are likely found in a setting that relies on shift work for continuous operation of

the facility. This analysis assumes that area/process monitors operate in areas continuously

occupied by an average of two shift workers whose average distance from the monitors is 5 m. 

Service personnel are assumed to spend 100 h/yr performing routine calibration and

maintenance for each monitor in situ at a distance of 0.5 m to the trunk of the body and 0.2 m

to the hands. Three radionuclides (133Ba, 60Co, and 137Cs) with nontrivial external exposure

potential were distributed in the greatest quantities for use as internal calibration or

standardization sources in area/process monitors between 1985 and 1995 (NRC, Licensee

Reports, Material Transfer Reports, 1985-1995), so this analysis assumes that each of these

radionuclides is contained in an instrument at the limiting Schedule-B quantity for determining

dose to the individual. 
2–144

Annual dose to the technician performing maintenance and calibration on a single instrument

under these conditions is 0.04 mSv (4 mrem) EDE and 0.2 mSv (20 mrem) to the hands. The

total annual dose to all technicians, based on the assumed 5-year mean lifetime, the annual

distribution shown in Table 2.10.2, and the assumptions used above, is estimated to be

0.006 person-Sv (0.6 person-rem) collective EDE. The incidental annual dose to others

occupying areas in which these monitors are installed is about 0.008 person-Sv

(0.8 person-rem) collective EDE. The total collective EDE (for both instrument maintenance

technicians and other occupants of the area) is about 0.01 person-Sv (1 person-rem).

The analysis depends entirely on the use of the specific gamma-ray dose constant for

estimating external exposures, without considering shielding provided by the instrument case or

source containers. If 0.2 cm of steel were considered as a generic shield, the external

exposure from 137Cs would be reduced by about 7%. Values reported here probably

overestimate actual exposures to instrument users for this reason alone.

2.10.4.3 Disposal

Because the materials were distributed as exempt items, no formal control over disposal is

assured. It is assumed that 90% of this byproduct material is disposed of as municipal waste,

5% is recycled and 5% is disposed of as radioactive waste. No incineration is assumed, since

the sources are typically incorporated into metal subassemblies used in instruments and are

noncombustible. A source is assumed to be used for 10 years before disposal. Doses to

landfill and recycle facility workers and to members of the public are estimated using the

disposal scenarios described in Appendix A.2. Annual disposal quantities are assumed to be

equal to quantities distributed annually, corrected for radiological decay over 10 years of use. 

Byproduct material used in ionizing radiation measuring instruments, then buried in landfills, is

assumed to be 10 times less dispersible in soil or air and 10 times less accessible to water than

loose materials, since sources (except for 241Am, which is discussed below) are typically

stabilized by electroplating or by incorporation in ion exchange resin or ceramic. These forms

are not expected to retain their physical or chemical integrity in a recycling process, however,

and both dispersibility in air and soil and accessibility to water for this disposal scenario are

assumed to be the same as for loose materials in the waste. Accessibility to water and soil for

the landfill scenario is assumed to be the same for 241Am dispersed in scintillator material as for

loose material, since container integrity is the only barrier preventing loss of contents, which are

often both hygroscopic and watersoluble. All of the 241Am is assumed to be distributed in this

form under the provisions of 10 CFR 30.15(a)(9) for the purposes of this analysis. Loss of

container integrity, which results in air in-leakage and exposure to atmospheric moisture, ruins

the devices and could be a primary reason for their disposal. Hygroscopic scintillator materials

typically expand on taking up moisture, further damaging the container and dispersing its

contents.

The doses from all landfill disposal and recycle scenarios were less than 1×105

 mSv (<0.001

mrem). The collective EDE corresponding to a year’s disposal of instruments containing

internal calibration sources is 4×105

 person-Sv (0.004 person-rem), due primarily to exposure

to future on-site residents to 137Cs at landfills for 1000 years after loss of institutional controls

over the sites. The collective EDE excluding exposures to future on-site residents is less than

1×105

 person-Sv (<0.001 person-rem).
2–145

2.10.4.4 Accidents and Misuse

Ionizing radiation monitoring instruments containing sources of byproduct material for

calibration or standardization are used primarily in industry and education, rather than in homes

or small businesses. Accidents involving release of byproduct material are most likely to occur

during transport, storage, or use. Doses from transportation accidents, warehouse fires, and

laboratory fires are assessed using the generic accident methodology described in

Appendix A.1. Doses from residential fires or spills are not evaluated, since sources are

typically solids used in industrial instruments. A release fraction” of 0.01% is assumed, since

byproduct material distributed under 10 CFR 30.15(a)(9) is enclosed within an instrument, and

descriptions provided by manufacturers indicate it may be further encased within

subassemblies and bound in salt or ceramic matrixes. Doses were assessed for benchtop

instruments and area/process monitors. Scenarios evaluated include a transportation accident,

warehouse fire, and a laboratory fire each involving 10 instruments or monitors. All accidents

doses were less than 1×105

 mSv (<0.001 mrem).

Misuse of an internal calibration or reference source might entail removal of a source or

subassembly from the instrument casing, followed by close hand work for or modification by an

unqualified individual. The greatest hypothetical dose from such a scenario would be received

from sources in the area/process monitor described above. A person spending two work weeks

attempting such repair or modification would receive a dose of 0.2 mSv (20 mrem) EDE (with

the trunk of the body located an average distance of 20 cm from the sources) and 20 mSv

(2 rem) to the hands (2 cm from the sources). A maximum hypothetical EDE of 0.8 mSv (80

mrem) and 80 mSv (8 rem) to the hands would be obtained from 10 exempt 137 Cs sources if

they were used in the same assembly. Again, a practical laboratory application for ten exempt

137Cs sources is not known.

2.10.5 Summary

The byproduct-material radionuclide reported to be distributed in the greatest quantities over a

10-year period (NRC, Licensee Reports, Material Transfer Reports, 1985-1995) to users

exempt from licensing requirements under the provisions of 10 CFR 30.15(a)(9) was 55Fe. 

Dosimetric impacts from 55Fe are negligible, however, because of the overwhelming importance

in this analysis of external exposures from sealed sources, normally inaccessible to untrained or

unqualified individuals, for which inhalation and ingestion are not important exposure pathways. 

There are no highly penetrating radiations from 55Fe, whose electrons and X-rays all have

energies less than 6.5 keV (Kocher, 1981) and are easily shielded.

The greatest dosimetric impact from ionizing radiation measurement instruments containing

one or more sources of byproduct material for purposes of internal calibration or

standardization is from 137Cs, second in quantity shipped only to 55Fe. The only scenario in

which 137Cs does not have the greatest impact is an accident involving a fire in which 241Am is

resuspended and inhaled.

Annual EDEs from typical distribution, use, and disposal of these instruments are summarized

in Table 2.10.3. The highest individual EDEs are expected for the routine use of a benchtop

instrument containing three sources of 137Cs with 100%, 33% and 10% of the exempt quantity

specified in 10 CFR 30.71, Schedule B, and in the event of repair or modification of an
2–146

instrument containing sources of 133Ba, 60Co, and 137Cs, each with 100% of its respective

exempt quantity. These doses are 0.02 mSv (20 mrem), each. Although a practical application

is not known, a maximum theoretical dose of 2 mSv (200 mrem) could be received from an

instrument containing the allowed ten exempt 137Cs sources. The greatest population dose,

0.13 person-Sv (13 person-rem), is anticipated from the use of handheld instruments, the most

numerous kind of instrument (by almost an order of magnitude) distributed under

10 CFR 30.15(a)(9).

Weaknesses in this analysis include the lack of measured exposure rates for byproduct

material as it is used in these instruments, lack of uniformity in the kind of information provided

by licensees who manufacture and distribute these devices, and no empirical information about

the total numbers of instruments or quantities of byproduct materials in use. The analysis

depends entirely on the use of the specific gamma-ray dose constant for estimating external

exposures, without considering shielding provided by the instrument case or source containers. 

If 0.2 cm of steel were considered as a generic shield, the external exposure from 137Cs would

be reduced by about 7%. Values reported here probably overestimate actual exposures to

instrument users for this reason alone. Reports from manufacturers did not always provide

information about the amount of byproduct material within sources, or about the number of

sources within instruments, so the analysis had to depend on the use of gross averages or on

extreme limiting values in some cases. Manufacturers’ reports to the NRC (NRC, Licensee

Reports, Material Transfer Reports, 1985-1995) were not always self-consistent, and cross-cut

totals of byproduct material activities did not always agree within a report. Telephone interviews

with manufacturers were required to clarify some of these inconsistencies, not all of which were

fully resolved.
2–147

Table 2.10.1 Reported Quantities of Byproduct Material and Numbers of Units

Distributed From 1985 to 1995 for Use as Internal Calibration or

Standardization Sources a

Radionuclide

Total Quantity

(Ci)b

Number of

Units

Average Unit

Quantityc

(Ci)b

Average Fractional

Part of Schedule B

Quantity per Unit

133Ba 640 129 4.9 0.49

14C 0.6 4 0.15 0.0015

36Cl 610 792 0.77 0.077

60Co 18 22 0.69 0.69

137Cs 12,000 6,937 1.7 0.17

55Fe 20,000 204 100 1

129I 0.001 4 0.00025 0.0025

90Sr 28 317 0.089 0.89

99Tc 1.4 59 0.024 0.0024

a

 NRC, Licensee Reports, Material Transfer Reports, 1985-1995.

b

 1 Ci = 37 kBq. c

 Ratio of total quantity and number of units.
2–148

Table 2.10.2 Estimated Annual Quantity of Byproduct Material and Number of Sources

Distributed in Each Ionizing Radiation Monitoring Instrument Classification a

TOTAL QUANTITY (Ci)b

Radionuclide

Area/Process

Monitors

Benchtop

Instruments Handheld Instruments

241Amc 0.15 0.15 

133Ba 190 

14C 0.18 

36Cl 180 

60Co 16 2.7

137Cs 660 67 2,500

55Fe 6,100 

129I 0.0003

90Sr 8.4 0.03

99Tc 0.43

TOTAL NUMBER OF SOURCES

Radionuclide

Area/Process

Monitors

Benchtop

Instruments Handheld Instruments

241Amc 6 6 

133Ba 39 

14C 1 

36Cl 240 

60Co 4 3

137Cs 75 8 2,000

55Fe 61 

129I 1

90Sr 84 11

99Tc 18

a

 Derived primarily from the same data used in generating Table 2.10.1; includes earlier

summary data (NRC, Unpublished Reports, Janney, 1990) and data from interviews with

manufacturers when recent data are not available in manufacturers’ reports to the NRC (NRC,

Licensee Reports, Material Transfer Reports, 1985-1995).

b

 1 Ci = 37 kBq. c

 Data for 241Am are estimates based on information from telephone interviews with

manufacturers and from NRC summary data from 1970 to 1989 (NRC, Unpublished Reports,

Janney, 1990).
2–149

Table 2.10.3 Estimated Effective Dose Equivalents From Distribution, 

Use, and Disposal of Ionizing Radiation Measuring Instruments Containing 

One or More Sources of Byproduct Material for Purposes of Internal

Calibration or Standardization

Exposure Pathway

Individual Annual

Effective

Dose Equivalent

 (mrem)a

Collective Effective

Dose Equivalentb

 (person-rem)a

Distribution and transportc 0.4 0.008

Routine use

 Handheld instruments

 Benchtop instruments

 Area and process monitors 

10c

20f

 4g

10d, c

1d, e

1d, e

Disposal as ordinary trashc

 Landfill

 Recycle facility

<0.001

<0.001

0.004

Accidents and misuse

 Transportation fireh

 Warehouse fireh

 Laboratory fireh

 Repair or modificationg

<0.001

<0.001

<0.001

20

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Refer to text discussion for time period for collective dose calculations.

c

 Each instrument contains 33 kBq (0.9 Ci) of 60 Co.

d

 Exposure from 137Cs is the predominate contributor. e

 Derived from total quantities in use, estimated as the product of useful life and annual

distribution.

f

 Each instrument contains three sources with 370 kBq (10 Ci), 122 kBq (3.3 Ci), and 37 kBq

(1 Ci) of 137Cs. The maximum theoretical dose is about 2 mSv (200 mrem) for the unlikely

scenario of an instrument containing the allowed exempt 137Cs sources.

g

 Instrument contains three sources with 370 kBq (10 Ci) 133Ba, 37 kBq (1 Ci) 60Co, and

370 kBq (10 Ci) 137Cs. For exposure to 10 exempt sources of 137Cs (the maximum

hypothetical exposure situation), the EDE is 0.8 mSv (80 mrem) and 80 mSv (8 rem) to the

hands.

h

 Ten instruments contain three sources with 1.9 kBq (0.05 Ci) 241Am, 37 kBq (1 Ci) 60Co, and

370 kBq (10 Ci) 137Cs. The greatest contribution to the estimated dose is from the

resuspension and inhalation of 241Am.

2–151

2.11 Spark Gap Irradiators

2.11.1 Introduction

In 10 CFR 30.15(a)(10), persons who receive, possess, use, transfer, own, or acquire spark

gap irradiators containing not more than 37 kilobecquerel (kBq) (1 microcurie (Ci)) of 60Co per

irradiator are exempted from licensing requirements for byproduct material, provided such

irradiators are used in electrically ignited fuel-oil burners having a firing rate greater than

11 liters per hour (>3 gallons per hour). The firing rate requirement ensures that spark gap

irradiators containing 60Co will be used almost exclusively in commercial and industrial

buildings, not in small furnaces used in private homes or in internal combustion engines. This

exemption was proposed on October 24, 1975 (40 FR 49801), and issued as a final rule on

January 17, 1978 (43 FR 2386).

The first Federal Register notice cited above includes dose estimates for individuals. These

estimates are discussed in Section 2.11.3 below. A more complete dose analysis is given in

the environmental impact statement for this exemption (NUREG–0319).

2.11.2 Description of Item

Spark gap irradiators containing cobalt are designed to minimize spark delay in some

electrically ignited commercial fuel-oil burners by generating free electrons in the spark gap. 

These free electrons are produced from beta particles emanating from the 60Co plating on the

irradiator. Their use is limited to spark-ignited fuel-oil burners with fuel input capacities greater

than 3 gallons per hour. Each irradiator is installed on a standard 1.6-cm diameter, hexagonal

pressure nozzle inside the burner. The 60Co is plated over an area of about 0.15 cm2

; the

thickness of the cobalt deposit is between 7×1010 and 2×109

 cm. Since the irradiators are

self-operating, the potential for exposure during actual use is limited to routine burner

maintenance (i.e., during cleaning, adjustment, or nozzle replacement of existing burner units).

A single burner manufacturer designed and has been the sole distributor of spark gap

irradiators. Apparently, the original estimates of spark gap irradiator demand were overly

optimistic. The irradiators are no longer being manufactured, only about 100 irradiators were in

stock in 1994, and no plans have been made to distribute them for use (phone call,

R. Westover, Ray Burner Co., Richmond, CA, September, 1994). The original manufacturer is

no longer in business. The number of irradiators actually distributed is unknown, but is not

thought to be significant.

2.11.3 Summary of Previous Assessments

Doses from routine use and disposal of spark gap irradiators containing 37 kBq (1 Ci) of 60Co

were previously estimated in the environmental impact statement for this exemption

(NUREG–0319). In the previous work, external exposure was assumed to be the only

important exposure mode. The majority of the exposure came from delivery, installation, and

maintenance of the irradiators. The annual distribution of 6000 irradiators was assumed

(i.e., six times the historic demand established in the original rulemaking petition). The dose

estimates were as follows:
2–152

The effective dose equivalent (EDE) to maximally exposed individuals (deliverers and

installers of irradiators) from photon irradiation was 0.12 millisievert (mSv)/yr

(12 mrem/yr).

The collective EDE from photon irradiation, about half of which is received by deliverers

and installers of irradiators and the other half by operators and servicemen at the oil

burners, was 0.04 person-Sv/yr (4 person-rem/yr).

The maximum dose from improper handling of irradiators was estimated to be 0.58 mSv

(58 mrem) to the total body (carrying an irradiator in a side pocket for 2000 h/yr).

The maximum dose from misuse (i.e., a child finding and saving an irradiator) was

estimated to be 0.02 mSv (2 mrem) to the total body (carrying it in a pocket for 10 h/day

for 1 week). A potential average skin dose to the hand from handling an irradiator for

1 h/day was estimated to be 0.015 Sv (1.5 rem).

Exposure from accidents involving fire was determined to be improbable since 60Co should not

become volatile in a vehicular or building fire.

2.11.4 Present Exemption Analysis

For this analysis, the original scenarios (i.e., number of persons involved, time of exposure,

distance to receptor, etc.) for distribution, transport, routine use, and misuse are considered

adequate (NUREG–0319). Collective doses were not determined as this product is not

believed to be currently manufactured or in wide-scale use. A hypothetical distribution of 1000

spark gap irradiators is assumed for the purpose of modeling distribution, transport, and

disposal. This was the annual demand originally estimated by the petitioner (NUREG–0319). 

For dose estimates from accidents and disposal, the generic methodologies presented in

Appendixes A.1 and A.2, respectively, are used. 

Dose factors for distribution, transport, routine use, and misuse were generated using

MicroShield (Computer Codes, Grove Engineering, 1996). The resultant dose factors were

then used to generate the individual and collective EDEs based on the quantity of radionuclide

in an irradiator or package of irradiators (as point sources), the duration of exposure for a

particular activity.

2.11.4.1 Distribution and Transport

The same assessment scenarios as previously used (NUREG–0319) were assumed for this

step, with the exception of an annual hypothetical distribution of 1000 irradiators. The original

petitioner estimated that 95% of the irradiators were shipped as a very small item packaged

with a much larger piece of mechanical equipment (the burner unit) via truck directly from the

manufacturer to the installer (Method 1). This would account for 950 irradiators, 95 units to

each of 10 installers, shipped as needed. The remaining 5% (50 irradiators) would involve

mailing a parcel containing 10 irradiators from the manufacturer directly to users or installers

(50 irradiators; one parcel containing 10 irradiators to each of five installers). Method 2 involves

several steps for distribution (i.e., post offices, sectional centers, airports). Assumptions used
2–153

in the scenarios for maximally exposed groups of people and corresponding annual individual

and collective dose estimates are included in Table 2.11.1.

The maximum individual EDEs during distribution and transport were estimated to be

0.003 mSv/yr (0.3 mrem/yr) to the truck driver for Method 1 and 7×104

 mSv/yr (0.07 mrem/yr)

to the post office carrier for Method 2. 

2.11.4.2 Routine Use (Installation and Maintenance)

The original scenarios (NUREG–0319) were also used in the routine use evaluation of this

assessment. Irradiators are installed on the pressure nozzle of spark-ignited oil burners used in

commercial or industrial facilities. An irradiator may be installed on a new burner or on an older

unit being serviced or remodeled. No one would come in direct contact with irradiators during

normal operation. However, during routine burner maintenance operations, such as cleaning,

adjustment, or replacement of nozzles, the irradiator might be handled, but only persons having

business in such areas would enter or work in them.

During installation, the installer would receive an irradiator that would eventually be taken to the

burner and installed. The actual installation procedure should require only a few seconds, but

transport from the shop to the installation site could take a few hours with the irradiator near the

installer.

The following assumptions are used in the calculations of potential doses:

Each irradiator contains 37 kBq (1 Ci) of 60Co.

Method 1 distribution assumes that 950 irradiators go directly to 10 installer facilities,

with each receiving and installing 95 units per year.

Method 2 distribution involves 50 irradiators, one parcel (each containing 10 irradiators)

going to each of five installer facilities.

A team of two workers from each shop delivers and installs the burners.

EDEs from installation are calculated for direct external exposure. Table 2.11.2 contains

annual EDEs from installation and maintenance. The maximum EDE is to installers and is

estimated to be 0.1 mSv (10 mrem).

2.11.4.3 Accidents and Misuse

Fires involving 60Co should not result in measurable impacts because the temperatures would

be much lower than those required to vaporize the radioactive material. Based on the generic

accident methodology in Appendix A.1, an irradiator involved in a transportation fire would yield

an EDE of less than 1×105

 mSv (<0.001 mrem) per irradiator. In a warehouse fire involving all

1000 irradiators, the potential maximum EDE would also be less than 1×105

 mSv

(<0.001 mrem).

Scenarios involving misuse of spark gap irradiators are the same as those postulated in the

original evaluation (NUREG–0319). Doses were evaluated using the methodology described in
2–154

Appendix A.4. The potential for misuse is considered small because of their packaging and

location of installation (i.e., on commercial burners). For the extreme assumption that an

installer or serviceman ignores warnings or is forgetful and carries an irradiator in a side pocket

for 2000 h/yr, the maximum annual EDE would be about 2 mSv/yr (200 mrem/yr) from photons. 

Conservatively, assuming an unshielded source at a distance of 1 cm from the skin and a

0.7 mm cloth cover (side pocket), the hypothetical dose to a small area of skin could be 1 gray

(Gy) (100 rads). The EDE from this skin exposure would be less than 0.01 mSv (<1 mrem),

considering a skin weighting factor of 0.01, an exposed area of 10 cm2

, and a total skin area of

1.8×104

 cm2.

If a lost irradiator was found and saved by an adult, the potential doses probably would be less

than those resulting from improper handling by installers and servicemen, as given above. In

an extreme and unlikely case, a child could carry an irradiator in a pocket for 10 h/day for

1 week. The resulting EDE could be 0.06 mSv (6 mrem) and the localized skin dose could be

0.04 Gy (4 rads).

2.11.4.4 Disposal

The generic disposal methodology discussed in Appendix A.2 was used to estimate the doses

from disposal of used spark gap irradiators. Deposition in a landfill, incineration (although an

irradiator supposedly will not vaporize in a conventional incinerator), and recycling of scrap steel

were considered. Table 2.11.3 includes the estimated EDE from each form of disposal.

It is assumed that 80% (or 800 irradiators) are disposed of in landfills and 20% (or 200

irradiators) are incinerated (see Appendix A.2). If an individual irradiator initially contained

37 kBq (1 Ci) of 60Co and had a useful life of 15 years, the activity remaining after 15 years

would be about 5.2 kBq (0.14 Ci). For 800 irradiators to be discarded as ordinary trash and

deposited in landfills, the annual EDE to the waste collector, landfill operator and future on-site

and off-site members of the public would be less than 1×105

 mSv (<0.001 mrem). For 200

irradiators to be incinerated, the highest dose is to the collector and is estimated to be less than

1×105

 mSv (<0.001 mrem).

There is a potential for recycle of scrap steel by mixing 1000 discarded irradiators with scrap

steel. For the off-site resident during smelter operation, the maximum individual annual EDE

would be less than 1×105

 mSv (<0.001 mrem). For a user of an automobile manufactured with

recycled steel, the resulting maximum individual annual EDE would be 4×105

 mSv

(0.004 mrem) if all 1000 irradiators are recycled.

2.11.5 Summary

Table 2.11.4 is a summary of the results of the current reanalysis of radiological impacts on the

public for use and disposal of spark gap irradiators containing cobalt. For distribution,

transport, routine use, and misuse of spark gap irradiators, the same scenarios generated in

the original assessment were used (NUREG–0319). The results of this assessment are based

on the hypothetical distribution of 1000 irradiators, each of which contains 37 kBq (1 Ci) of

60Co.
2–155

Previous studies (NUREG–0319) estimated a maximum individual EDE of 0.12 mSv (12 mrem)

to deliverers and installers. This assessment has yielded a maximum EDE of about 0.1 mSv/yr

(10 mrem/yr) to the same group of workers.
2–156

Table 2.11.1 Exposure Conditions for Distribution and Transport of 1,000 Irradiators

and Corresponding Doses

Exposure Event

Duration

of Event

(h)

Average

Distance

from

Irradiatorsa

(cm)

Maximum Annual

Effective Dose

Equivalent per

Unitb

(mrem)c

Number of

Units

Involvedb

Total Annual 

Effective

Dose

 Equivalent

per Individual

(mrem)c

DISTRIBUTION AND TRANSPORT BY METHOD 1d

Truck drivers

 (1 irradiator

 per trip)

0.03-10 15-300 0.003 95 0.3

DISTRIBUTION AND TRANSPORT BY METHOD 2d

Postal system

 driverse

0.03-0.5 30-450 0.0004 10 0.004

Post office

 receiver

2.5 90 0.003 10 0.03

Post office

 carrier

0.03-2 30-300 0.007 10 0.07

Sectional center

 (receiving,

 sorting,

 loading)

0.03-2 30-150 0.001 10 0.01

DISTRIBUTION AND TRANSPORT BY METHOD 2d

Airline loaders

 and unloaders

 (15 parcels)

0.03-0.25 30-450 4×104 10 0.004

Airline

 passengers

 (1parcel/plane)

2.5 280 3×106 10 <0.001

a

 The range of distances reflects the variability during certain operations within each specific group of

people.

b Unit” may be one irradiator in the case of distributing an irradiator as part of a burner unit (Method 1)

or as a parcel containing 10 irradiators (Method 2). 

c

 1 mrem = 0.01 mSv.

d

 Method 1 distribution involves 95% of the 1,000 irradiators (i.e., 950) going directly from manufacturer

to each of 10 installers (95 irradiators each). Method 2 involves distribution of the remaining 50

irradiators through the postal system (i.e., post offices, sectional centers) and airports to the installer

(each of five installers receives one parcel containing 10 irradiators).

e

 Postal system drivers include those driving to and from post offices and sectional centers.
2–157

Table 2.11.2 Exposure Conditions for Installation and Maintenance of 1,000 Irradiators

and Corresponding Doses

Exposure Event

Duration

of Event

(h)

Average

Distance

from

Irradiatorsa

(cm)

Maximum

Annual

Effective Dose

Equivalent per

Unitb

(mrem)c

Number

of Units

Involvedb

Total Annual

Collective

Effective

Dose

Equivalent

(person-rem)c

INSTALLATIONd

For 950 irradiators

 (Method 1)

 Delivery and

 installation: max.

 individual carrying

 1 irradiator

1-8 20-150 0.2 50/team 10

For 50 irradiators

 remaining; 9

 irradiators per shop

960 610 0.02 9/shop 0.2

Delivery and

 installation: max.

 individual carrying

 1 irradiator

8 20-150 0.2 10/shop 2

MAINTENANCE

Operators doing

 burner check

Operators doing

 other work

350 

1400

150

460

0.14

0.06

1

1

0.2

Service

 In vicinity

 During irradiator

 service

8

1

150

 30

0.003

0.01

1

1

0.001

a

 The range of distances reflects the variability during certain operations within each specific group of

people.

b Unit” may be one irradiator in the case of distributing an irradiator as part of a burner unit (Method 1)

or as a parcel containing 10 irradiators (Method 2). A team of two persons each installs 50 irradiators.

c 1 mrem = 0.01 mSv. 1 person-rem = 0.01 person-Sv.

d

 Method 1 distribution involves 95% of the 1,000 irradiators (i.e., 950) going directly from manufacturer

to each of 10 installers (95 irradiators each). Method 2 involves distribution of the remaining 50

irradiators through the postal system (i.e., post offices, sectional centers) and airports to the installer

(each of five installers receives one parcel containing 10 irradiators).
2–158

Table 2.11.3 Exposure Conditions for Disposal of 1,000 Irradiators

and Corresponding Doses

Exposure Event

Maximum Annual Individual Effective Dose

Equivalent

(mrem)a

LANDFILL

Collector <0.001

Operator <0.001

On-site resident <0.001

Off-site resident <0.001

INCINERATOR

Collector 0.001

Worker <0.001

Off-site resident <0.001

RECYCLE

Off-site resident <0.001

Userb 0.004

a

 1 mrem = 0.01 mSv.

b

 An individual driving an automobile containing recycled steel for 2000 h/yr (e.g., taxicab

driver).
2–159

Table 2.11.4 Hypothetical Radiation Doses From Spark Gap Irradiators 

Containing 60Co

Exposure

Pathway

Maximum Annual 

Individual Effective

Dose Equivalent per

Unit

(mrem)a

Total Maximum Annual

Individual Effective Dose

Equivalentb

(mrem)a

DISTRIBUTION AND TRANSPORT

Method 1

 (950 irradiators)

0.003 0.3

Method 2

 (50 irradiators)

0.007 0.07

INSTALLATION

Method 1

 (950 irradiators)

0.2 10

Method 2

 (50 irradiators)

0.02 0.2

Maintenance 0.2 0.2

Disposal 0.004c

Accidents and

 misuse

200e

a

 1 mrem = 0.01 mSv.

b

 Maximum individual effective dose equivalent (EDE) includes the number of irradiators being

handled per person.

c

 Highest individual EDE was to a user of an automobile made from recycled steel.

d

 Misuse involving a worker inadvertently carrying one irradiator in pocket 2,000 h/yr. Other

situations were considered, but the doses were much less.

2–161

2.12 Resins Containing 46Sc for Sand Consolidation in Oil Wells

2.12.1 Introduction

In 10 CFR 30.16, persons who receive, possess, use, transfer, own, or acquire synthetic resins

containing 46Sc that are designed for sand consolidation in oil wells are exempted from licensing

requirements for byproduct material. The exemption does not authorize the manufacture or

initial transfer for sale or distribution of any resins containing 46Sc. According to 10 CFR 32.17,

the concentration of 46Sc in the final product at the time of distribution will not exceed

52 becquerel (Bq)/mL (1.4×103

 microcurie (Ci)/mL). The exemption was proposed on

December 14, 1966 (31 FR 15747), and issued as a final rule on March 18, 1967 (32 FR 4241).

As proposed in the 1966 rule, it was considered unlikely that an oil field worker, even if

continuously performing the sand consolidation process, would be exposed to radiation in

excess of the annual standard applicable to an individual in the public (5 millisieverts (mSv)

(500 mrem) whole-body irradiation). In addition, the resins were not intended for use by the

public and, because of the low concentrations used and the short half-life, the resulting doses

to members of the public would not constitute an unreasonable risk to their health and safety.

Scandium-46, is used as a tracer in sand consolidation operations as well as in other oil and

gas studies.

In researching this exemption, it was found that 46Sc is no longer tagged to plastic resins.

Rather, 46Sc is now primarily adsorbed onto glass beads or encapsulated in ceramic or alumina

beads, a process not covered under this exemption (i.e., it is a licensed activity). 

2.12.2 Description of Item

Radioactive tracers have been used by petroleum engineers to monitor the effectiveness of

formation fracturing and primary cementing, as well as to assess progress of water flow and

tertiary recovery techniques (Williams and McCarthy, 1987). Sand consolidation is used to

prevent loose sand from entering the oil, thereby plugging the well and limiting production. The

formation is still porous, which allows for continued oil production. Sand consolidation is also

used in fractured stimulated wells. 

When the exemption was proposed in 1966, resins were used to consolidate loose sand and

were tagged with 46Sc. Currently, resin-coated sand is used to consolidate loose sand, and the

46Sc tracer, which is encapsulated in a ceramic or alumina bead, is added to the sand, a

process that is not covered under this exemption. Resins as the primary cementing media are

no longer used (Phone call, F. Calloway, Radiation Safety Officer, Isotag, Inc., Odessa, TX,

February 1997; phone call, L. Stephenson, Director of Environmental Compliance, Protechnics

International, Inc., Houston, TX, March 1995 and April 1997). Because of these new

radioactive tracer configurations, it is very unlikely that resins will be tagged with 46Sc. Resins

tagged with 46Sc were somewhat difficult to use, and often 46Sc washed off the resin. Resins

tagged with 46Sc have not been used in at least 10 years (Phone call, F. Calloway, Radiation

Safety Officer, Isotag, Inc., Odessa, TX, February 1997; phone call, L. Stephenson, Director of

Environmental Compliance, Protechnics International, Inc., Houston, TX, March 1995 and April
2–162

1997; phone call, F. Hamiter, Bureau of Radiation Control, Texas Department of Health, March 1995.).

2.12.3 Summary of Previous Analyses and Assessments 

Chevron Research Company filed a petition with the Atomic Energy Commission (AEC)

requesting that the Commission’s regulations be amended to provide a general license to use

and introduce into oil wells 46Sc-labeled resins for sand consolidation. Because of the low 46Sc

concentrations, short half-life, low level of exposure to oil workers, and the conclusion that the

likelihood that this product could be diverted was remote, the AEC believed an exemption rather

than a general license would be appropriate. The primary source of the proceeding analyses is

the proposed rule issued in the Federal Register on December 14, 1966 (31 FR 15747). The

46Sc concentration in resins and the dose rate to oil field workers exposed to resin drums, which

were used to substantiate the exemption, are described in this section.

2.12.3.1 Oil Field Worker Dose Assessment

Based on experimental determination, the minimum feasible 46Sc concentration that could be

used with the resins was 52 Bq/mL (1.4×103

Ci/mL) (31 FR 15747). By limiting the 46Sc

concentration and minimizing resin handling (direct injection of the resin into oil wells with no

special field preparation), the external dose-equivalent rate around resin drums (with the

expected form of packaging) was reported to be less than 0.005 mSv/h (<0.5 mrem/h). 

Therefore, according to 31 FR 15747, it was unlikely that an oil field worker, even if

continuously involved in sand consolidation, would be exposed in excess of the radiation

standards applicable to individual members of the public. When the proposed rule was issued,

the annual nonoccupational exposure limit was 5 mSv (500 mrem). Ingestion of the resin was

considered unlikely. Inhalation of significant amounts of 46Sc was also considered unlikely,

since scandium is not volatile, the resin is a viscous material, and the resin is pumped directly

from the drum through a hose. However, there were no quantitative analyses supporting these

statements.

2.12.3.2 Other Members of the Public

According to the Federal Register (31 FR 15747), the resins are designed to be introduced into

oil wells, and the probability of a diversion to other uses appeared to be remote. Even if part of

the tagged resin was introduced into the plastics industry, the low concentration and short halflife of 46Sc would result in low levels of exposure. Transfer of 46Sc to another medium, either

inadvertently or by design, appeared extremely unlikely. However, there were no quantitative

analyses supporting these statements.

2.12.4 Present Exemption Analysis

Resins tagged with 46Sc are no longer used. However, 46Sc is still used as a tracer in sand

consolidation operations and in other oil well studies using a process not covered by this

exemption. The workers who use 46Sc and other radioactive tracers for oil well activities do this

under the authority of a specific Nuclear Regulatory Commission or specific Agreement State

license. For these reasons, no present exemption analyses were done for distribution and

transport, routine use, and accidents and misuse.
2–163

2.13 Quantities of Byproduct Material

2.13.1 Introduction

In 10 CFR 30.18, persons who receive, possess, use, transfer, own, or acquire byproduct

material in individual quantities that do not exceed the values for radionuclides listed in

Schedule B of 10 CFR 30.71 are exempted from licensing requirements. The exemption does

not authorize the production, packaging, repackaging, or transfer of byproduct material for

purposes of commercial distribution or the incorporation of byproduct material into any

manufactured or assembled commodity, product, or device intended for commercial distribution.

Requirements for licensees who manufacture, process, produce, package, repackage, or

transfer quantities of byproduct material for commercial distribution to persons exempt pursuant

to 10 CFR 30.18 are specified in 10 CFR 32.18. These regulations specify that quantities

authorized for exempt distribution are not to be contained in any food, beverage, cosmetic,

drug, or other commodity designed for ingestion or inhalation by, or application to, a human

being, and that the byproduct material is to be in the form of processed chemical elements,

compounds, or mixtures, tissue samples, bioassay samples, counting standards, plated or

electroplated sources, or similar substances, which are identified as radioactive and are to be

used for their radioactive properties.

Licensees who manufacture, process, produce, package, repackage, or transfer quantities of

byproduct material pursuant to 10 CFR 32.18 also are subject to the following requirements

specified in 10 CFR 32.19. First, no more than 10 quantities shall be sold or transferred in any

single transaction. For purposes of this requirement, an individual quantity may be composed

of fractional parts of one or more of the quantities, provided that the sum of such fractions shall

not exceed unity. Second, each quantity of byproduct material shall be packaged separately

and individually, no more than 10 such packaged quantities shall be contained in any outer

package for transfer to persons exempt pursuant to 10 CFR 30.18, and the dose equivalent

rate at the surface of the outer package shall not exceed 5 millisieverts (mSv)/h (500 mrem/h). 

Finally, several requirements for labeling are specified, including that the immediate container

or an accompanying brochure shall bear the words Radioactive Material—Not for Human

Use—Introduction into Foods, Beverages, Cosmetics, Drugs, or Medicinals, or Into Products

Manufactured for Commercial Distribution is Prohibited—Exempt Quantities Should Not Be

Combined.”

This exemption was proposed on August 10, 1968 (33 FR 11414), and issued as a final rule on

April 22, 1970 (35 FR 6426), except the quantity for 133Ba was proposed on May 25, 1971

(36 FR 9468), and issued as a final rule on August 26, 1971 (36 FR 16898).

In addition to the regulatory requirements noted above, the Federal Register notice for the

proposed rule stated that licensees of the Nuclear Regulatory Commission (NRC) or an

Agreement State who are authorized to manufacture, process, or produce byproduct material

also are authorized to make transfers, on a noncommercial basis, of quantities of byproduct

material possessed under their license. This provision is designed to accommodate the

occasional transfers between laboratories of small quantities of byproduct material in such

items as tissue samples, bioassay samples, tagged compounds, and counting standards. Such

transfers are expected to involve a negligible risk.
2–164

The basis for the quantities of byproduct material authorized for exempt distribution established

by the Atomic Energy Commission (AEC) is described in the first Federal Register notice cited

above. Except for 85Kr, each selected quantity was derived based on one of two radiological

criteria. First, since inhalation was considered the most likely route of entry into the body, the

quantity was calculated for each radionuclide that would be inhaled by a reference individual

from continuous exposure over a year to the maximum permissible concentration (MPC) in air

for members of the public, as listed in Table II of Appendix B of 10 CFR 20 (AEC,

25 FR 10914). Second, for each gamma-emitting radionuclide, the quantity that would produce

a radiation level of 0.26 microcoulomb (C)/kg-h (1 milliroentgen (mR)/h) at a distance of 10 cm

from a point source was calculated. Then, the smaller of the two quantities calculated for

inhalation and external exposure was logarithmically rounded to the nearest decade, in

microcuries, and adopted as the quantity in Schedule B of 10 CFR 30.71. For 85Kr, the quantity

was based on the limitation of external dose to the skin from beta irradiation.

When a quantity in Schedule B was based on the MPC in air for members of the public, its

value was intended to correspond approximately to an annual committed dose equivalent from

continuous inhalation exposure to 5 mSv (0.5 rem) to the whole body, 30 mSv (3 rem) to the

bone or thyroid, or 15 mSv (1.5 rem) to any other organ (AEC, 25 FR 10914), as calculated

using the dosimetric and metabolic models in Publication 2 of the International Commission on

Radiological Protection (ICRP 2). When the quantity was based on the potential external

exposure, the specified criterion of an exposure rate of 0.26 C/kg-h (1 mR/h) at a distance of

10 cm from a point source corresponds to an annual dose equivalent to the whole body from

continuous exposure at this distance of about 60 mSv (6 rem), which is about an order of

magnitude greater than the then-existing dose criterion for limiting external exposure to the

public of 5 mSv/yr (0.5 rem/yr) (AEC, 25 FR 10914). 

In adopting the procedure described above for determining quantities in Schedule B, the AEC

reasoned that under the conditions of the exemption, it is unlikely any individual would inhale (or

ingest) more than a very small fraction of any radioactive material being used or receive

excessive doses of external radiation when realistic source-to-receptor distances and exposure

times are assumed. Therefore, it was considered highly improbable that any member of the

public exposed to byproduct material in quantities less than the limits for exemption would

receive an annual dose equivalent more than a small fraction of recommended limits for the

public at the time the exemption was developed.

2.13.2 Description of Materials

As described in the previous section, quantities of byproduct material authorized for distribution

are not to be incorporated into any manufactured or assembled commodity, product, or device

intended for commercial distribution. Instead, they are intended primarily for use in products or

materials that have teaching or research applications, including tissue samples, bioassay

samples, tagged compounds, and counting standards. Particular examples of products

containing quantities of byproduct material authorized for exempt distribution include sources

for calibration of radiation detectors and sources for use in spiked counting samples.

Quantities of byproduct material authorized for exempt distribution have been established for

more than 170 radionuclides. However, most of these radionuclides apparently have not been

distributed commercially to any significant extent. An indication of the particular radionuclides
2–165

that have been distributed in the greatest amounts in recent years is provided by the data in

Table 2.13.1, which were obtained from a review of materials licensee reports sent to the NRC

(NRC, Licensee Reports, Material Transfer Reports, 1985-1995). These data do not

necessarily represent the total activity of the various radionuclides that have been distributed,

primarily because noncommercial transfers are not included. Reports of transfers are required

only from commercial distributors.

2.13.3 Summary of Previous Analyses and Assessments

As discussed in Section 2.13.1, the Federal Register notice for the proposed rule indicated that,

under the conditions of the exemption, it is unlikely that any individual would inhale (or ingest)

more than a very small fraction of any radioactive material being used or receive excessive

doses of external radiation when realistic source-to-receptor distances and exposure times are

assumed. This conclusion was based on several factors, including the requirement that the

materials be identified as radioactive by appropriate labels and brochures, the low likelihood

that radionuclides in the materials would be inhaled or ingested, and the low allowable external

dose rates near the materials. In addition, collective doses should be limited by the

requirement that quantities of byproduct material authorized for exempt distribution not be

incorporated into any manufactured or assembled commodity, product, or device intended for

commercial distribution and by the condition that no more than 10 quantities shall be sold or

transferred in a single transaction. However, example quantitative analyses of individual and

collective doses for particular materials and particular exposure scenarios were not presented.

Only one previous assessment is known of the potential radiological impacts on the public

associated with the use of quantities of byproduct material authorized for exempt distribution. 

Specifically, the NRC staff (NRC, Memoranda, Paperiello, 1994) considered inhalation

exposure to workers in a laboratory while using a chemical solution containing 14C in amounts

significantly less than the quantity in Schedule B. The solution is highly volatile and is used to

evaluate the level of micro-organic residues in solvents or directly on surfaces, and the

deposited contents are deliberately allowed to evaporate into the air. However, the solution

normally is used only under an exhaust hood, which reduces the amounts of 14C released into

the breathing space of the workers.

In evaluating inhalation exposure for the situation described above, the materials licensee

(Anderson, 1994) stated that the solution normally contains 120 kilobecquerel (kBq)

(3.3 microcurie (Ci)) of 14C per mL and that 17 L of solution normally would be deposited in

each test sequence. Thus, for each deposition, the amount of 14C that would be released into

the air would be 2.1 kBq (0.056 Ci). By assuming a laboratory with dimensions of

4 m × 5 m × 6 m, an air ventilation rate of 20 volume changes per hour, 10 tests per hour, and

exhaustion of the released 14C to outside air with an efficiency of 90%, the licensee estimated

that the inhalation exposure to workers in the laboratory to 14C in the air would be 1.1 Bq/mL

(3×1011 Ci/mL) per hour (Anderson, 1994).

In considering the exposure scenario described above, the NRC judged that some of the

assumptions used by the materials licensee were not sufficiently conservative (NRC,

Memoranda, Paperiello, 1994). The NRC assumed instead that the release rate of 14C into the

air would be 8 kBq/h (0.22 Ci/h), the air exchange rate in the laboratory would be 4 per hour,

and the ventilation efficiency of hoods in the laboratory would be 70%. Based on these
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assumptions, the annual effective dose equivalent (EDE) to a worker in the laboratory as

estimated by the NRC was 0.07 mSv (7 mrem).

Information relevant to assessing dose from external exposure to quantities of gamma-emitting

radionuclides that are authorized for exempt distribution, is provided by a materials licensee

who manufactures gauging and calibration sources containing 60Co or 137Cs (Cahill, 1994). The

licensee stated that for an unshielded source containing a quantity of either 37 kBq (1 Ci) of

60Co or 370 kBq (10 Ci) of 137Cs, the external exposure rate is no greater than 0.008 C/kg-h

(0.03 mR/h) at a distance of 30 cm and 0.0008 C/kg-h (0.003 mR/h) at a distance of 1 meter. 

These exposure rates are consistent with the criterion of 0.26 C/kg-h (1 mR/h) at a distance of

10 cm used in establishing quantities for gamma-emitting radionuclides when the procedure of

logarithmic rounding to the nearest decade is taken into account (see Section 2.13.1).

2.13.4 Present Exemption Analysis

A rigorous quantitative assessment of the potential radiological impacts on the public from use

of quantities of byproduct material authorized for exempt distribution is a difficult undertaking,

primarily because the exemption does not specify any limits on the number of quantities of

byproduct material that may be used by any individual or groups of individuals. An additional

complicating factor is the variety of allowable uses of quantities of byproduct material.

In this assessment, simple scenarios for external and internal exposure are developed for

example materials containing quantities of byproduct material authorized for exempt

distribution. These scenarios are intended to provide reasonable upper bounds on doses that

might be experienced by individual members of the public from routine use or from accidents

and misuse. Although a rigorous assessment of individual dose is difficult, upper bound

estimates of dose can be based on the radiological criteria used to establish the quantities, as

described in Section 2.13.1.

Similarly, rigorous estimates of collective dose are difficult to obtain, because of the wide range

of radionuclides, physical and chemical forms, uses, and exposure conditions, as well as the

changing patterns of radionuclide use over time for this exemption. However, representative

collective doses can be estimated based on simple conservative assumptions regarding the

conditions of use for the radionuclides that have been distributed commercially to any

significant degree in recent years. In this assessment, estimates of collective dose are based

on reported commercial distributions of quantities of byproduct material. Most noncommercial

transfers should occur between specific licensees. Some of the materials commercially

distributed for use under this exemption also are used by specific licensees. In the case of this

particular exemption, licensed users are not exempt from the requirements of 10 CFR Part 20

with respect to these materials. Thus, exposures would be controlled under the requirements of

10 CFR Part 20. As a result, the collective doses calculated on the basis of total commercial

distribution should provide an overestimate of doses resulting from the exemption.

2.13.4.1 Individual Dose During Routine Use

In this section, three different approaches are presented to assessing individual dose from

routine use of quantities of byproduct material authorized for exempt distribution. In the first, an

assessment of dose from external exposure based directly on the stated criterion for
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determining the quantity of a photon-emitting radionuclide, as described in Section 2.13.1, is

presented. In the second, an assessment of dose from inhalation exposure based directly on

the intended use of 14C in a volatile solution, as described in Section 2.13.3, is presented. In

the third, simple scenarios for external and internal exposure are applied in estimating dose for

all quantities of byproduct material listed in Schedule B of 10 CFR 30.71.

2.13.4.1.1 Assessment of External Dose Based on Exemption Criterion

As an example of potential individual doses from external exposure to quantities of byproduct

material authorized for exempt distribution, exposure to a source that emits significant

intensities of high-energy photons is considered, such as might be used for calibration of

radiation detectors. As noted in Section 2.13.1, such calibration sources are one of the

intended uses of quantities of byproduct material authorized for exempt distribution, and they

are commonly used in teaching and research. Furthermore, such sources essentially are

unshielded and, thus, estimates of external dose in this case would provide an upper bound on

external dose from any shielded source.

As discussed in Section 2.13.1, the quantity for any radionuclide that emits sufficient intensities

of high-energy photons (e.g., 60Co and 137Cs) is that which results in an exposure of

0.26 C/kg-h (1 mR/h). By assuming that an exposure of 0.26 mC/kg (1 R) corresponds to an

EDE of 0.01 Sv (1 rem) to an exposed individual at the same location, this criterion can be used

to estimate dose from external exposure to the source. The assumed equivalence between

exposure and EDE is consistent with the approach used by the NRC (56 FR 23360) in

evaluating deep-dose equivalent (i.e., the dose equivalent at a depth in tissue of 1 cm) for

occupational exposure, which is used as a surrogate for EDE, and it provides a slightly

conservative estimate of EDE (e.g., by about 15 to 30% for a rotational irradiation geometry) for

photon energies above about 0.1 MeV (ICRP 51).

In this assessment, exposure to a single, unshielded calibration source is assumed to give an

EDE rate at a distance of 10 cm of 0.01 mSv/h (1 mrem/h). As noted above, this assumption

corresponds to the stated criterion for the exemption. It is then assumed that an individual is

located in the same room as the calibration source for 1000 h/yr, and that the average distance

between the source and the exposed individual is 2 meters. The exposure time is based on the

assumption that an individual spends about half of the normal working hours during a year in

the room containing the source, and the assumed distance from the source is intended to

represent an average distance in a typical laboratory. For an unshielded source of high-energy

photons, the dose rate varies approximately as the inverse of the square of the distance from

the source. Based on these assumptions, the resulting annual EDE from external exposure

would be 0.02 mSv (2 mrem).

The dose estimate given above should be somewhat conservative for the assumed exposure

time and distance from the source, because it does not take into account any shielding between

the source and receptor locations. For example, users of calibration sources often are provided

with safety instructions and proper equipment for storing sources in a shielded configuration

when they are not in use (Cahill, 1994). In addition, the assumed exposure time could be a

considerable overestimate for many realistic exposure situations, because calibration sources

often are stored in rooms (e.g., teaching laboratories) that individuals normally would occupy

only infrequently during the normal working year.
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On the other hand, rooms in which calibration sources are stored could be occupied on a

continuous basis for up to twice as long as the exposure time assumed in this analysis, and the

average distance of an individual from the source could be less than 2 meters. For an

exposure time of 2000 h/yr at an average distance of 1 meter, for example, the estimated dose

given above would be increased by a factor of 8, although such a higher dose should be

considerably less likely to occur. In addition, multiple sources containing quantities of

byproduct material authorized for exempt distribution could be located in the same room, in

which case the external dose would increase in proportion to the number of sources. This

situation also should be less likely to occur, because information provided to the user includes a

statement that exempt quantities should not be combined, although it cannot be assumed that

users will comply with this caution at all times.

Another factor also should be considered in the estimate of external dose given above. As

discussed in Section 2.13.1, quantities of byproduct material calculated on the basis of the

specified radiological criteria are logarithmically rounded to the nearest decade. Therefore, the

external dose corresponding to a quantity of a photon-emitting radionuclide could be as much

as a factor of 3 higher or lower than the dose calculated from the specified criterion on

exposure rate in air. For example, the external dose rates reported by Cahill (1994) for 60Co

and 137Cs and discussed in Section 2.13.3 are about a factor of 3 less than the criterion used in

establishing the quantities of byproduct material authorized for exempt distribution. This effect

is investigated in more detail in Section 2.13.4.1.3.

Considering all of the factors discussed above, the following conclusions about potential doses

to individuals from external exposure to quantities of byproduct material authorized for exempt

distribution appear warranted:

Based on reasonable assumptions about exposure conditions, the annual EDE could

range from 0.01 to 0.1 mSv (1 to 10 mrem).

Based on somewhat more pessimistic assumptions, which should be less likely to occur

but are nonetheless credible, the annual EDE could be on the order of a few tenths of

a mSv (a few tens of an mrem).

Based on quite pessimistic assumptions, which should occur only rarely, the annual

EDE could approach or exceed 1 mSv (100 mrem).

The highest doses could occur, for example, if multiple sources were stored without shielding in

occupied locations during a substantial portion of the year and at distances close to individuals.

2.13.4.1.2 Assessment of Internal Dose Based on Intended Use

As discussed in Section 2.13.1, potential inhalation or ingestion exposures to quantities of

byproduct material authorized for exempt distribution are inherently limited by two factors. The

first is the requirement that quantities of byproduct material authorized for exempt distribution

not be contained in any product or material designed for intake by, or application to, humans. 

The second is the use of MPCs in air for members of the public in defining the quantities of

byproduct material authorized for exempt distribution for radionuclides that are not significant

photon emitters.
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In this assessment, it is assumed that inhalation of quantities of byproduct material authorized

for exempt distribution in solid form and ingestion of quantities of byproduct material in any form

are unlikely during routine use. However, some byproduct materials (e.g., 3

H and 14C) often

occur in a volatile liquid form, and inhalation exposures could occur during normal use of such

materials.

A representative upper bound estimate of individual dose from inhalation exposure during

routine use of quantities of byproduct material authorized for exempt distribution is assumed to

be provided by a dose assessment for use of a chemical solution containing 14C, as described

in Section 2.13.3. In this case, the solution is deposited directly onto surfaces and the

deposited contents are deliberately evaporated into the air. However, the solution normally is

used only inside an exhaust hood in a laboratory, which would limit the activity of 14C released

into the room and, thus, the activity that could be inhaled by an individual working in the room.

Inhalation doses from exposure to a laboratory worker to 14C in a volatile solution are estimated

based on data provided by the materials licensee (Anderson, 1994). Specifically, the solution

contains 120 kBq/mL (3.3 Ci/mL) of 14C and, furthermore, 17 L of solution containing 2.1 kBq

(0.056 Ci) of 14C would be released into the air in each test sequence. The licensee expects

there would be no loss or spillage in transferring the solution from an ampule to the

microsyringe used to deposit the solution when the licensee’s instructions are followed. Finally,

the licensee assumed that a maximum of 10 test sequences per hour could be performed,

taking into account the time required for analysis and sample insertion. Therefore, the release

rate of 14C to the air assumed by the licensee, and adopted in this assessment, is 21 kBq/h

(0.56 Ci/h).

In the generic accident methodology presented in Appendix A.1, the assumptions are a typical

laboratory volume of 180 m3

 and an air ventilation rate of 6 volume changes per hour, and the

materials licensee (Anderson, 1994) assumed an exhaust efficiency for a laboratory hood of

90%. Based on these assumptions and the release rate of 14C to the air given above, the

concentration of 14C in the room air at steady state would be 1.9 Bq/m3

 (52 pCi). If it is then

assumed that the exposure is over a typical working year of 2000 hours (which should be a

conservative estimate of exposure time for most workers who would not spend all of their

working time in the laboratory), that the breathing rate of an individual is 1.2 m3

/h while working

in a laboratory, and that the inhalation dose coefficient for 14C is as provided in Table 2.1.2, the

resulting annual EDE from inhalation would be 0.003 mSv (0.3 mrem).

The dose estimate obtained in this assessment is somewhat lower than the estimate of about

0.07 mSv/yr (7 mrem/yr) obtained by the NRC (NRC, Memoranda, Paperiello, 1994), as

described in Section 2.13.3. The reason for this discrepancy, in part, is due to different

assumptions regarding laboratory volume and ventilation turnover rate. Other assumptions

used by the NRC in estimating dose, including the breathing rate and inhalation dose

coefficient, are not stated. In addition, the release rate of 14C to the air of 8 kBq/h (0.22 Ci/h)

assumed by the NRC, is less than the licensee’s value of 21 kBq/h (0.56 Ci/h) (Anderson,

1994), if the NRC’s value applies to 10 test sequences per hour. However, the number of test

sequences per hour assumed by the NRC was not given.

The dose assessment for an intended use of a quantity of byproduct material described above

should provide a reasonable upper bound for the inhalation dose to individuals from routine use

of other quantities of byproduct material authorized for exempt distribution, primarily because
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most other such materials are not expected to be used in a manner that would result in

deliberate releases into the air.

2.13.4.1.3 Example Assessment of External and Internal Dose

In this section, simple scenarios are assumed for the purpose of estimating individual doses

from external and internal exposure during routine use of the quantities authorized for exempt

distribution for all of the byproduct materials listed in Schedule B of 10 CFR 30.71. This

assessment has two purposes. First, it can be used to investigate the effects on estimates of

dose due to the use of logarithmic rounding to the nearest decade in determining the quantities

(see Section 2.13.1) and changes in dose coefficients for external and internal exposure since

the quantities were established. Although the quantities were intended to correspond

approximately to the same external or inhalation dose, depending on the criterion used to

establish the quantity for exempt distribution for any radionuclide, the effects of logarithmic

rounding and the newer dosimetry data could result in calculated external or inhalation doses

for assumed exposure scenarios that vary significantly among the different radionuclides. 

Second, it can be used to investigate the relative importance of external and internal exposure

during routine use. Such a comparison is not readily obtainable from the criteria for external

and inhalation exposure used in establishing the quantities. Note that for the purposes of this

comparison, no consideration has been given to radioactive decay; radionuclide activity is

assumed to remain constant over the exposure period. This assumption may result in

significant overestimates for radionuclides with short half-lives.

In estimating external dose from routine use of quantities of byproduct material authorized for

exempt distribution, the exposure scenario described in Section 2.13.4.1.1 is assumed. That is,

an individual is assumed to be exposed for 1000 h/yr at an average distance of 2 meters from

an unshielded source containing a quantity in Schedule B. For any radionuclide that emits

photons with energies predominantly above about 0.1 MeV, the EDE for this scenario is

estimated using the specific gamma-ray dose constant calculated by Unger and Trubey (1981),

as listed for some radionuclides in Table 2.1.2 of Section 2.1, and the assumption that the dose

rate varies inversely with the square of the distance from the source. As also discussed in

Section 2.13.4.1.1, the specific gamma-ray dose constant for these radionuclides is assumed to

correspond to the EDE. 

External dose is not estimated for radionuclides that emit photons with energies predominantly

less than about 0.1 MeV, because the specific gamma-ray dose constant in these cases would

substantially overestimate the EDE, especially if any shielding exists between the source and

receptor locations. Furthermore, the primary purpose of this part of the assessment is to

estimate external dose for those radionuclides for which the quantity was based on the criterion

for external exposure, and this is the case only for radionuclides that emit sufficient intensities

of higher energy photons.

The development of representative scenarios for inhalation or ingestion exposure during routine

use of quantities of byproduct material authorized for exempt distribution is rather arbitrary

compared with the case of external exposure, because the sources normally would be

contained and, furthermore, would not be incorporated in any product designed for intake by, or

application to, humans. Therefore, appreciable inhalation or ingestion exposures normally

would not be expected to occur during routine use, especially when the byproduct materials are

in a solid form.
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In this assessment, inhalation and ingestion doses during routine use of quantities of byproduct

material authorized for exempt distribution are estimated based on the following assumptions. 

First, the materials are assumed to be in a readily dispersible liquid or powder form, and they

are assumed to be used in an open container in a laboratory in such a way that the materials

easily could be inhaled or ingested. This assumption should result in overestimates of

inhalation and ingestion doses for most routine uses of quantities of byproduct material

authorized for exempt distribution.

Second, the generic accident methodology for spills of liquids or powders in Appendix A.1 is

assumed to be appropriate for estimating inhalation dose from routine exposure to quantities of

byproduct material in liquid or powder form. Thus, for each working day, 0.1% of the material is

assumed to be released into the air and 0.1% of the released material is assumed to be

inhaled; i.e., 106

 of a quantity is assumed to be inhaled per day. For a working year of

250 days, the fraction of a quantity inhaled is assumed to be 2.5×104

.

Third, for each working day, 106

 of a quantity also is assumed to be ingested. In the generic

accident methodology for spills of liquids or powders in Appendix A.1, the fraction of the amount

of a spilled liquid or powder ingested is assumed to be 104

, based on assumptions that 10% of

the material would be deposited on an individual’s skin and 0.1% of the deposited material

would be ingested. However, for routine use of liquids or powders, the amount of the available

material deposited on an individual’s skin presumably would be considerably less than 10%,

given the considerable care that normally would be taken in handling the material, and it is

arbitrarily assumed this fraction is 0.1%. For a working year of 250 days, the fraction of a

quantity ingested is assumed to be 2.5×104

, the same as for inhalation.

Finally, in estimating dose from inhalation or ingestion exposure, the dose coefficients from

EPA–520/1–88–020 are used. For inhalation, the highest dose coefficient for any clearance

class is chosen and, for ingestion, the highest dose coefficient for any uptake fraction from the

gastrointestinal tract is chosen. However, inhalation and ingestion exposure is assumed not to

occur for radionuclides in the form of noble gases.

For the radionuclides listed in Schedule B of 10 CFR 30.71, the results of the dose assessment

based on the exposure scenarios described above are given in Table 2.13.2. Bearing in mind

that this assessment may not provide realistic estimates of dose from exposure to quantities of

byproduct material authorized for exempt distribution, especially for inhalation and ingestion

exposures, these results may be summarized as follows.

First, for radionuclides that emit significant intensities of high-energy photons, the estimated

dose from external exposure generally exceeds the estimated dose from inhalation or ingestion

exposure by about an order of magnitude or more. Therefore, given that the assumed

scenarios for inhalation and ingestion exposure during routine use should be conservative

compared with the scenario for external exposure, potential doses from external exposure to

quantities of byproduct material authorized for exempt distribution apparently are substantially

higher than potential doses from inhalation or ingestion.

Second, for radionuclides for which the estimated dose from external exposure is higher than

the estimated dose from inhalation or ingestion, the external dose varies by about three orders

of magnitude, depending upon the particular radionuclide, from about 2×104

 mSv/yr

(0.02 mrem/yr) to about 0.2 mSv/yr (20 mrem/yr). Doses from the middle to the upper end of
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this range correspond to the best estimates (i.e., reasonable assumptions) for high-energy,

photon-emitting radionuclides discussed in Section 2.13.4.1.1. This large variability presumably

reflects a number of factors, including the calculation of quantities using logarithmic rounding to

the nearest decade, changes in external dosimetry data since the quantities were established,

and the possibility that the quantities for radionuclides with external doses toward the low end of

this range were based on the criterion for inhalation exposure rather than external exposure.

The one notable exception to the range of external doses given above occurs for the

positron-emitting radionuclide 18F. In this case, the estimated upper bound external dose is

2 mSv/yr (200 mrem/yr). It can be speculated that in establishing the quantity in Schedule B for

this radionuclide, the AEC may not have considered the significant contribution to external dose

from the 0.511-MeV photons produced by annihilation of the emitted positrons at rest and,

therefore, the quantity may have been based inappropriately on the criterion for inhalation

exposure. However, considering its 110-minute half-life, any actual dose from use of a quantity

of this radionuclide from Schedule B during the course of a year should be considerably less.

Finally, for radionuclides for which the estimated dose from inhalation or ingestion exposure is

higher than the estimated dose from external exposure, the internal dose also varies by about

three orders of magnitude, depending upon the particular radionuclide, from about

2×106

 mSv/yr (2×104 mrem/yr) to about 3×103 mSv/yr (0.3 mrem/yr). The exceptions include

the alpha-emitting radionuclide 241Am, for which the estimated inhalation dose is 0.06 mSv/yr

(6 mrem/yr), and 115In, for which the estimated inhalation dose is 0.09 mSv/yr (9 mrem/yr). In

the case of 115In, a quantity for exempt distribution corresponds to a mass of 1400 kg, an

amount no one would be expected to have in practice. The observed variability in internal

doses among the different radionuclides appears reasonable, given the calculation of quantities

using logarithmic rounding to the nearest decade and the significant differences in dose

coefficients for many radionuclides between those used by the AEC (ICRP 2) and those used in

the present assessment (EPA–520/1–88–020).

2.13.4.2 Collective Dose During Routine Use

It is difficult to obtain realistic estimates of collective dose during routine use of quantities of

byproduct material authorized for exempt distribution, due primarily to the variety of materials

and conditions of use. However, representative estimates of collective dose can be obtained

based on assumed scenarios. The following paragraphs provide example assessments of

collective dose from external and inhalation exposure, based on credible uses of quantities of

byproduct material authorized for exempt distribution.

2.13.4.2.1 Collective Dose from External Exposure

In Section 2.13.4.1.1, it was estimated that external exposure to a single calibration source

containing a quantity from Schedule B of a high-energy, photon-emitting radionuclide would

result in a nominal annual EDE to an individual of 0.02 mSv (2 mrem). The collective dose

during routine use can be estimated from this individual dose and assumptions about the

annual distribution of sources and their useful lifetime. It is assumed for this assessment that

10,000 sources are distributed annually. Also, assuming an individual is exposed for 1000 h/yr

at 2 meters is reasonably conservative and bounds potential exposure to other individuals in the

work area. If each source contains a quantity from Schedule B, this assumption is roughly
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consistent with the distribution data in Table 2.13.1 for the important photon-emitting

radionuclides 60Co and 137Cs. 

The useful lifetime of a calibration source depends on the half-life of the radionuclide. To

provide a reasonable upper bound on the collective dose, it is assumed, consistent with the

distribution data in Table 2.13.1, that most of the sources are 137Cs and that the sources have a

useful lifetime of 30 years (i.e., about one half-life of the radionuclide). Based on these

assumptions, and taking into account radioactive decay over 30 years, the resulting collective

EDE during routine use from 1 year’s distribution of sources would be 0.6 person-Sv

(60 person-rem). If the actual EDE for external exposure to 137Cs was 0.009 mSv/yr

(0.9 mrem/yr) from Table 2.13.2, the collective dose would be 0.2 person-Sv (20 person-rem).

Based on the distribution data in Table 2.13.1, it appears that 60Co and 137Cs are by far the most

important photon-emitting radionuclides distributed under this exemption in regard to the

potential collective dose from external exposure. This conclusion is based on the total activities

distributed for the different photon-emitting radionuclides and their half-lives (i.e., expected

useful lifetimes). Therefore, the estimate of collective dose given above should provide a

reasonable upper bound for the collective dose from external exposure for the recent exempt

commercial distribution of all quantities of byproduct material.

2.13.4.2.2 Collective Dose From Internal Exposure

In Section 2.13.4.1.2, it was estimated that inhalation exposure to 14C contained in a solution

that is deliberately evaporated into the air would result in an annual EDE 0.003 mSv (0.3 mrem)

to a laboratory worker. This estimate resulted from an assumed use of 41 MBq (1.1 mCi) of 14C

in a single laboratory. Based on the annual distribution of 14C in the solution of 150 MBq

(4 mCi) reported by a single licensee (Anderson, 1994), individuals would be exposed in only

about four laboratories. If it is also assumed, based on the intended use of the solution, that all

of the solution would be used during the year in which it is distributed, the resulting collective

EDE during routine use from1 year’s distribution of 14C would be 9×106

 person-Sv

(9×104

 person-rem).

Based on the distribution data in Table 2.13.1, the annual distribution of 14C by a single licensee

given above is only a small fraction of the total distribution of this radionuclide. If it is assumed

from the data in Table 2.13.1 that the total annual distribution of 14C is 680 GBq (18 Ci) and that

the collective dose for this distribution can be obtained by linear scaling of the estimated

collective dose for the 14C solution given above, the estimated collective EDE during routine use

from 1 year’s distribution of all 14C would be 0.04 person-Sv (4 person-rem). However, this

estimate should be quite conservative, because most of the 14C distributed as quantities of

byproduct material authorized for exempt distribution presumably is not deliberately evaporated

into the air during normal use.

Based on the distribution data in Table 2.13.1, it appears that 14C is by far the most important

nonphoton-emitting radionuclide distributed under this exemption in regard to the potential

collective dose from inhalation exposure. Relatively large quantities of 3

H also have been

distributed, but the amounts are substantially less than for 14C. Furthermore, the inhalation

dose per unit exposure for 3

H is considerably less than for 14C in the form of labeled organic

compounds (EPA–520/1–88–020). Therefore, the estimate of collective dose given above
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should provide an upper bound for the collective dose from inhalation exposure for the recent

exempt commercial distribution of all quantities of byproduct material.

Based on the results of this assessment, it appears that the collective dose from routine use of

quantities of byproduct material authorized for exempt distribution that are not significant

photon-emitters (i.e., radionuclides for which only inhalation exposure would be important)

should be substantially less than the collective dose from routine use of radionuclides for which

the criterion for external exposure provided the basis for the quantity in Schedule B. This

conclusion is consistent with the results of the individual dose assessments in Section 2.13.4.1.

2.13.4.3 Distribution and Transport

During routine distribution and transport of quantities of byproduct material authorized for

exempt distribution, external exposure normally would be the only pathway of concern, except

for 3

H where the normal leakage of volatile materials could result in inhalation exposure. 

Furthermore, because of the requirement in 10 CFR 32.19(a) that no more than 10 quantities

can be sold or transferred in any single transaction, no more than 10 quantities normally would

be included in a single shipment.

Based on these considerations, individual and collective doses during distribution and transport

of quantities of byproduct material authorized for exempt distribution are estimated using the

generic methodology in Appendix A.3. It is assumed that quantities of byproduct material

authorized for exempt distribution are shipped primarily by ground parcel delivery. A local

parcel-delivery driver is assumed to pick up the radioactive materials from the distributor and

transport them to a local terminal for shipment to customers. It is further assumed that semitrucks are used to transport the materials between local terminals, and that the materials are

transported to an average of four regional terminals before delivery to the customer. The

radiation doses to workers at local and regional terminals are assumed to be the same as those

estimated for workers at a large warehouse.

In estimating individual dose during distribution and transport, it is assumed that the same

parcel-delivery driver picks up all radioactive materials from the distributor and transports them

to the first local terminal. This should provide a conservative estimate of individual dose, but it

also takes into account the limited number of distributors of quantities of byproduct material

authorized for exempt distribution and the few drivers normally used by local parcel-delivery

services. The dose to individual workers at the first local terminal would be less than the dose

to the local truck driver (see Appendix A.3). Furthermore, once the radioactive materials are

dispersed throughout the regional and national distribution system, the dose to individual truck

drivers and terminal workers would decrease substantially compared with the doses during the

first stage of distribution and transport.

In estimating the external dose to an individual truck driver who picks up quantities of byproduct

material from the distributor, it is assumed that the radionuclides shipped are 60Co or 137Cs. As

discussed previously and indicated in Table 2.13.1, these are the most important

photon-emitting radionuclides distributed as quantities of byproduct material authorized for

exempt distribution. Based on the data over a 6-year period given in Table 2.13.1, an annual

distribution of 28 MBq (0.75 mCi) of 60Co and 1.2 GBq (32 mCi) of 137Cs is assumed. These

annual distributions are roughly consistent with an assumption that a single driver would deliver

10 quantities each day for 250 working days during the year, which suggests that the estimates
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of individual dose obtained using these assumptions, while conservative, are not extreme. 

Using the results in Table A.3.2 of Appendix A.3 for a small express-delivery truck under

conditions of average exposure, the annual EDE to a truck driver who is assumed to deliver an

entire year’s exempt distribution of quantities of byproduct material would be 2×103

 mSv

(0.2 mrem) for 60Co and 2×102

 mSv (2 mrem) for 137Cs. Use of the results in Table A.3.1 of

Appendix A.3, assuming that packages in all shipments during the year would be located close

to the driver, is presumed to be unreasonable.

An upper bound estimate of the inhalation dose to an individual truck driver from distribution

and transport of quantities of byproduct material authorized for exempt distribution of 3

H can be

obtained by assuming, based on the data in Table 2.13.1, that there is an annual distribution of

110 GBq (3 Ci) and that a single driver would be involved in all shipments from the distributor. 

Using the results in Table A.3.1 or A.3.2 of Appendix A.3 for a small express-delivery truck, the

annual EDE to a truck driver would be 6×105

 mrem (0.006 mrem). The calculated dose from

inhalation exposure to 3

H is much less than the estimated doses from external exposure to 60Co

and 137Cs.

As described previously, the collective dose from distribution and transport of quantities of

byproduct material authorized for exempt distribution is estimated by assuming two shipments

in a small express-delivery truck (i.e., the initial pickup from the distributor and the final delivery

to the customer), three shipments between terminals in a semi-truck, and temporary storage in

four terminals (i.e., large warehouses). For the assumed annual distributions of 60Co and 137Cs

given above and using the results in Tables A.3.3 and A.3.8 of Appendix A.3, the estimated

annual collective EDEs are 5×105

 person-Sv (5×103 person-rem) for 60Co and

5×104

 person-Sv (5×102 person-rem) for 137Cs. Most of the collective dose would be received

by terminal workers. The collective dose from inhalation exposure for 1 year’s distribution of 3

H

would be much less.

Thus, in summary, based on the reviewed data for commercial distribution of quantities of

byproduct material authorized for exempt distribution , the dose during distribution and transport

would be due almost entirely to the distribution of 60Co and 137Cs. The following dose estimates

are obtained:

The annual EDE to individual truck drivers during the initial pickup of quantities of

byproduct material from the distributor would be about 0.02 mSv (2 mrem). The dose to

other individuals would be considerably less.

The annual collective EDE to truck drivers and terminal workers, most of which would be

received by terminal workers, would be about 5×104

 person-Sv (5×102 person-rem).

The contributions to individual and collective dose from other photon-emitting radionuclides that

have been distributed as quantities of byproduct material authorized for exempt distribution and

from 3

H, which could result in inhalation exposure during distribution and transport, would be

negligible by comparison.

2.13.4.4 Disposal

Individual and collective doses from disposal of byproduct material are estimated using the

generic methodology described in Appendix A.2. Disposal in landfills and by incineration is
2–176

assumed to occur, with 80% of all disposals going to landfills and 20% to incinerators. Users of

quantities of byproduct material authorized for exempt distribution may be cautioned that these

quantities are not meant to be incinerated (Cahill, 1994). Recycling is assumed not to occur. 

Doses are estimated for disposal of 14C, 60Co, and 137Cs. As indicated by the data in

Table 2.13.1, these are the most important radionuclides that have been distributed recently

and, therefore, should be the most important in regard to doses from disposal of all quantities of

byproduct material authorized for exempt distribution. Based on the 6 year distribution data in

Table 2.13.1, the annual distributions of these radionuclides are assumed to be 680 GBq (18

Ci) for 14C, 28 MBq (0.75 mCi) for 60Co, and 1.2 GBq (0.033 Ci) for 137Cs. For 60Co and 137Cs,

the dose estimates for disposal take into account radioactive decay, based on an assumption

that disposal occurs at 1 half-life after distribution.

2.13.4.4.1 Disposal in Landfills

Based on the generic methodology in Appendix A.2, the following estimates are obtained of

individual and collective dose from disposal in landfills of the quantities of byproduct materials

described above.

For 14C, the annual EDE to individual waste collectors would be 2×105

 mSv (0.002 mrem), and

the annual doses to individual landfill workers or other members of the public would be

considerably less. The collective EDE from 1 year’s disposals would be 0.009 person-Sv

(0.9 person-rem), due almost entirely to exposure to off-site residents from releases to

groundwater over 1000 years after disposal.

For 60Co, the annual EDE to individual waste collectors would be less than 1×105

 mSv

(<0.001 mrem), and the annual doses to individual landfill workers or other members of the

public would be considerably less. The collective EDE from 1 year’s disposals would be

3×105

 person-Sv (0.003 person-rem), due almost entirely to exposure to waste collectors and

workers at landfills.

For 137Cs, the annual EDE to individual waste collectors would be 6×105

 mSv (0.006 mrem),

and the annual doses to individual landfill workers or other members of the public would be

considerably less. The collective EDE from 1 year’s disposals would be 0.001 person-Sv

(0.1 person-rem), due primarily to exposure to future on-site residents over 1,000 years after

disposal. If exposure to future on-site residents were ignored, the collective EDE would be 

2×104 person-Sv (0.02 person-rem), due almost entirely to exposure to waste collectors and

landfill workers.

Thus, in summary, based on recent data on the exempt commercial distribution of byproduct

materials, the dose from disposal in landfills would be due almost entirely to the distribution of

14C, 60Co, and 137Cs. The following dose estimates are obtained:

The annual EDE to individuals, i.e., waste collectors, would be about 9×105

 mSv

(0.009 mrem), due primarily from disposals of 137Cs.

The collective EDE from 1 year’s disposals, which would be received primarily by off-site

residents from releases of 14C to groundwater, would be about 0.009 person-Sv

(0.9 person-rem).
2–177

2.13.4.4.2 Disposal in Incinerators

Based on the generic methodology in Appendix A.2, the following estimates are obtained of

individual and collective dose from disposal in incinerators of the quantities of byproduct

materials described at the beginning of Section 2.13.4.4.

For 14C, the annual EDE to individual waste collectors would be 9×105

 mSv (0.009 mrem), and

the annual doses to individual incinerator workers or other members of the public would be

considerably less. The collective EDE from 1 year’s disposals would be 1×105

 person-Sv

(1×103

 person-rem), due almost entirely to exposure to waste collectors.

For 60Co, the annual EDE to individual waste collectors would be less than 1×105

 mSv

(<0.001 mrem), and the annual doses to individual incinerator workers or other members of the

public would be considerably less. The collective EDE from 1 year’s disposals would be less

than 1×105

 person-Sv (<0.001 person-rem).

For 137Cs, the annual EDE to individual waste collectors would be 4×104

 mSv (0.04 mrem), and

the annual doses to individual incinerator workers or other members of the public would be

considerably less. The collective EDE from 1 year’s disposals would be 5×105

 person-Sv

(0.005 person-rem), due almost entirely to exposure to waste collectors.

Thus, in summary, based on the reviewed data on the exempt commercial distribution of

byproduct materials, the dose from disposal in incinerators would be due almost entirely to the

distribution of 14C, 60Co, and 137Cs. The following dose estimates are obtained:

The annual EDE to individuals, i.e., waste collectors, would be about 5×104

 mSv

(0.05 mrem), primarily from disposals of 137Cs.

The collective EDE from 1 year’s disposals, which would be received primarily by waste

collectors, would be about 6×105

 person-Sv (0.006 person-rem), primarily from disposal

of 137Cs.

2.13.4.4.3 Additional Disposal Considerations

Particularly in the case of quantities of byproduct material authorized for exempt distribution in

liquid form, a substantial portion of the material could be disposed in sanitary sewers. An

assessment of doses resulting from such disposals is beyond the scope of this study, but

exposure pathways resulting from disposal of radioactive materials into sanitary sewers have

been evaluated elsewhere (NUREG/CR–5814).

2.13.4.5 Accidents and Misuse

Potential doses from accidents and misuse involving byproduct material are inherently limited

by the conditions of the exemption and the criteria used to establish the quantities of byproduct

material authorized for exempt distribution, i.e., the use of MPCs in air for members of the

public or an external dose rate of about 0.01 mSv/h (1 mrem/h) at a distance of 10 cm from a

point source, both of which result in low values. As examples, three scenarios are considered

that should bound potential doses from accidents or misuse, involving external exposure to a

worker, internal exposure in a laboratory, and a transportation accident.
2–178

First, a scenario for misuse is considered in which an individual inadvertently places a

calibration source containing a quantity of a photon-emitting radionuclide from Schedule B in a

shirt pocket. It is assumed that the individual might be exposed for 50 hours (i.e., about

16 h/day for 3 days) before the presence of the source would be discovered, and it is further

assumed that the source is located at an average distance of 10 cm from internal organs of the

body (Refer to Appendix A.4.) An assumption of exposure times considerably longer than

50 hours would not be reasonable for inadvertent misuse of a source, because of the

requirement in 10 CFR 32.18(c) that the source must be identified as radioactive and the low

likelihood that the presence of the source in the shirt pocket would go unnoticed for a long time. 

Based on these assumptions and the assumption that the external dose rate at a distance of

10 cm is 0.01 mSv/h (1 mrem/h), the resulting EDE would be 0.5 mSv (50 mrem) without any

consideration for attenuation. The actual EDE should be considerably less.

Second, an accident scenario is considered that involves a spill of a quantity of byproduct

material from Schedule B in liquid or powder form in a laboratory and subsequent ingestion and

inhalation exposure to an individual. For such a scenario, the results in Table A.1.8 of the

generic accident methodology in Appendix A.1 can be used to estimate dose. For example, for

a spill of a quantity of 37 MBq (1 mCi) of 3

H in liquid form, the resulting EDE from ingestion

would be 0.06 mSv (6 mrem), and the dose from inhalation would be much less. For a spill of a

quantity of 3.7 MBq (0.1 mCi) of 137Cs in powder form, the resulting EDE from ingestion would

be 5×104

 mSv (0.05 mrem), and the dose from inhalation would be about two orders of

magnitude less. In general, for radionuclides other than 3

H and 14C that normally would not

occur in liquid form, the internal dose from an accidental spill of a quantity for exempt

distribution would be considerably less than the value for 3

H obtained here.

Finally, an accident scenario is considered that involves a transportation fire and subsequent

inhalation and external exposure to an individual firefighter during the fire and cleanup after the

fire. For such a scenario, the results in Table A.1.4 of the generic accident methodology in

Appendix A.1 can be used to estimate dose. For example, a single parcel-service pickup from

a distributor is assumed to include 100 quantities of the radionuclides 3

H, 14C, 51Cr, and 137Cs. 

The data in Table 2.13.1 indicate these are the most important radionuclides distributed

recently under this exemption. Although each package for an individual customer may include

no more than 10 quantities (see Section 2.13.1), multiple packages for different customers

could be included in a single shipment. Using the quantities for these radionuclides in

Table 2.13.1 and the results in Table A.1.4, with 51Cr evaluation performed using methodology

described in Appendix A.1, the following estimates of EDEs are obtained: 8×105

 mSv

(0.008 mrem) for 3

H and less than 1×105 mSv (<0.001 mrem) for 14C, 51Cr, and 137Cs.

The results described above clearly indicate the limited potential for high doses from accidents

and misuse of quantities of byproduct material authorized for exempt distribution. High doses,

e.g., on the order of 10 mSv (1 rem) or greater, from external exposure to photon-emitting

radionuclides could occur only if extreme and highly unlikely exposure times of about

1000 hours or greater are assumed. For internal exposure, there do not appear to be any

credible scenarios that would result in such high doses. For example, even assuming

inadvertent ingestion of an entire quantity of 37 MBq (1 mCi) of 3

H, which is an extreme

exposure scenario, the resulting EDE would only be 0.6 mSv (60 mrem).
2–179

2.13.5 Summary

In this assessment, estimates of individual and collective dose to the public from routine use,

distribution and transport, and disposal of byproduct material authorized for distribution were

obtained based on the radiological criteria used to define the quantities and recent information

on the amounts of the most important radionuclides that have been distributed. Doses from

accidents and misuse of these quantities also were considered.

In all dose assessments, scenarios for internal exposure to nonphoton-emitting radionuclides

were considered separately from scenarios for external exposure to photon-emitting

radionuclides. In general, for any scenario, potential doses from inhalation or external exposure

are inherently limited by the low values of the quantities.

The results of this assessment are summarized in Table 2.13.3. Except for the nominal

estimates of individual and collective dose from external exposure to photon-emitting

radionuclides during routine use, the estimated doses are intended to represent credible upper

bounds. Based on this assessment, the following general conclusions about radiological

impacts on the public associated with this exemption can be made:

During most routine uses of quantities of byproduct material authorized for exempt

distribution, individual and collective doses from external exposure to photon-emitting

radionuclides should be considerably higher than doses from inhalation of nonphotonemitting radionuclides. Even in cases where byproduct materials are deliberately

released into the air during routine use, potential inhalation doses appear to be

somewhat less than potential doses from external exposure to photon-emitting

radionuclides.

Individual and collective doses should be higher during routine use of quantities of

byproduct material authorized for exempt distribution than during distribution and

transport or following disposal. Although the individual dose during distribution and

transport in Table 2.13.3 is essentially the same as the individual doses during routine

use from external and inhalation exposure, the dose estimate for distribution and

transport is based on the conservative assumption that a single truck driver would be

exposed to an entire annual distribution of the most important photon-emitting

radionuclide. However, the doses for routine use are intended to be nominal best

estimates and the dose from external exposure could be considerably higher if more

pessimistic, but nonetheless credible, assumptions are used, such as exposure to

multiple sources combined with quite pessimistic and highly unlikely assumptions about

exposure conditions could result in annual dose approaching or exceeding 1 mSv

(100 mrem) (see Section 2.13.4.1.1).

There do not appear to be any credible scenarios for accidents and misuse of quantities

of byproduct material authorized for exempt distribution that could result in doses

exceeding about 0.5 mSv (50 mrem). Especially for accidents involving inhalation or

ingestion exposure, doses are inherently limited by the low values of the quantities. For

scenarios involving external exposure to photon-emitting radionuclides, doses of several

tens of mSv (several rem) could be obtained by assuming inadvertent exposure to a

source next to the body for thousands of hours. However, since current regulations
2–180

require these quantities of byproduct material to be identified as radioactive, such a

scenario could occur only as a result of deliberate misuse. Exposure times of no more

than a few tens of hours are more reasonable for scenarios involving inadvertent misuse

of photon-emitting sources.

This assessment has indicated that potential external doses during routine use of byproduct

material containing photon-emitting radionuclides are particularly important. Based on the

criterion that the external exposure rate from a quantity of byproduct material should not exceed

0.26 C/kg-h (1 mR/h) at a distance of 10 cm from an unshielded point source and taking into

account that the quantities were obtained using logarithmic rounding to the nearest decade, the

following results were obtained:

The annual EDE to individuals from exposure to a single source could range from 0.01

to 0.1 mSv (1 to 10 mrem), based on reasonable assumptions about the exposure time,

average distance from the source, and amount of shielding present.

The annual EDE from exposure to a single source could be a few tenths of a mSv (a few

tens of a mrem), based on more pessimistic assumptions about the exposure time and

average distance from the source that would be less likely to occur but are nonetheless

credible for routine exposure situations.

The annual EDE from exposure to multiple sources could approach or exceed 1 mSv

(100 mrem), based on quite pessimistic assumptions which should rarely occur if

multiple sources are stored without shielding in occupied locations during a substantial

portion of the year and at distances close to individuals.

However, concerns about the magnitude of potential external doses during routine use of

quantities of byproduct material authorized for exempt distribution, containing photon-emitting

radionuclides may be mitigated by the following circumstances. First, given the requirements

for labeling of containers for quantities as radioactive material and the precautions that normally

would be taken by users of radioactive material, including shielding in containers and storage at

locations away from individuals when not in use, it is highly unlikely that quantities of byproduct

material authorized for exempt distribution containing photon-emitting radionuclides would be

used or stored in unshielded configurations in close proximity to individuals for most of a

working year.

Second, many of the institutions using multiple quantities of byproduct material would be

licensed by the NRC or an Agreement State to possess byproduct material in amounts

exceeding quantities of byproduct material authorized for exempt distribution. Therefore,

exposure to individuals who work in such institutions would be monitored routinely, and

excessive doses from external exposure to these quantities of byproduct material would be

detected and appropriate actions taken to reduce unwarranted exposures. Exposures to other

members of the public would not be monitored, but their exposure times and doses should be

less than those for workers at licensed facilities who are more likely to be exposed over

substantial portions of a year.
2–181

Table 2.13.1 Total Reported Quantities of Radionuclides in Byproduct Material

Distributed During 1989 to 1995 a

Radionuclide

Total Activity

Distributedb

(Ci)c

Quantity Authorized

for Exempt Distributiond

(Ci)c

3

H 19 1,000

14C 110 100

32P 0.083 10

35S 0.015 100

51Cr 2.3 1,000

55Fe 0.0065 100

57Co 0.031 0.1e

60Co 0.0045 1

125I 0.036 1

137Cs 0.20 10

204Tl 0.0084 10

a

 Estimates based on review of materials licensee transfer reports sent to the NRC (NRC,

Licensee Reports, Material Transfer Reports, 1985-1995). For radionuclides not listed,

reported quantity distributed was less than 37 MBq (<1 mCi). Data provide indication of relative

importance of different radionuclides authorized for exempt commercial distribution in recent

years, but do not account for noncommercial transfers.

b

 Total activity reported for all products or materials.

c

 1 Ci = 37 GBq; 1 Ci = 37 kBq.

d

 Value from Schedule B of 10 CFR 30.71.

e

 Default value for any byproduct material, other than alpha-emitting byproduct material, not

listed in Schedule B of 10 CFR 30.71.
2–182

Table 2.13.2 Estimates of Potential Radiation Doses From External, Inhalation, and

Ingestion Exposure to Quantities of Byproduct Material Authorized for Exempt Distribution

Radionuclide

Qa

(Ci)

External Doseb

(rem/yr)

Inhalation Dosec

(rem/yr)

Ingestion Dosed

(rem/yr)

Americium-241 0.05e f

6×103 5×105

Antimony-122 100 8×103 1×104 2×104

Antimony-124 10 3×103 6×105 3×105

Antimony-125g 10 1×103 4×105 9×106

Arsenic-73 100

f

9×105 2×105

Arsenic-74 10 1×103 2×105 1×105

Arsenic-76 10 7×104 9×106 1×105

Arsenic-77 100 2×104 3×105 3×105

Barium-131 10 1×103 2×106 5×106

Barium-133 10 1×103 2×105 9×106

Barium-140g 10 4×103 2×105 4×105

Bismuth-210 1

f

5×105 2×106

Bromine-82 10 4×103 4×106 4×106

Cadmium-109 10

f

3×104 3×105

Cadmium-115m 10 3×105 2×104 4×105

Cadmium-115g 100 9×103 1×104 1×104

Calcium-45 10

f

2×105 8×106

Calcium-47g 10 2×103 2×105 2×105

Carbon-14 100

f

5×105 5×105

Cerium-141 100 2×103 2×104 7×105

Cerium-143 100 6×103 8×105 1×104

Cerium-144g 1 1×105 9×105 5×106

Cesium-131 1000

f

4×105 6×105

Cesium-134m 100 2×103 1×106 1×106

See end of table for footnotes.
2–183

Table 2.13.2 Estimates of Potential Radiation Doses From External, Inhalation,

and Ingestion Exposure to Quantities of Byproduct Material

Authorized for Exempt Distribution (continued)

Radionuclide

Qa

(Ci)

External Doseb

(rem/yr)

Inhalation Dosec

(rem/yr) 

Ingestion Dosed

(rem/yr)

Cesium-134 1 2×104 1×105 2×105

Cesium-135 10

f

1×105 2×105

Cesium-136 10 3×103 2×105 3×105

Cesium-137g 10 9×104 8×105 1×104

Chlorine-36 10

f

5×105 8×106

Chlorine-38 10 2×103 3×107 6×107

Chromium-51 1000 6×103 8×105 4×105

Cobalt-57h 0.1 4×106 2×107 3×108

Cobalt-58m 10

f

2×107 2×107

Cobalt-58 10 2×103 3×105 9×106

Cobalt-60 1 3×104 5×105 7×106

Copper-64 100 3×103 7×106 1×105

Dysprosium-165 10 6×105 3×107 9×107

Dysprosium-166g 100 2×103 3×104 3×104

Erbium-169 100

f

5×105 4×105

Erbium-171 100 7×103 1×105 4×105

Europium-152m 100 5×103 2×105 5×105

Europium-152 1 2×104 6×105 2×106

Europium-154 1 2×104 7×105 2×106

Europium-155 10 2×104 1×104 4×106

Fluorine-18 1000 2×101 2×105 3×105

Gadolium-153 10 4×104 6×105 3×106

Gadolium-159 100 1×103 2×105 5×105

Gallium-72 10 4×103 5×106 1×105

See end of table for footnotes.
2–184

Table 2.13.2 Estimates of Potential Radiation Doses From External, Inhalation,

and Ingestion Exposure to Quantities of Byproduct Material

Authorized for Exempt Distribution (continued)

Radionuclide

Qa

(Ci)

External Doseb

(rem/yr)

Inhalation Dosec

(rem/yr)

Ingestion Dosed

(rem/yr)

Germanium-71 100

f

3×106 2×107

Gold-198 100 7×103 8×105 1×104

Gold-199 100 2×103 4×105 4×105

Hafnium-181 10 1×103 4×105 1×105

Holmium-166 100 6×104 8×105 1×104

Hydrogen-3 1000

f

2×105 2×105

Indium-113m 100 6×103 1×106 3×106

Indium-114mg 10 4×104 2×104 4×105

Indium-115m 100 5×103 3×106 9×106

Indium-115i 10

f

9×103 4×104

Iodine-125 1

f

6×106 1×105

Iodine-126 1 1×104 1×105 2×105

Iodine-129 0.1

f

4×106 7×106

Iodine-131 1 7×105 8×106 1×105

Iodine-132 10 4×103 1×106 2×106

Iodine-133 1 1×104 1×106 3×106

Iodine-134 10 4×103 3×107 6×107

Iodine-135g 10 2×103 3×106 6×106

Iridium-192 10 1×103 7×105 1×105

Iridium-194 100 2×103 7×105 1×104

Iron-55 100

f

7×105 2×105

Iron-59 10 2×103 4×105 2×105

Krypton-85 100 4×103 j k

Krypton-87 10 1×103 j k

See end of table for footnotes.
2–185

Table 2.13.2 Estimates of Potential Radiation Doses From External, Inhalation,

and Ingestion Exposure to Quantities of Byproduct Material

Authorized for Exempt Distribution (continued)

Radionuclide

Qa

(Ci) 

External Doseb

(rem/yr)

Inhalation Dosec

(rem/yr) 

Ingestion Dosed

(rem/yr)

Lanthanum-140 10 3×103 1×105 2×105

Lutetium-177 100 7×104 6×105 5×105

Manganese-52 10 5×103 1×105 2×105

Manganese-54 10 1×103 2×105 7×106

Manganese-56 10 2×103 9×107 2×106

Mercury-197m 100 2×103 3×105 5×105

Mercury-197 100 2×103 2×105 2×105

Mercury-203 10 6×104 2×105 3×105

Molybdenum-99 g 100 6×103 1×104 1×104

Neodymium-147 100 3×103 2×104 1×104

Neodymium-149 100 8×103 6×106 1×105

Nickel-59 100

f

3×105 5×106

Nickel-63 10

f

8×106 1×106

Nickel-65 100 7×103 6×106 2×105

Niobium-93m 10

f

7×105 1×106

Niobium-95 10 1×103 1×105 6×106

Niobium-97 10 1×103 2×107 6×107

Osmium-185 10 1×103 3×105 6×106

Osmium-191m 100

f

8×106 1×105

Osmium-193 100 1×103 5×105 8×105

Palladium-103 100

f

4×105 2×105

Palladium-109 g 100 3×103 3×105 5×105

Phosphorus-32 10

f

4×105 2×105

Platinum-191 100 6×103 2×105 4×105

Platinum-193m 100 4×104 2×105 5×105

See end of table for footnotes.
2–186

Table 2.13.2 Estimates of Potential Radiation Doses From External, Inhalation,

and Ingestion Exposure to Quantities of Byproduct Material

Authorized for Exempt Distribution (continued)

Radionuclide

Qa

(Ci)

External Doseb

(rem/yr) 

Inhalation Dosec

(rem/yr)

Ingestion Dosed

(rem/yr)

Platinum-193 100

f

6×106 3×106

Platinum-197m 100 2×103 3×106 8×106

Platinum-197 100 5×104 1×105 4×105

Polonium-210 0.1

f

2×104 5×105

Potassium-42 10 4×104 3×106 3×106

Praseodymium-142 100 7×104 7×105 1×104

Praseodymium-143 100

f

2×104 1×104

Promethium-147 10

f

1×104 3×106

Promethium-149 10 2×105 7×106 1×105

Rhenium-186 100 5×104 8×105 7×105

Rhenium-188 100 1×103 5×105 8×105

Rhodium-103m 100

f

1×107 3×109

Rhodium-105 100 1×103 2×105 4×105

Rubidium-86 10 1×104 2×105 2×105

Rubidium-87i 10

f

8×106 1×105

Ruthenium-97 100 1×102 1×105 2×105

Ruthenium-103 10 8×104 2×105 8×106

Ruthenium-105g 10 1×103 1×106 3×105

Ruthenium-106g 1 3×105 1×104 7×106

Samarium-151 10

f

7×105 1×106

Samarium-153 100 2×103 5×105 7×105

Scandium-46 10 3×103 7×105 2×105

Scandium-47 100 2×103 5×105 6×105

Scandium-48 10 5×103 1×105 2×105

Silicon-31 100 1×105 6×106 1×105

See end of table for footnotes.
2–187

Table 2.13.2 Estimates of Potential Radiation Doses From External, Inhalation,

and Ingestion Exposure to Quantities of Byproduct Material

Authorized for Exempt Distribution (continued)

Radionuclide

Qa

(Ci)

External Doseb

(rem/yr)

Inhalation Dosec

(rem/yr)

Ingestion Dosed

(rem/yr)

Selenium-75 10 2×103 2×105 2×105

Silver-105 10 1×103 m 1×105 5×106

Silver-110m 1 4×104 2×105 3×106

Silver-111 100 5×104 2×104 1×104

Sodium-24 10 5×103 3×106 4×106

Strontium-85 10 2×103 1×105 5×106

Strontium-89 1

f

1×105 2×106

Strontium-90g 0.1

f

3×105 4×106

Strontium-91g 10 2×103 4×106 8×106

Strontium-92 10 2×103 2×106 5×106

Sulfur-35 100

f

6×105 2×105

Tantalum-182 10 2×103 1×104 2×105

Technetium-96 10 5×103 6×105 7×106

Technetium-97m 100

f

1×104 3×105

Technetium-97 100

f

2×105 4×106

Technetium-99m 100 3×103 8×107 2×106

Technetium-99 10

f

2×105 4×106

Tellurium-125m 10

f

2×105 9×106

Tellurium-127mg 10 9×106 5×105 2×105

Tellurium-127 100 9×105 8×106 2×105

Tellurium-129mg 10 3×104 6×105 3×105

Tellurium-129 100 2×103 2×106 5×106

Tellurium-131mg 10 2×103 2×105 2×105

Tellurium-132g 10 4×103 2×105 2×105

Terbium-160 10 2×103 6×105 2×105

See end of table for footnotes.
2–188

Table 2.13.2 Estimates of Potential Radiation Doses From External, Inhalation,

and Ingestion Exposure to Quantities of Byproduct Material

Authorized for Exempt Distribution (continued)

Radionuclide

Qa

(Ci)

External Doseb

(rem/yr)

Inhalation Dosec

(rem/yr)

Ingestion Dosed

(rem/yr)

Thallium-200 100 2×102 1×105 2×105

Thallium-201 100 2×103 6×106 8×106

Thallium-202 100 9×103 2×105 4×105

Thallium-204 10

f

6×106 8×106

Thulium-170 10

f

7×105 1×105

Thulium-171 10

f

2×105 1×106

Tin-113g 10 1×103 3×105 8×106

Tin-125 10 4×104 4×105 3×105

Tungsten-181 10

f

4×107 9×107

Tungsten-185 10

f

2×106 5×106

Tungsten-187 100 8×103 2×105 7×105

Vanadium-48 10 4×103 3×105 2×105

Xenon-131m 1000 8×103 m j k

Xenon-133 100 2×103 m j k

Xenon-135 100 5×103 m j k

Ytterbium-175 100 8×104 4×105 4×105

Yttrium-90 10

f

2×105 3×105

Yttrium-91 10 5×106 1×104 2×105

Yttrium-92 100 4×103 2×105 5×105

Yttrium-93 100 1×103 5×105 1×104

Zinc-65 10 8×104 5×105 4×105

Zinc-69mg 100 7×103 2×105 4×105

Zinc-69 1000

f

1×105 2×105

Zirconium-93 10

f

8×104 4×106

Zirconium-95g 10 2×103 1×104 1×105

Zirconium-97g 10 2×103 1×105 2×105

See following page for footnotes.
2–189

Footnotes to Table 2.13.2

a

 Quantity listed in Schedule B of 10 CFR 30.71, except as noted; 1 Ci = 37 kBq.

b

 Estimated dose assumes exposure for 1,000 h/yr at average distance from source of 2 meters

(see Section 2.13.4.1.3). 1 rem = 0.01 Sv.

c

 Estimated dose assumes inhalation of 106 of a quantity per day for 250 day/yr (see

Section 2.13.4.1.3).

d

 Estimated dose assumes ingestion of 106 of a quantity per day for 250 day/yr (see

Section 2.13.4.1.3).

e

 Quantity is not listed in Schedule B of 10 CFR 30.71, but is defined for purposes of exemption

for ionizing radiation measuring instruments containing sources for internal calibration or

standardization in 10 CFR 30.15(a)(9)(iii).

f

 Radionuclide does not emit significant intensities of photons with energies of about 0.1 MeV or

greater, and external dose is not estimated in these cases (see Section 2.13.4.1.3).

g

 Contributions to dose from shorter lived decay products, which are assumed to be in activity

equilibrium with parent radionuclide, are included.

h

 Radionuclide is not listed in Schedule B of 10 CFR 30.71, but has been distributed in

significant quantities under this exemption (see Table 2.13.1); quantity is default value for any

byproduct material other than alpha-emitting byproduct material.

i

 Radionuclide has very low specific activity, and use of maximum quantity would be impractical

due to extremely large mass.

j

 Inhalation exposure is assumed not to occur during routine use of noble gases.

k

 Ingestion exposure is not relevant for noble gases.

m Value estimated from calculated exposure rate in air and ratio of effective dose equivalent to

exposure for rotational irradiation geometry of 0.8 (ICRP 51).
2–190

Table 2.13.3 Summary of Potential Radiation Doses to the Public From Use

of Quantities of Byproduct Material Authorized for Exempt Distribution a

Exposure Scenario

Individual Annual Effective

Dose Equivalent

(mrem)b

Collective Effective

Dose Equivalent

(person-mrem)b

Routine use

 External exposurec

 Inhalation exposuref

2d

0.3g

60e

<4h

 Distribution and transport 2i 0.05j

Disposal

 Landfills

 Incinerators

0.009k

0.05k

 0.9l

 0.006l

Accidents and misuse

 External exposurem

 Internal exposuren

 50

6

See following page for footnotes.
2–191

Footnotes to Table 2.13.3

a

 Except as noted, dose estimates are intended to represent credible upper bounds for most

important radionuclides authorized for exempt distribution.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 Exposure to occupants of laboratory housing calibration sources containing a quantity of

photon-emitting radionuclide authorized for exempt distribution.

d

 Nominal best estimate for exposure to single source; depending upon particular radionuclide,

annual dose from exposure to single source could be range from 0.01 to 0.1 mSv (1 to

10 mrem). More pessimistic, but relatively unlikely, assumptions about exposure time and

distance from single source could result in annual dose of a few tens of a mrem, and exposure

to multiple sources combined with quite pessimistic and highly unlikely assumptions about

exposure conditions could result in annual dose approaching or exceeding 1 mSv (100 mrem)

(see Section 2.13.4.1.1).

e

 Dose from 1 year’s distribution based on nominal estimate of individual dose from exposure to

single calibration source of 0.025 mSv/yr (2.5 mrem/yr) and assumptions that 10,000 calibration

sources are distributed annually and each source has useful lifetime of 30 years (see

Section 2.13.4.2).

f

 Exposure to workers in laboratory using solution containing 14C that is deliberately evaporated

into air. Doses for this scenario should be considerably higher than doses resulting from

routine use of other selected quantities that would not be released into air deliberately.

g

 Dose estimate based on the quantity of 14C in each solution ampule reported by materials

licensee (Anderson, 1994); the quantity per ampule is about 5% of quantity authorized for

exempt distribution.

h

 Upper bound estimate based on assumption that 680 GBq (18 Ci) of 14C per year is distributed

that would be released into air deliberately. Actual collective dose from assumed annual

distribution of 14C would be considerably less if most materials would not be deliberately

released into air.

i

 Dose to truck driver during initial pickup from distributor, based on assumed annual

distributions of byproduct materials (see Section 2.13.4.3).

j

 Dose to truck drivers and terminal workers, based on assumed annual distributions of

byproduct materials (see Section 2.13.4.3).

k

 Dose to waste collectors, based on assumed annual distributions of byproduct materials (see

Section 2.13.4.4).

l

 Dose from year’s disposals, based on assumed annual distributions of byproduct materials

(see Section 2.13.4.4).

m Exposure to individual who inadvertently places single calibration source containing a quantity

of photon-emitting radionuclide authorized for exempt distribution in shirt pocket.

n

 Exposure to individual who spills 3H in liquid form.

2–193

2.14 Self-Luminous Products

2.14.1 Introduction

In 10 CFR 30.19, any person is exempt from licensing requirements to the extent that person

receives, possesses, uses, transfers, owns, or acquires tritium (3

H), 85Kr, or 147Pm in

self-luminous products; however, the exemption does not apply to products intended primarily

for frivolous purposes or in toys and adornments. This exemption was proposed on June 21,

1968 (33 FR 9198), and issued as a final rule on June 6, 1969 (34 FR 9025).

Requirements for a license to manufacture, process, produce, or initially transfer self-luminous

products containing 3

H, 85Kr, or 147Pm and intended for use under 10 CFR 30.19 are contained

in 10 CFR 32.22. Conditions of licenses are also included in 10 CFR 32.25; these include

requirements for quality control, labeling, and reporting of transfers.

Section 32.22 specifies that the license applicant must demonstrate that the product will meet

certain safety criteria set forth in 10 CFR 32.23. These safety criteria are the primary factors in

controlling the radiation doses associated with this exemption and are described below:

In normal use and disposal of a single exempt unit, it is unlikely that the annual dose

from external exposure, or the annual dose commitment resulting from intake of

radioactive material, to a suitable sample of the group of individuals expected to be the

most highly exposed to radiation or radioactive material from the product will exceed

0.01 millisievert (mSv) (1 mrem) to the whole body, the head and trunk, active

blood-forming organs, the gonads, or the lens of the eye: 0.15 mSv (15 mrem) to the

hands and forearms, feet and ankles, or localized areas of the skin averaged over areas

no larger than 1 cm2

 ; and 0.03 mSv (3 mrem) to any other organs.

In normal handling and storage of the quantities of exempt units that are likely to

accumulate in one location during marketing, distribution, installation, and servicing of

the product, it is unlikely that the annual dose from external exposure, or the annual

dose commitment resulting from intake of radioactive material, to a suitable sample of

the group of individuals expected to be the most highly exposed to radiation or

radioactive material from the product will exceed 0.1 mSv (10 mrem) to the whole body,

the head and trunk, active blood-forming organs, the gonads, or the lens of the eye;

1.5 mSv (150 mrem) to the hands and forearms, feet and ankles, or localized areas of

the skin averaged over areas no larger than 1 cm2

; and 0.3 mSv (30 mrem) to any other

organs.

It is unlikely there will be a significant reduction in the effectiveness of the containment,

shielding, or other safety features of the product from wear and abuse that is likely to

occur in normal handling and use of the product during its useful life.

In use and disposal of a single exempt unit, or in handling and storage of the quantities

of exempt units that are likely to accumulate in one location during marketing,

distribution, installation, and servicing of the product, the probability is low that the

containment, shielding, or other safety features of the product would fail under such

circumstances that an individual would receive an external dose or internal dose
2–194

commitment in excess of 5 mSv (0.5 rem) to the whole body, the head and trunk, active

blood-forming organs, the gonads, or the lens of the eye; 75 mSv (7.5 rem) to the hands

and forearms, feet and ankles, or localized areas of the skin averaged over areas no

larger than 1 cm2

, and 15 mSv (1.5 rem) to any other organs. The probability is

negligible that an individual would receive an external dose or internal dose commitment

in excess of 150 mSv (15 rem) to the whole body, the head and trunk, active bloodforming organs, the gonads, or the lens of the eye; 2 Sv (200 rem) to the hands and

forearms, feet and ankles, or localized areas of the skin averaged over areas no larger

than 1 cm2

; and 0.5 Sv (50 rem) to any other organs.

These criteria for failure of safety features have the effects of establishing an upper limit on the

amount of radioactive material in an exempt unit and of establishing a need to provide some

degree of containment for the material even under accident conditions. Furthermore, the

probabilities associated with radiation dose limits for accidental exposures ensure that the risk

to individuals would be very small.

In the last set of safety criteria for accidental exposures, it is intended that as the magnitude of

the potential dose increases above that permitted under normal conditions, the probability that

any individual will receive such a dose must decrease. The probabilities of failure of safety

features in products are expressed in general terms to emphasize the approximate nature of

the required estimates, but the following values may be used as guides in estimating

compliance with the safety criteria:

A probability of failure is low” if there is not more than one failure per year for each

10,000 exempt units distributed.

A probability of failure is negligible” if there is not more than one failure per year for

each 1 million exempt units distributed.

Therefore, the safety criteria for self-luminous products containing 3

H, 85Kr, or 147Pm used under

10 CFR 30.19 provide: (1) radiation dose limits for individuals from normal handling, storage,

use, and disposal of these products, and (2) radiation dose limits for individuals in conjunction

with associated probabilities of occurrence for accidents. The safety criteria do not include any

limits on collective dose.

2.14.2 Description of Items

A variety of self-luminous products that have existed at one time or another could potentially

cause exposure to consumers (NCRP 95, NUREG/CR–1775). These products include

wristwatches, telephone dials and push buttons, night map and document readers, leveling

bubbles, marine compasses, gun sights, and aircraft and building exit signs. Some of these

products were distributed for use under general license and some were distributed for use

under exemptions other than 10 CFR 30.19. The only items to be distributed for use in

significant numbers under this exemption have been wristwatches and gun sights containing

sealed tubes of 3

H gas.
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2.14.2.1 Wristwatches

Electronic digital watches started out as a gimmick consumer item in the early 1970s

(NUREG/CP–0001; Ristagno). Two competing technologies were the light-emitting diode

(LED), which offered a time readout on demand, and the liquid crystal display (LCD), which

offered a continuous time readout. The reason for the LED time-on-demand readout was the

high drain on the batteries used to power the watch. Because of the short battery life

expectancy of the LED, the LCD became the dominant technology. The battery life expectancy

was anywhere from 3 to 9 months for the LED and 18 to 24 months for the LCD. The LCD was

visible under normal ambient lighting conditions, but it became marginally visible at twilight and

invisible at night. Thus, backlighting with either miniature incandescent light bulbs or selfluminous gaseous tritium light sources (GTLSs) was used to make the LCD readable at night.

The self-luminous LCD wristwatches contained about 7.4 gigabecquerel (GBq) (200 millicurie

(mCi)) 3

H with a nominal composition, according to watch manufacturers, of elemental 3H 99%

and tritium oxide 1% in two sealed pyrex (borosilicate) glass tubes (NUREG/CR–0215). A

typical tube was approximately 2.0 cm long, with a 0.3-cm by 0.08-cm elliptical cross-section

and a 0.02-cm-thick wall. The tubes were coated on the inside with an inorganic phosphor. 

Light was produced when the phosphor was activated by beta particles emitted during the

decay of the 3

H in the sealed tube. The sealed tubes were well protected in an assembled

watch and were not easily accessible. Both tubes were bonded to a metal tray with a shockabsorbent silicone adhesive. The tray was sealed to a transflector-display panel, and the entire

assembly was encapsulated in a plastic collar. A printed circuit board was affixed to the plastic

collar, covering the bottom of the tray, and the entire module was then inserted into the watch

case as an integral unit. About 40,000 LCD watches backlighted by GTLSs were distributed

during 1976 (NUREG/CP–0001; Ristagno), and a total of several million were distributed during

the late 1970s and early 1980s (NUREG/CR–1775). Data reported by industry sources have

indicated that LCD watches backlighted by GTLSs were produced for only a few years during

the early 1980s (NCRP 95).

The self-luminous wristwatches now being distributed in the United States contain 14 to 15 tiny

GTLSs to light the dial and are imported mainly from Canada and Switzerland. In a typical

Swiss watch design (Yeaple, 1989), 14 GTLSs containing a total of about 1.9 GBq (50 mCi) of

3

H gas (HT) are used to provide good readability in any light. They are made of pyrex glass

and sealed with a laser process. Twelve of the GTLSs are 0.25 cm long with outer diameters of

0.095 cm, and the other two are 0.41 cm and 0.66 cm long with outer diameters of 0.07 cm. 

The two longer GTLSs are bonded to the minute and hour hands so they do not hard set using

elastomeric adhesives applied in a double layer, and the shorter GTLSs are placed at the

12-hour indexes on the dial. They are held in place by clamping pressure created when a

plastic holder ring having 12 U-shaped channels for the GTLSs is press fitted into the back of

the dial. The ring is then adhesively bonded to the dial. The wristwatch is shockproof and

waterproof to a depth of 30 meters. In a typical Canadian watch design, 15 tiny GTLSs

containing a total of about 1.1 GBq (30 mCi) of 3

H are used, with the extra GTLS being located

on the bezel of the watch. GTLSs containing 147Pm do not appear to be in current use.

2.14.2.2 Gun Sights

Self-luminous night sights are relatively new for firearms, having been introduced in the

mid-1980s (Kasler, 1992). On handguns, the total amount of 3

H contained in the three tiny
2–196

GTLSs of a typical three-dot set is approximately 1.9 GBq (50 mCi). The GTLSs glow with a

distinctive green color that quickly draws the eye to the proper sight alignment in low-light

situations. Night sights are currently standard issue with several Federal law enforcement

agencies, many State police departments, and hundreds of municipal and county departments. 

Statistically, about 80% of law enforcement shooting encounters take place in low-light or

no-light situations (Petty, 1992). Homeowners may also confront intruders during evening

hours, and anyone who may have to fire in low light can benefit from the use of night sights.

The GTLSs for some night sights are manufactured in the United States, but most are imported

from either Switzerland or Israel. In a typical Swiss sight assembly for a handgun, the GTLSs

consist of three pyrex tubes about 0.5 cm in length, 0.1 cm in outside diameter, and 0.02 cm in

wall thickness. Each tube is coated with an inorganic phosphor on the inside, filled with HT,

sealed with a laser, and leak tested. The next step in the assembly process is to shock mount

the GTLSs inside small aluminum tubes using silicone rubber cement. This provides shock

isolation to protect the GTLSs from weapon recoil. The viewing end of the aluminum tubes are

then sealed with polished industrial sapphires. Clarity of the sapphire provides maximum

brightness and its hardness (second only to a diamond) also provides protection for the GTLSs. 

Finally, the aluminum tubes are precision turned on a lathe to provide a controlled fit into holes

drilled in the steel sight assembly of the weapon, and they are sealed in place using a modified

cyanoacrylate adhesive (super glue). In a typical Israeli sight assembly for a handgun, the

GTLSs are entirely encased in a synthetic bonding material that completely seals the GTLSs

and makes them impervious to gun-cleaning solvents. Neither the Israeli nor the Swiss sights

are damaged by wiping and cleaning with conventional gun-cleaning solvents (Petty, 1992).

Self-luminous sights are also available as bow pin” sights for archery bows and as night

scopes for rifles and other large firearms. A bow pin sight typically uses a single GTLS

containing 0.19 to 0.37 GBq (5 to 10 mCi) of 3

H, and a night scope for a large firearm typically

uses one to three GTLSs containing 3.0 to 3.7 GBq (80 to 100 mCi) of 3

H per GTLS. The

distribution of bow pin sights and scopes for large firearms is small when compared to the

number of gun sights being distributed for handguns.

2.14.3 Summary of Previous Assessments

Assessments of radiological impacts on the public from self-luminous wristwatches containing

HT were performed by McDowell-Boyer and O’Donnell (NUREG/CR–0215) and Buckley et al.

(NUREG/CR–1775) and the results were summarized by the National Council on Radiation

Protection and Measurements (NCRP) (NCRP 95). The analysis of Buckley et al., was based

in large part on the work of McDowell-Boyer and O’Donnell. 

McDowell-Boyer and O’Donnell calculated doses from the escape of HT into air using either

CONDOS (Computer Codes, O’Donnell et al., 1975) or AIRDOS-II (Computer Codes, Moore,

1977). For calculational purposes individuals were divided into three groups: (1) wearers of the

wristwatches, (2) bystanders, and (3) distant persons. The CONDOS code was used in the

individual dose calculations for wearers of the watches and bystanders. Bystanders were

individuals who work near or associate with wearers of the watches. Three such individuals

were assumed to be near each wearer. For distant persons, the individual and collective doses

were calculated using the AIRDOS-II code. The 3

H leakage was considered to be a

ground-level release that was dispersed into the atmosphere, then absorbed through the skin
2–197

and inhaled by the distant persons within an 80-km radius of the watch wearer of the watch. An

annual distribution of 1 million watches and a 10-year effective lifetime for the watches were

used in the estimation of collective doses.

The exposure scenarios for routine use of self-luminous wristwatches that were considered

include:

exposures to workers and members of the public during distribution and transport;

exposures to wearers of watches, bystanders near wearers, and other members of the

public;

exposures to watch repairmen; and

exposures to members of the public from storage of discarded watches in the home and

disposal in landfills or by incineration.

The exposure scenarios considered involving accidents or misuse of self-luminous wristwatches

include:

exposure following breakage of a watch in the home;

exposure to a watch repairman following breakage in a repair shop; and

exposure following a fire in a warehouse.

In the dose analyses for the routine use scenarios, a leakage rate of 3

H from wristwatches

containing up to 7.4 GBq (0.2 Ci) of HT was assumed, on the basis of available data, to be 1.9

KBq/day (50 nCi/day) (i.e., 10 ppb/h), and all 3

H was assumed to be in the oxide form (i.e.,

tritiated water vapor (HTO)). The latter assumption probably resulted in overestimates of dose,

because some of the 3

H would be in elemental form (i.e., HT) and the dose per unit activity is

much less for HT than for HTO. 

In the dose analyses for storage in homes, the decay of 3

H over the previous 10 years was

considered and the leakage rate was reduced to 1.1 KBq/day (30 nCi/day) (i.e., 10 ppb/h). All

3

H was also assumed to be converted to HTO. 

In the dose assessments for accident and misuse scenarios, instantaneous release of 7.4 GBq

(0.2 Ci) of 3

H from watches was considered, and the 3H was assumed to be 99% in HT form

and 1% in HTO form. Doses resulting from accidents and misuse were also estimated by

assuming that all of the 3

H was HTO, but these results will not be discussed here because most

of the 3

H from an instantaneous release would be in the HT form.

The dose estimates obtained by McDowell-Boyer and O’Donnell (NUREG/CR–0215) for the

routine use scenarios are summarized as follows:

During distribution, truck drivers who pick up shipments could receive an annual dose

equivalent of 0.003 mSv (0.3 mrem), but other parcel system workers likely would

receive annual dose equivalents less than 1×104

 mSv (<0.01 mrem). Some marketing
2–198

employees could receive an annual dose equivalent of 0.0013 mSv (0.13 mrem), but

most workers likely would receive annual dose equivalents of less than 5×105 mSv

(<0.05 mrem). Annual dose equivalents to individual customers probably would be less

than 1×107

 mSv (<1×105 mrem). The collective dose equivalents from the annual

distribution of 1 million wristwatches was estimated to be about 0.1 person-Sv

(10 person-rem).

During routine use, wristwatch wearers could receive annual dose equivalents in the

range of 3×105

 to 2×104 mSv (0.003 to 0.02 mrem), depending on the location of the

wearer. The same doses could be received by bystanders who work near, live with, or

associate with the watch wearers. Other individual members of the public who are

distant from wearers could receive annual dose equivalents of 3×108

 mSv

(3×106

 mrem) or less. Annual collective dose equivalents from use of the wristwatches

were estimated to total 2.3 person-Sv (230 person-rem) (i.e., about 0.65 person-Sv

(65 person-rem) to wearers, 1.6 person-Sv (160 person-rem) to bystanders close to the

wearer, and 0.01 person-Sv (1 person-rem) to distant persons).

During repair, watch repairmen could receive annual dose equivalents in the range of

3×104

 to 6×104 mSv (0.03 to 0.06 mrem), depending on the size of the repair shop,

and the annual collective dose equivalent could be about 0.01 person-Sv

(1 person-rem).

During storage of discarded wristwatches in the home, individuals in the home could

receive annual dose equivalents as high as 3×105

 mSv (0.003 mrem), and distant

persons could receive annual dose equivalents 2×108

 mSv (2×106 mrem) or less. The

annual collective dose equivalent from storage could be about 0.5 person-Sv

(50 person-rem). If half of the watches are disposed of in landfills after storage and the

other half are burned, maximally exposed individuals could receive annual dose

equivalents of 0.001 mSv (0.1 mrem) near a landfill and 0.2 mSv (20 mrem) near a

location where watches were burned at ground level, no longer a common practice. 

Burning in an incinerator would not produce individual doses higher than those from

burning at ground level. The annual collective dose equivalents could be about

0.01 person-Sv (1 person-rem) from disposal in landfills and about 2 person-Sv

(200 person-rem) from incineration and burning at ground level.

The total annual collective dose equivalent from all normal use scenarios thus would be

5 person-Sv (500 person-rem). About half of the collective dose would be received

during use of wristwatches, and most of the rest results from incineration or burning at

ground level.

The total collective dose estimated by McDowell-Boyer and O'Donnell is about a factor of 15

less than the value of 77 person-Sv (7700 person-rem) adopted in NCRP 95 on the basis of

results given by Buckley et al. (NUREG/CR–1775). This difference results almost entirely from

the difference in estimated collective dose from disposal of wristwatches in landfills or by

incineration. For these scenarios, Buckley et al. used more conservative assumptions in the

dose analysis than those used by McDowell-Boyer and O’Donnell.

The dose estimates obtained by McDowell-Boyer and O’Donnell for the accident and misuse

scenarios are summarized as follows:
2–199

For breakage of a single wristwatch in a home, the dose equivalent to an individual over

the next 24 hours would be 0.005 mSv (0.5 mrem).

For breakage of a single watch in a small repair shop, the dose equivalent to a

repairman over the next 10 hours would be 0.1 mSv (10 mrem).

For a fire in a warehouse containing 60 wristwatches, and assuming no ventilation of air

for 15 minutes, the dose equivalent to a worker in the warehouse during that time would

be 0.01 mSv (1 mrem).

Thus, it is concluded from the published analyses that the individual and collective effective

dose equivalents (EDEs) to distant persons are extremely small and can be ignored in the case

of routine use and storage of old watches in homes. Also, it is concluded that (1) the individual

and collective doses for other exposure pathways should be recalculated for the current annual

distribution and 3

H activity in self-luminous wristwatches, (2) exposures from accidents,

disposal, and distribution of the watches should be reevaluated using the generic

methodologies in Appendixes A.1, A.2, and A.3 of this report, and (3) the 3

H intake through skin

in contact with the watch case should be considered in estimating individual and collective

doses to wearers during routine usage. The 3

H intake through skin in contact with a watch case

has been discussed in several recent articles by Brunner et al. (1996), Turvey (1996), Thüler

(1996), and Beyer et al. (1996).

For gun sights containing HT, there are no known previous analyses of the radiological impacts,

and a totally new assessment is needed for this self-luminous product.

2.14.4 Current Assessment for Wristwatches

Table 2.14.1 presents results of the current assessment of potential radiation doses for an

assumed annual distribution of 100,000 wristwatches initially containing 1.9 GBq (50 mCi) of 3

H

each. The effective lifetime of the watches is assumed to be 10 years, and the leakage rate of

3

H from the GTLSs in the watches is assumed to be 10 ppb/h under normal circumstances

(NUREG/CR–0215). For 3

H, the only important modes of exposure are inhalation and

absorption though the skin. However, the chemical form of the 3

H at the time of exposure is

very important with respect to the dose calculations. Doses from exposure to HT are estimated

to be about 10,000 times less than those from exposure to the same concentration of HTO in

air (see, for example, ICRP 68, Table C.1).

The 3

H in the GTLSs is generally more than 99% HT and less than 1% HTO. Thus, the 3H that

leaks from the GTLSs will mainly be in the HT form, but it will contact many surfaces that

catalyze oxidation before leaking into air (NUREG/CR–0215). For this reason, a conservative

approach is taken. For exposure to airborne levels, it is assumed that 3

H leaks from the GTLSs

as HTO under normal circumstances. This assumption may overestimate dose particulary for

persons in enclosed spaces (i.e., houses), but it should have less of an effect on exposures

from distant releases in the outdoors (i.e., from disposal). For skin uptake and exposure from

direct contact with surfaces containing GTLSs, special dosimetry has been developed (see

Section 2.14.4.2.1). Since accidents and misuse may involve an instantaneous release of the

3

H in the watches, no assumptions are made concerning elemental 3H conversion since contact
2–200

with other surfaces that catalyze oxidation would be less likely and room ventilation would

rapidly vent the HT (NUREG/CR–0215).

2.14.4.1 Distribution and Transport

The potential radiation doses from distribution and transport of the self-luminous watches are

estimated using the generic methodology of Appendix A.3. In applying this methodology, it is

assumed that local parcel-delivery drivers in large trucks pick up the watches from suppliers

and take them to local terminals, where they are shipped by semi-truck to other local terminals

for delivery to individual buyers or stores. Also, it is assumed that each shipment passes

through an average of four regional terminals before reaching its final destination.

It is assumed further that (1) the radiation dose to workers at both local terminals and regional

terminals are the same as those estimated for workers in a large warehouse, (2) a local parceldelivery driver could pick up an average of 200 watches per day (250 day/yr) from a single

supplier, (3) 50% of the watches are shipped directly to individual buyers and 50% are shipped

directly to stores, (4) retail clerks in some large department stores could be exposed

continuously to an average of 10 watches during the year, and (5) the leakage rate from the

watches is 10 ppb/h or 100 times less than the value of 1 ppm/h used in the development of the

generic methodology in Appendix A.3. The direct shipment of small, expensive items, such as

watches to retail stores, is commonly used to minimize possibility of theft during distribution.

Based on these assumptions and generic methodology in Appendix A.3, the individual EDE

could be 0.001 mSv (0.1 mrem) for a local parcel-delivery driver who is assumed to pick up

1000 self-luminous watches from the same supplier each week (50 wk/yr). Individual doses to

other truck drivers, terminal workers, retail clerks, store customers, and members of the public

along truck routes would be less. The total collective EDE from 1 year’s distribution of 100,000

self-luminous watches containing 1.9 GBq (50 mCi) of 3

H per watch could be 1×102 person-Sv

(1 person-rem), due almost entirely to exposure to store clerks and shoppers.

2.14.4.2 Routine Use

Two modes of exposure can occur during routine use of self-luminous wristwatches containing

3

H. One mode is exposure to airborne releases of 3H from the wristwatches, and the other

mode is exposure from skin contact with the case of the wristwatch. The latter mode of

exposure applies only to the wristwatch wearer, whereas the first mode of exposure applies to

the wristwatch wearer as well as coworkers and other family members.

The following parts of this section discuss individual doses to wearers from skin contact with the

wristwatches (Section 2.14.4.2.1); individual doses to wearers, coworkers, and other family

members due to airborne releases during routine use (Section 2.14.4.2.2); individual doses due

to storage of old watches in homes (Section 2.14.4.2.3); and total individual and collective

doses from all types of exposures during routine use and storage of watches in homes

(Section 2.14.4.2.4).

2.14.4.2.1 Skin Contact With the Wristwatch Case

The importance of this pathway of exposure is noted from a publication by Brunner et. al

(1996). In the study by Brunner et al., release rates from 82 different plastic case watches were
2–201

determined. The observed release rates ranged from 110 to 160,000 Bq/day (3 to

4300 nCi/day) with a median of 16,000 Bq/day (430 nCi/day); mean of 24,400 Bq/day

(660 nCi/day). The reported cumulative frequency distribution of the release rates was

consistent with a log normal distribution and, therefore, the central value is given by the median. 

The watches contained 200 to 300 MBq (5.4 to 8.1 mCi) and the fractional release rate is about

3×106

 per hour, or 3 ppm/h. The concentration of 3H in the urine of 108 individuals wearing

such watches also was measured. Observed concentrations ranged from minimum detectable

activity (MDA) to about 1100 Bq/L (30 nCi/L) with a median of 99 Bq/L (2.7 nCi/L); mean of

197 Bq/L (5.3 nCi/L). Exposure time of subjects ranged from 2 to 24 h/day. Comparing the

cumulative frequency distribution of the data for tritium release rate from the watches with the

measured concentration in urine provides a means for estimating the fractional uptake of tritium

by individuals wearing watches. If it is assumed that 50% of the tritium uptake is excreted in the

urine (ICRP 23), then the fractional uptake can be calculated, based on a standard excretion

rate of 1.4 liters per day, by assuming equilibrium conditions between uptake and excretion. 

Based on the data for the median, 75%, and maximum values for watch leak rate and for

concentration in urine as reported by Brunner et al. (1996), the corresponding uptake can be

calculated to be approximately 2%. While the chemical form and actual amounts of tritium

uptake via skin absorption versus inhalation are unknown, this 2% absorption is not inconsistent

with that measured by Eakins et. al., (1974) for exposure to skin to metal foils containing

absorbed tritium.

The dosimetry for tritium uptake from watches is not well established. Johnson and Dunford

(1984) modeled the uptake, transfer, and elimination of tritium via skin absorption based on the

data from study conducted by Eakins et. al. Their modeling yielded dose factors for the skin

dose equivalent per becquerel uptake of 4.5×108

 Sv (1.7×105 rem/Ci) and 4.9×109 Sv

(1.8×104

 rem/Ci) over a 40 cm2 skin area for maximum and minimum skin retention times,

respectively. As acknowledged in their study, much uncertainty surrounds the modeling and

assumptions for the dosimetry for skin uptake of tritium, and at best these dose factors provide

a crude estimate for approximating potential dose significance. For purposes of this study, the

decision was made to use the maximum skin retention dose factor of 1.8×103

 mSv-cm2/Bq

(6.7×103

 mrem-cm2/Ci), derived from the Johnson and Dunford study by normalizing the

uptake for the exposed skin area. This factor provides what is believed to be an upper bound

assessment, based on the limited data. It should be recognized that a more probable dose

could be an order of magnitude (or more) lower.

To estimate the potential radiation doses due to skin absorption of 3

H from a self-luminous

wristwatch initially containing 1.9 GBq (50 mCi) of 3

H, the analysis proceeds as follows:

First, the 3

H leakage from the self-luminous watch is (1.9 GBq (50 mCi)) × (10 ppb/h) ×

(16 h/day) or 300 Bq/day (0.008 Ci/day), where 10 ppb/h is the assumed rate of HTO

leakage from the self-luminous watch (see Section 2.14.4). 

Second, the intake of HTO through the skin in contact with the case of the watch is

(300 Bq/d (0.008 Ci/d)) × (0.02), or 6 Bq/day (1.6×104 Ci/day), where 0.02 is the

fractional absorption of 3

H released from the watch by the skin in contact with the case

of the watch (see above). 

Third, the annual dose equivalent from the HTO to the skin in contact with the case is

(6 Bq/d (1.6×104 Ci/day)) × (365 d/yr) × (1.8×103

 mSv-cm2/Bq
2–202

(6.7×103

 mrem-cm2/Ci)) ÷ (10 cm2), or about 0.4 mSv (40 mrem) over an area of

10 cm2

. The area of skin in contact with the wristwatch is assumed to be about 10 cm2

(Turvey, 1996).

The average annual dose equivalent to the skin of the whole body from the distributed

wristwatch source is calculated by multiplying the localized skin dose by the ratio of the

exposed area of 10 cm2

 to the total skin area of the body of 1.8×104 cm2 (ICRP 23) ((0.4 mSv

(40 mrem)) × (10 cm2

/1.8×104 cm2), or about 2×104 mSv (0.02 mrem). The contribution of this

skin dose equivalent to the annual EDE is (2×104

 mSv (0.02 mrem) × (0.01), or less than

1×105

 mSv (<0.001 mrem), where 0.01 is the organ weighting factor for skin of the whole body

(ICRP 60). 

The annual EDE to the internal organs of the body for the absorption of HTO through the skin in

contact with the case of the watch is (6 Bq/day (1.6×104 Ci/day)) × (365 day/yr) ×

(1.7×108

 mSv/Bq (6.4×102 mrem/Ci)), or 4×105 mSv (0.004 mrem), where 1.7×108 mSv/Bq

(6.4×102

 mrem/Ci) is the dose conversion factor for either absorption through the skin or

ingestion of 3

H (see Section 2.1). 

In summary, the annual dose equivalent to a small area of skin is estimated to be 0.4 mSv

(40 mrem) over an area of 10 cm2

 in contact with the wristwatch. The skin dose due to the

distributed wristwatch source of 3

H makes a negligible contribution to the annual EDE, and the

total annual EDE to a wearer from skin absorption of 3

H in contact with the case of a

self-luminous watch is estimated to be 4×105

 mSv (0.004 mrem).

The collective EDE of 1 year’s distribution of 100,000 wristwatches initially containing 1.9 GBq

(50 mCi) of 3

H each could be 0.004 person-Sv (0.4 person-rem) during the first year of routine

use. The total collective dose is estimated to be about 0.03 person-Sv (3 person-rem) over the

10-year useful life.

2.14.4.2.2 Airborne Releases from Wristwatch

Because watches may be worn in a variety of ways during routine use, the following two

scenarios have been chosen to indicate the potential dose from airborne releases of 3

H from

self-luminous wristwatches.

Scenario I. A wearer spends 12 hours at home each day (4380 h/yr) and exposes three other

family members to airborne releases of 3

H from the wristwatch. The home has an enclosed

volume of 450 m3

 and a ventilation rate of 1 volume change per hour. The equilibrium

concentration of HTO in the air of the home is approximately 37 MBq/m3

 (1 pCi/m3) and the

breathing rate of the individuals is 0.9 m3

/h. Thus, the annual EDE to the wearer and to other

family members would be less than 1×105

 mSv (<0.001 mrem), assuming the other family

members are exposed over the same 12 h/day as the wearer.

Scenario II. A wearer works 8 h/day (2000 h/yr) in an office or shop and exposes two

coworkers to airborne release of 3

H from a wristwatch. The office or shop has an enclosed

volume of 34 m3

 and a ventilation rate of 1 volume change per hour. The equilibrium

concentration of HTO in the air of the office or shop is approximately 0.55 Bq/m3

 (15 pCi/m3)

and the breathing rate of the individuals is 1.2 m3

/h. Thus, the annual EDE to the wearer and
2–203

two coworkers would be less than 1×105

 mSv (<0.001 mrem), assuming the coworkers are

exposed over the same 8 h/day as the wearer.

To estimate the collective dose from airborne releases of 3

H during routine use of the

self-luminous watches, it is assumed that 50% of the wearers work outdoors or in areas with

high ventilation rates so that their annual EDEs at work are essentially zero, and the other 50%

work in small offices or shops with low ventilation rates so that their annual individual EDEs at

work are similar to those in Scenario II. While the estimated individual doses are presented as

less than values, the calculated values used for estimating collective dose are 4×106

 mSv

(4×104

 mrem) for Scenario I and 5×105 mSv (0.005 mrem) for Scenario II. Thus, the

collective EDEs from Scenarios I and II would total about 0.009 person-Sv (0.9 person-rem) for

the first year and 0.07 person-Sv (7 person-rem) over 10 year useful life.

2.14.4.2.3 Total Individual and Collective Doses

For a 16-hour-per-day wearer of a self-luminous wristwatch containing 1.9 GBq (50 mCi), the

annual dose equivalent to skin from routine use could be 0.4 mSv (40 mrem) over an area of

10 cm2

 in contact with the wristwatch. The annual individual EDE to such a wearer from routine

use could be approximately 4×105

 mSv (0.004 mrem), essentially all due to absorption of 3H

through the skin in contact with the case of the watch (see Section 2.14.4.2.1). The individual

dose to coworkers and other family members is significantly less.

The total collective dose equivalent to both wearers and other members of the public from

routine use of 1 year’s distribution of 100,000 watches containing 1.9 GBq (50 mCi) of 3

H could

be 0.1 person-Sv (10 person-rem). Of this total, 0.03 person-Sv (3 person-rem) is due to

exposure to the wearer by absorption of 3

H through the skin in contact with the case of the

watch (see Section 2.14.4.2.1), and 7×102

 person-Sv (7 person-rem) is due to airborne

releases from the watch while at work or at home (see Section 2.14.4.2.2).

2.14.4.3 Watch Repair

Watch repairmen may adjust or replace batteries in the watches. Other repair operations are

not considered because work involving batteries is likely to account for the majority of repairs. 

It is unlikely that 3

H will be released catastrophically from the watches during repair, but normal

leakage of 3

H from the watch will continue. The unlikely breakage of GTLSs in a watch shop is

considered in Section 2.14.4.5.

For a repairman at a watch repair shop, the potential doses were estimated using the following

scenarios: 

the shop had an enclosed volume of 34 m3

 and a ventilation rate of 1 volume change

per hour, and the watch repairman was exposed to airborne leakage of 3

H from a watch

for 1 day (8 hours) between the time the watch was received at the shop and returned to

the owner;

the average time for adjustment and battery replacement was 10 minutes, and the

repairman was exposed during this time to airborne releases of 3

H into a small

hemispherical air space with a radius of 1.5 meters, a volume of 7 m3

, and a ventilation

rate of 1 volume change per hour; and 
2–204

each watch was handled for 10 minutes during watch repair, and 2% of the 3H escaping

from the watch was absorbed through a skin area of 3 cm2

 on the ends of the

repairman’s fingers. 

Based on these assumptions, a breathing rate of 1.2 m3

/h for light activity, and the repair of 100

watches per year, the EDE to the watch repairman from all exposure pathways could be 

1×105

 mSv (0.001 mrem), and the dose equivalent to the skin could be 0.004 mSv (0.4 mrem)

over an area of 3 cm2

 in contact with the watches.

If the watches are serviced every 2 years mainly to have batteries replaced, the collective EDE

to repairmen could be 7×106

 person-Sv (7×104 person-rem) for servicing one half the watches

on an annual basis and 5×105

 person-Sv (0.005 person-rem) over the 10-year useful life of the

100,000 watches distributed annually.

2.14.4.4 Disposal of Watches

Half of the watches are assumed to be disposed of as ordinary trash after 10 years of use, and

the other half are assumed to be used for 10 years, then stored in homes for an additional

5 years before disposal. The total amount of 3

H in 1 year’s distribution of 100,000 watches is

estimated to be about 92 TBq (2.5 kCi) at the time of disposal.

To estimate potential individual and collective doses to the public from disposal of self-luminous

wristwatches, the generic disposal methodology in Appendix A.2 was used. It has been

assumed in applying this methodology that most of the GTLSs remain intact during waste

collection and landfill disposal. Thus, a reduction factor of 10 was applied to the following doseto-source ratios for inhalation and ingestion in Appendix A.2: (1) waste collectors at both

landfills and incineration, (2) workers at landfills, (3) off-site members of the public exposed to

airborne releases during landfill operations and releases to groundwater following disposal in

landfills, and (4) future on-site residents at landfills.

For disposal at landfills, the annual individual EDE would be about less than 1×105

 mSv

(<0.001 mrem) to waste collectors. The annual individual doses to workers at landfills, off-site

members of the public, and future on-site residents would be less. The total collective EDE was

found to be about 2×103

 person-Sv (0.2 person-rem), due almost entirely to exposure to offsite members of the public from groundwater releases.

For disposal by incineration, the annual EDE would be 4×105

 mSv (0.004 mrem) to waste

collectors. The annual individual dose is less to workers at incinerators and off-site members of

the public. The total collective EDE is about 3×103

 person-Sv (0.3 person-rem), due mainly to

exposures to off-site members of the public from airborne releases during incinerator

operations.

The above dose estimates are for exposure to multiple exempt units during waste collection

and disposal. For exposure to a single exempt unit used for 10 years before disposal, the

individual EDE for waste collectors at either landfills or incinerators would be less than

1×105

 mSv (<0.001 mrem). The individual doses to other workers and members of the public

from exposure during disposal of a single exempt unit would be less.
8

 The dose conversion factors for effective dose equivalent due to ingestion of 3H or absorption

of 3

H through the skin are the same numerically.

2–205

2.14.4.5 Accidents and Misuse

In the case of accidents, the following was considered: (1) a catastrophic release from crushing

of a single watch in a repair shop, (2) an accident involving the crushing of a single watch in a

home, and (3) a shipping accident in a storeroom or cargo-handling area involving the crushing

of a shipment of 200 watches. The 3

H contained in the watches is assumed to be 99% HT and

1% HTO. Based on these assumptions and the generic accident methodology in Appendix A.1,

the potential radiation doses from the crushing of self-luminous watches containing 3

H can be

summarized as follows:

For a watch repairman, the individual EDE from crushing of a single watch containing

1.9 GBq (50 mCi) of 3

H could be 0.03 mSv (3 mrem) at a small repair shop or 0.02 mSv

(2 mrem) at a large repair shop.

For a person at home, the individual EDE from crushing of a single watch containing

1.9 GBq (50 mCi) of 3

H could be 0.001 mSv (0.1 mrem).

For a worker in a storeroom or cargo-handling area, the individual EDE from crushing of

200 watches containing a total of 370 GBq (10 Ci) of 3

H could be 0.1 mSv (10 mrem).

In the case of misuse, this analysis considers the exposure to a 5-year-old child who plays with

a self-luminous watch as a glow-in-the-dark” toy at night while going to sleep during one year. 

It is assumed that (1) the watch is a 10-year-old watch containing 1.1 GBq (30 mCi) of 3

H, (2)

the child handles the watch 10 min/day, (3) the child absorbs 2% of the 3

H released from the

watch through a skin area of 10 cm2

 while handling the watch, and (4) the child sleeps in a

closed bedroom with the watch 12 h/day. It is further assumed that (1) the bedroom has an

enclosed volume of 27 m3

 and a ventilation rate of 1 air change per hour (see Appendix A.1),

(2) the child’s breathing rate is 0.24 m3

/h while sleeping (ICRP 66), (3) the dose conversion

factors for inhalation and ingestion8

 are about twice those for an adult (ICRP 67; ICRP 71), and

(4) the total surface area of the child's skin is approximately 0.8 m2

 (ICRP 23). Based on these

assumptions, the potential radiation doses to the 5-year-old child can be summarized as

follows:

The dose equivalent to the skin of the 5-year-old child due to absorption of 3

H from the

watch could be 0.002 mSv (0.2 mrem) over a skin area of 10 cm2

 in contact with the

watch.

The EDEs would be less than 1×105

 mSv (<0.001 mrem) due to absorption of 3H

through the skin in contact with the watch and 2×105

 mSv (0.002 mrem), due to

airborne releases of 3

H from the watch.

2.14.4.6 Safety Criteria

Table 2.14.2 compares results of the current dose assessment for self-luminous wristwatches

with safety criteria for self-luminous products set forth in 10 CFR 32.23 (see Section 2.14.1). 
2–206

First, the limiting 10 CFR 32.23 values for dose to the whole body are compared with the

current assessment’s maximum estimates of the annual individual EDE due to 3H releases from

the self-luminous watches for the current assessment. Second, the limiting 10 CFR 32.23

values for dose to skin are compared with the current assessment’s maximum estimates of the

annual dose equivalent to skin of individuals exposed by 3

H absorption through the skin in

contact with the watches. It is important to note that the skin dose from normal use of a single

self-luminous wristwatch could exceed the limiting skin dose set forth in 10 CFR 32.23. This

result contains some rather large uncertainties for the following reasons: (1) the rate of leakage

of 3

H from the GTLSs in the self-luminous watches is not well known, and (2) the conversion

factor for relating skin dose to absorption of 3

H through the skin in contact with the wristwatch is

not well established (see Section 2.14.4.2.1). However, the currently estimated annual dose

equivalent of 0.4 mSv (40 mrem) for skin of a wearer in contact with the self-luminous

wristwatch is much less than the annual dose limit of 50 mSv (5 rem) for skin of members of the

public that is recommended in both ICRP 60 and NCRP 116.

2.14.5 Current Assessment for Gun Sights

Table 2.14.3 provides the results of our assessment of potential radiation doses from an

assumed annual distribution of 100,000 handgun sights initially containing 1.9 GBq (50 mCi) of

3

H each. It should be noted that the average activity of the handgun sights is significantly less

than the maximum activity of 11 GBq (300 mCi) used under this exemption for rifle scopes, and

the number of rifle scopes distributed annually is quite small in comparison to the number of

handgun sights. Hence, individual doses resulting from exposures to a single exempt item

during routine use and during accidents are assumed to be controlled by the maximum amount

of activity in a rifle scope. All other individual doses resulting from exposure to multiple exempt

items and, thereby, all collective doses are assumed to be controlled by the average activity per

handgun sight and by the number distributed annually. However, individual doses based on

both rifle scopes and handgun sights for routine exposures are provided here, because the

latter were developed for use in the estimation of collective dose. It is further assumed that the

useful lifetime of the handgun sights and rifle scopes is 10 years, and the leakage rate of 3

H is

10 ppb/h.

2.14.5.1 Distribution and Transport

The potential radiation doses from distribution and transport of the gun sights are estimated

using the generic methodology of Appendix A.3. In applying this methodology, it is assumed

that local parcel-delivery drivers in large trucks pick up the self-luminous gun sights from

suppliers and take them to a local terminal, where they are shipped by semi-truck to other local

terminals for delivery to gunsmiths or manufacturers. Furthermore, it is assumed that each

shipment passes through an average of four regional terminals before reaching its final

destination. It is also assumed that the radiation dose to workers at both local terminals and

regional terminals are the same as those estimated for workers in a large warehouse and the

leakage rate from the gun sights is 10 ppb/h, or 100 times less than the value of 1 ppm/h used

in the development of the generic methodology in Appendix A.3.

Most of the handgun sights are assumed to be shipped in lots of 1000 or less to gun

manufacturers, who mount the sights as an option for buyers. These handguns are then picked

up by local drivers and returned to local terminals, where they are shipped to other local
2–207

terminals for delivery to the gun buyers. It is assumed further that essentially all of the gun

sights are mounted by installers in factories and the above scenario applies during shipment of

these mounted sights to gun buyers. Based on these assumptions and generic methodology in

Appendix A.3, the individual EDE could be 0.001 mSv (0.1 mrem) for a local parcel-delivery

driver who is assumed to pick up 1000 unmounted gun sights from the same supplier each

week (50 wk/yr). Individual doses to other truck drivers, terminal workers, and members of the

public along truck routes would be less. The total collective EDE from distribution of a total of

100,000 unmounted and mounted gun sights could be 3×105

 person-Sv (0.003 person-rem).

2.14.5.2 Gun Sight Installers

In a gun repair shop, the time taken by a gunsmith to mount and align a handgun sight is

estimated to be about 30 minutes, and in a factory, about 10 to 15 minutes. A factory installer

could mount and align about 5000 sights per year (100 sights per week for 50 wk/yr). A total of

about 20 factory installers would be needed to mount and align 1 year’s distribution of 100,000

gun sights if it is assumed that essentially all gun sights are installed at factories. It is unlikely

that 3

H will be released catastrophically from the gun sights during installation, but normal

leakage of 3

H will continue. The unlikely breakage of GTLSs in a self-luminous sight at a gun

repair shop is considered in Section 2.14.5.5.

For a gun installer at a factory, the potential doses were estimated using the follow scenarios:

(1) the installer was exposed for 40 h/wk (2000 h/yr) to airborne releases from 1 week’s supply

of 100 gun sights into a hemispherical air space with a radius of 1.5 meters, a volume of 7 m3

,

and a ventilation rate of 5 volume changes per hour, and (2) each sight was handled for 15

minutes while being installed on a handgun and 2% of the 3

H escaping from the sight was

absorbed through a skin area of about 3 cm2

 on the ends of the installer’s fingers. Based on

these assumptions, a breathing rate of 1.2 m3

/h for light activity, and the installation of 5000

gun sights per year, the annual EDE to a single gun installer could be 0.003 mSv (0.3 mrem),

essentially all from inhalation intake. The annual dose equivalent to the skin could be 0.3 mSv

(30 mrem) over a skin area of 3 cm2

 in contact with the gun sights. The collective EDE to all

gun installers from 1 year’s distribution of 100,000 gun sights is estimated to be approximately

7×105

 person-Sv (0.007 person-rem).

2.14.5.3 Routine Use

Because a handgun with self-luminous sights may be used in a variety of ways by law

enforcement agents and private gun owners, the following two scenarios have been chosen to

indicate potential doses from routine use.

Scenario I. A police officer spends 6 h/day (1500 h/yr) in a cruiser with another police officer

(exposure to sights on two guns) and 12 h/day (4380 h/yr) at home (exposure to sights on one

gun). The cruiser is assumed to have a volume of 6.2 m3

 and a ventilation rate of 5 volume

changes per hour, and the residence is assumed to have a volume of 450 m3

 and a ventilation

rate of 1 volume change per hour (see Appendix A.1). The breathing rates are assumed to be

0.9 and 1.2 m3

/h while at home and at work, respectively.

The equilibrium concentrations of 3

H in the cruiser and in the home are approximately

1.11 Bq/m3

 (30 pCi/m3) and 0.037 Bq/m3 (1 pCi/m3), respectively. For the police officer, the

annual EDE is estimated to be 6×105

 mSv (0.006 mrem) in the first year. For an assumed
2–208

average of three other family members, the initial annual EDEs are estimated to be less than

1×105 mSv (<0.001 mrem), assuming the other family members are exposed at home over the

same 12-hour period as the police officer.

Scenario II. A private gun owner stores the handgun at home for protection. The equilibrium

concentration is 0.037 Bq/m3

 (1 pCi/m3), under the above assumptions, and the annual EDE is

less than 1×105

 mSv (<0.001 mrem) for the gun owner and three other family members,

assuming they each spend 12 h/day at home.

To estimate the collective EDE from the routine use of the 100,000 handgun sights used for

10 years, it was assumed that 20% of the sights were purchased by private gun owners and

80% were purchased for police officers (or other law enforcement agents). The collective EDE

would total about 4×103

 person-Sv (4 person-rem), essentially all attributable to police officers,

for 1 year’s distribution of 100,000 handgun sights initially containing 1.9 GBq (50 mCi) of 3

H

each.

In the case of a private gun owner who stores a rifle with a self-luminous scope containing

11 GBq (300 mCi) of 3

H at home, the annual individual EDE could be 3×105 mSv (0.003 mrem)

the case of a police officer who patrols alone and carries such a rifle in a police cruiser, the

annual individual EDE could be 3×104

 mSv (0.03 mrem). These estimates are based on the

same assumptions as above, except the rifle used by the police officer is assumed to be stored

at a police station when not in use, rather than at home.

2.14.5.4 Disposal

To estimate the potential radiation doses due to the disposal of self-luminous gun sights

containing 3

H, the generic disposal methodology of Appendix A.2 has been used. It is assumed

in applying this methodology that most of the GTLSs remain intact during waste collection and

landfill disposal. Thus, a reduction by a factor of 10 is applied to the following dose-to-source

ratios for inhalation and ingestion in Appendix A.2: (1) waste collectors at both landfills and

incinerators, (2) workers at landfills, (3) off-site members of the public exposed to airborne

releases during landfill operation and releases to groundwater following disposal in landfills, and

(4) future on-site residents. The total activity of 3

H from 1 year’s distribution of 100,000 gun

sights after 10 years of use is approximately 110 TBq (3,000 Ci).

For disposal at landfills, the annual individual EDE would be less than 1×105

 mSv

(<0.001 mrem) to waste collectors. The annual individual doses to workers at landfills, off-site

members of the public, and future on-site residents would be less. The total collective EDE was

found to be about 3×103

 person-Sv (0.3 person-rem), due almost entirely to exposures to offsite members of the public from groundwater releases.

For disposal by incineration, the annual EDE would be 4×105

 mSv (0.004 mrem) to waste

collectors. The annual individual dose to workers at incinerators and off-site members of the

public are less. The total collective EDE is about 4×103

 person-Sv (0.4 person-rem), due

mainly to exposures to off-site members of the public from airborne releases during incinerator

operations.

The above dose estimates are for exposure to multiple exempt units during waste collection

and disposal. For a gun sight containing 1.9 GBq (50 mCi) of 3

H initially, the activity is about
9

 The dose conversion factors for EDE due to ingestion of 3H or absorption of 3H through the

skin are the same numerically.

2–209

1.1 GBq (30 mCi) at time of disposal, and the individual EDE to a waste collector would be less

than 1×105

 mSv (<0.001 mrem). For a rifle scope containing 11 GBq (300 mCi) of 3H initially,

the activity is about 6.7 GBq (180 mCi) at time of disposal, and the individual EDE to a waste

collector would also be less than 1×105

 mSv (<0.001 mrem).

2.14.5.5 Accidents and Misuse

In the case of accidents, the following scenarios have been considered: (1) a catastrophic

release from crushing of a rifle scope in a repair shop, (2) an accident involving the crushing of

a rifle scope in a home, and (3) a shipping accident in a storeroom or cargo-handling area

involving the crushing of a shipment of 1000 handgun sights. The 3

H contained in the rifle

scopes and handgun sights is assumed to be 99% HT and 1% HTO. Based on these

assumptions and the generic accident methodology in Appendix A.1, the potential radiation

doses from the crushing of self-luminous watches containing 3

H can be summarized as follows:

For a gunsmith at a repair shop (see data for watch repair shops in Tables A.1.2 and

A.1.9 of Appendix A.1), the individual EDE from crushing of a rifle scope containing

11 GBq (300 mCi) of 3

H could be 0.2 mSv (20 mrem) at a small shop or 0.1 mSv

(10 mrem) at a large shop.

For a person at home (see Tables A.1.2. and A.1.9 of Appendix A.1), the individual EDE

from crushing of a rifle scope containing 11 GBq (300 mCi) of 3

H could be 8×103 mSv

(0.8 mrem).

For a worker in a storeroom or cargo-handling area (see Tables A.1.2 and A.1.9 of

Appendix A.1), the individual EDE from crushing of 1000 handgun sights containing a

total of 1.9 TBq (50 Ci) of 3

H could be 0.5 mSv (50 mrem).

In the case of misuse, the analysis considers the exposure to a 5-year-old child who plays with

a self-luminous sight from a handgun as a glow-in-the-dark” toy at night while going to sleep

during one year. It is assumed that the (1) handgun sight is a 10-year-old sight containing 1.1

GBq (30 mCi) of 3

H, (2) the child handles the sight 10 min/day, (3) the child absorbs 2% of the

3

H released from the sight through a skin area of 3 cm2 while handling the sight, and (4) the

child sleeps in a closed bedroom with the sight 12 h/day. It is further assumed that (1) the

bedroom has an enclosed volume of 27 m3

 and a ventilation rate of 1 air change per hour (see

Appendix A.1), (2) the child’s breathing rate is 0.24 m3

/h while sleeping (ICRP 66), (3) the dose

conversion factors for inhalation and ingestion9

 are about twice those for an adult (ICRP 67;

ICRP 71), and (4) the total surface area of the child's skin is approximately 0.8 m2

 (ICRP 23). 

Based on these assumptions, the potential radiation doses to the 5-year-old child can be

summarized as follows:

The dose equivalent to the skin of the 5-year-old child due to absorption of 3

H from the

gun sight could be 0.008 mSv (0.8 mrem) when averaged over 3 cm2

 of skin area in

contact with the sight.
2–210

The EDE would be less than 1×105

 mSv (<0.001 mrem) due to absorption of 3H

through the skin in contact with the sight and 2×105

 mSv (0.002 mrem), due to airborne

releases of 3

H from the sight.

2.14.5.6 Safety Criteria

Table 2.14.4 compares the results of the current dose assessment for self-luminous gun sights

with the safety criteria for self-luminous products set forth in 10 CFR 32.23 (see Section

2.14.1). First, the limiting 10 CFR 32.23 values for dose to the whole body are compared with

the current assessment’s maximum estimates of the annual individual EDE due to 3

H releases

from the self-luminous gun sights. Second, the limiting 10 CFR 32.23 values for dose to skin

are compared with the current assessment’s maximum estimates of the annual dose equivalent

to skin of individuals exposed by 3

H absorption through the skin in contact with the sights. The

individual doses to the whole body (EDE) and the dose equivalents to skin do not exceed the

dose limits set forth in 10 CFR 32.23 (see Table 2.14.6).

2.14.6 Summary 

Tables 2.14.1 and 2.14.3 present results of the current assessments of potential radiation

doses to the public from self-luminous products containing 3

H. These results are based on an

annual distribution of 100,000 self-luminous wristwatches and 100,000 self-luminous

gun sights. The self-luminous wristwatches are assumed to contain 1.9 GBq (50 mCi) of 3

H per

watch, and the self-luminous gun sights are assumed to contain from 1.9 to 11 GBq (50 to

300 mCi) of 3

H per sight. The watches and sights are both assumed to have an effective

lifetime of 10 years.

The estimate of the total collective EDE to the public from self-luminous products is

0.2 person-Sv (20 person-rem) (see Tables 2.14.1 and 2.14.3). This estimate is smaller than

the previous estimate of 77 person-Sv (7700 person-rem) in NCRP 95. Part of the difference is

due to design modifications of the wristwatches and assumptions regarding the annual

distribution of these consumer products (see Section 2.14.2.1). However, the difference is

largely due to the conservative assumptions used in making dose estimates for landfill burial

and incineration of self-luminous products in the previous study by Buckley et al.

(NUREG/CR–1775) (see, also, Section 2.14.3).

Tables 2.14.2 and 2.14.4 also present comparisons of the current estimates of annual radiation

doses to individuals with the safety criteria in 10 CFR 32.23 (see Section 2.14.1). It is important

to note that the annual dose equivalent to skin from normal use of a self-luminous watch could

exceed the skin dose limit set forth in 10 CFR 32.23 (see Table 2.14.3). This result contains

some rather large uncertainties for the following reasons: (1) the rate of 3

H leakage from the

GTLSs in the self-luminous wristwatches is not well known, and (2) the conversion factor for

relating skin dose to absorption of 3

H through the skin in contact with the wristwatch is not well

established (see Section 2.14.4.2.1). Therefore, studies are needed to obtain better data for

both of these important dose-related parameters.
2–211

Table 2.14.1 Potential Radiation Doses From Self-Luminous Wristwatches

Containing 3

H

Exposure Pathway

Individual Annual Effective

Dose Equivalent

(mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport 0.1c 1.0

Routine use 0.004d 10

Watch repair 0.001e 0.005

Disposal

 Landfills

 Incinerators

<0.001f

0.004g

0.2

0.3

Accidents and misuse

 Toy for a child

 Crushing of watches

0.003h

10i

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Collective doses are based on an assumed annual distribution of 100,000 wristwatches having

an effective lifetime of 10 years and initially containing 1.9 GBq (50 mCi) of gaseous 3

H per

watch. Refer to text for time period of calculations.

c

 Dose estimate applies to a local parcel-delivery driver exposed to multiple exempt units during

distribution and transport; dose estimates are less for other truck drivers, workers at truck

terminals and distribution warehouses, store clerks and shoppers, and members of public along

truck routes (see Section 2.14.4.1).

d

 Dose estimate applies to 16-hour-per-day wearers of wristwatches; dose estimate for annual

dose equivalent to skin of such wearers is 0.4 mSv (40 mrem) over an area of 10 cm2

 in contact

with the case of the wristwatch (see Section 2.14.4.2).

e

 Dose estimate applies to watch repairmen exposed to multiple exempt units; dose estimate for

annual dose equivalent to skin is 0.004 mSv (0.4 mrem) over a skin area of 3 cm2

 in contact

with watches (see Section 2.14.4.3). 

f

 Dose estimate applies to waste collectors exposed to multiple exempt units during disposal at

landfills; dose estimates are less for workers at landfills, off-site members of public, and future

on-site residents (see Section 2.14.4.4).

g

 Dose estimate applies to waste collectors exposed to multiple exempt units during disposal at

incinerators; dose estimates are less for workers at incinerators and off-site members of public

(see Section 2.14.4.4).

h

 Dose estimate applies to a 5-year-old child who is allowed to use an old self-luminous watch

as a toy during a year; dose estimate for dose equivalent to skin is less than 1×105

 mSv

(<0.001 mrem) due to absorption of 3

H through the skin and 2×105 mSv (0.002 mrem) due to

airborne releases of 3

H from the watch (see Section 2.14.4.5).

i

 Dose estimate applies to a worker exposed during a shipping accident in a storeroom or

cargo-handling bay involving the crushing of 200 self-luminous watches; dose estimate for a

watch repairman following the crushing of a single exempt unit in a small watch repair shop is

0.03 mSv (3 mrem) (see Section 2.14.4.5).
2–212

Table 2.14.2 Comparison of Estimated Maximum Individual Doses From Self-Luminous

Wristwatches Containing 3

H and Limiting Organ Doses From 10 CFR 32.23

Exposure Conditionsa Maximum Individual

Dose (mrem/yr or mrem)b

Regulatory Limit for

Organ Dose 

(mrem/yr or rem)b

EFFECTIVE DOSE EQUIVALENT

Routine use and disposal of a

 single exempt unit

Accidents involving a single

 exempt unit

Normal handling and storage of

 multiple exempt units

Accidents involving multiple

 exempt units

0.004 mrem/yrc

3 mreme

0.1 mrem/yrh

10 mremj

 1 mrem/yrd

0.5 remf

15 remg

10 mrem/yri

0.5 remf

15 remg

DOSE EQUIVALENT TO SKIN

Normal use and disposal of a 

 single unit

40 mrem/yrk 15 mrem/yr

a

 See Section 2.14.1 of this report.

b

 1 mrem = 0.01 mSv; 1 rem = 0.01 Sv

c

 Dose estimate applies to a 16-hour-per-day wearer of a self-luminous wristwatches (see

Section 2.14.4.2).

d

 Limit on whole-body dose during normal use and disposal of a single exempt unit (see

Section 2.14.1).

e

 Dose estimate applies to a watch repairman following crushing of a single watch in a small

repair shop (see Section 2.14.4.5).

f

 Limit on whole-body dose when probability of failure of safety features in product is less than

one failure per year for each 10,000 exempt units distributed (see Section 2.14.1).

g

 Limit on whole-body dose when probability of failure of safety features in product is less than

one failure per year for each 1 million exempt units distributed (see Section 2.14.1).

h

 Dose estimate applies to a local parcel-delivery driver (see Section 2.14.4.1).

i

 Limit on whole-body dose during normal handling and storage of multiple exempt units (see

Section 2.14.1).

j

 Dose estimate applies to a worker exposed during a shipping accident involving the crushing

of 200 self-luminous watches (see Section 2.14.4.5).

k

 Skin dose to a 16-hour-per-day wearer of a self-luminous wristwatch (see Section 2.14.4.2).
2–213

Table 2.14.3 Potential Radiation Doses From Self-Luminous Gun Sights 

Containing 3

H

Exposure Pathway

Individual Annual

Effective

Dose Equivalent

 (mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport 0.1c 0.003

Gun sight installers 0.3d 0.007

Routine use 0.03e 4

Disposal

 Landfills

 Incinerators

<0.001f

0.004g

0.3

0.4

Accidents and misuse

 Toy for a child

 Crushing of gun sights

0.002h

 50i

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Collective doses are based on an assumed annual distribution of 100,000 handgun sights

having an effective lifetime of 10 years and initially containing 1.9 GBq (50 mCi) of gaseous 3

H

per sight. Refer to text for time period for collective dose calculations.

c

 Dose estimate applies to a local parcel-delivery driver exposed to multiple exempt units during

distribution and transport; dose estimates are less for other truck drivers, workers at truck

terminals, store clerks and customers, and members of public along truck routes (see

Section 2.14.5.1).

d

 Dose estimate applies to a factory installer who mounts and aligns 5,000 gun sights per year;

dose estimate for annual dose equivalent to skin of a factory installer is 0.3 mSv (30 mrem)

when averaged over 3 cm2

 of skin area in contact with gun sights (see Section 2.14.5.2). e

 Dose estimate applies to police officer who patrols alone in a police cruiser and carries a rifle

with a self-luminous scope containing 11 GBq (300 mCi) of 3

H (Section 2.14.5.3).

f

 Dose estimate applies to waste collectors exposed to multiple exempt units during disposal at

landfills; dose estimates are less for workers at landfills, off-site members of the public, and

future on-site residents (see Section 2.14.5.4).

g

 Dose estimate applies to waste collectors exposed to multiple exempt units during disposal at

incinerators; dose estimates are less for workers at incinerators and off-site members of the

public (see Section 2.14.5.4).

h

 Dose estimate applies to a 5-year-old child who is allowed to use an old self-luminous

gunsight as a toy during one year; dose estimate for dose equivalent to skin is 0.008 mSv

(0.8 mrem) when averaged over a skin area of 3 cm2

 in contact with gun sight (see Section

2.14.5.5).

i

 Dose estimate applies to a worker exposed during a shipping accident in a storeroom or 

cargo-handling bay involving the crushing of 1,000 handgun sights containing 1.9 GBq (50 mCi)

of 3

H each; dose estimate is 0.2 mSv (20 mrem) for a gunsmith at a small gun shop following

the crushing of a single rifle scope containing 11 GBq (300 mCi) of 3

H (see Section 2.14.5.5).
2–214

Table 2.14.4 Comparison of Estimated Maximum Individual Doses From 

Self-Luminous Gun Sights Containing 3

H and Limiting Organ Doses 

From 10 CFR 32.23

Exposure Conditionsa Maximum Individual

Dose (mrem/yr or mrem)b

Regulatory Limit For

Organ Dose

(mrem/yr or rem)b

EFFECTIVE DOSE EQUIVALENT

Routine use and disposal of a

 single exempt unit

Accidents involving a single

 exempt unit

Normal handling and storage of

 multiple exempt units

Accidents involving multiple

 exempt units

0.03 mrem/yrc

20 mreme

0.3 mrem/yrh

50 mremj

1 mrem/yrd

0.5 remf

15 remg

10 mrem/yri

0.5 remf

15 remg

DOSE EQUIVALENT TO SKIN

Normal handling and storage of

 multiple units

30 mrem/yrk 150 mrem/yr

a

 See Section 2.14.1 of this report.

b

 1 mrem = 0.01 mSv; 1 rem = 0.01 Sv.

c

 Dose estimate applies to a police officer using a self-luminous rifle scope containing 11 GBq

(300 mCi) of 3

H (see Section 2.14.5.3).

d

 Limit on whole-body dose during normal use and disposal of a single exempt unit (see

Section 2.14.1).

e

 Dose estimate applies to a gunsmith in a small gun shop following crushing of a self-luminous

rifle scope containing 11 GBq (300 mCi) of 3

H (see Section 2.14.5.5).

f

 Limit on whole-body dose when probability of failure of safety features in product is less than

one failure per year for each 10,000 exempt units distributed (see Section 2.14.1).

g

 Limit on whole-body dose when probability of failure of safety features in product is less than

one failure per year for each 1 million exempt units distributed (see Section 2.14.1).

h

 Dose estimate applies to an installer at a factory exposed to multiple self-luminous handgun

sights containing 1.9 GBq (50 mCi) of 3

H each (see Section 2.14.5.2).

i

 Limit on whole-body dose during normal handling and storage of multiple exempt units.

j

 Dose estimate applies to a worker exposed during a shipping accident involving the crushing

of 1,000 handgun sights containing 1.9 GBq (50 mCi) of 3

H (see Section 2.14.5.5). 

k

 Skin dose for an installer at a factory exposed to multiple self-luminous handgun sights

containing 1.9 GBq (50 mCi) of 3

H each (see Section 2.14.5.2).
2–215

2.15 Gas and Aerosol Detectors

2.15.1 Introduction

In 10 CFR 30.20, any person is exempt from licensing requirements to the extent that the

person receives, possesses, uses, transfers, owns, or acquires byproduct material in gas and

aerosol detectors designed to protect life or property from fire and airborne hazards. This

class exemption” was proposed on November 1, 1968 (33 FR 16089), and issued as a final

rule on April 18, 1969 (34 FR 6653).

Requirements for a license to manufacture, process, produce, or initially transfer gas and

aerosol detectors intended for use under 10 CFR 30.20 are contained in 10 CFR 32.26. 

Conditions of license are also included in 10 CFR 32.29; these include requirements for quality

control, labeling, and reporting of transfers.

Section 32.26 specifies that the license applicant must demonstrate the product will meet

certain safety criteria set forth in 10 CFR 32.27 and includes the specific information to be

submitted with the application. The safety criteria are the primary factors in controlling the

doses associated with this exemption and are described below:

In normal use and disposal of a single exempt unit and in normal handling and storage

of the quantities of exempt units that are likely to accumulate in one location during

marketing, distribution, installation, and servicing of the product, it is unlikely that the

annual dose from external exposure, or the annual dose commitment resulting from

intake of radioactive material, to a suitable sample of the group of individuals expected

to be the most highly exposed to radiation or radioactive material from the product will

exceed 0.05 millisievert (mSv) (5 mrem) to the whole body, head and trunk, active

blood-forming organs, gonads, or lens of the eye; 0.75 mSv (75 mrem) to the hands and

forearms, feet and ankles, or localized areas of the skin averaged over areas no larger

than 1 cm2

; and 0.15 mSv (15 mrem) to any other organs.

It is unlikely there will be a significant reduction in the effectiveness of the containment,

shielding, or other safety features of the product from wear and abuse that is likely to

occur in normal handling and use of the product during its useful life.

In use and disposal of a single exempt unit and in handling and storage of the quantities

of exempt units that are likely to accumulate in one location during marketing,

distribution, installation, and servicing of the product, it is unlikely that the containment,

shielding, or other safety features of the product would fail under such circumstances

that an individual would receive an external dose or internal dose commitment in excess

of 5 mSv (0.5 rem) to the whole body, head and trunk, active blood-forming organs,

gonads, or lens of the eye; 75 mSv (7.5 rem) to the hands and forearms, feet and

ankles, or localized areas of the skin averaged over areas no larger than 1 cm2

; and

15 mSv (1.5 rem) to any other organs. The likelihood is negligible that an individual

would receive an external dose or dose commitment in excess of 150 mSv (15 rem) to

the whole body, head and trunk, active blood-forming organs, gonads, or lens of the

eye; 2 Sv (200 rem) to the hands and forearms, feet and ankles, or localized areas of
2–216

the skin averaged over areas no larger than 1 cm2

; and 0.5 Sv (50 rem) to any other

organs.

These safety criteria have the effect of establishing an upper limit on the amount of radioactive

material in an exempt unit and of establishing a need to provide some degree of containment

for the material even under accident conditions. Furthermore, the probabilities associated with

the dose limits for accidental exposures ensure that the risk to individuals would be very small.

In the last set of safety criteria for accident exposures, the guidance for quantifying low and

negligible probabilities of failure of safety features in products for purposes of demonstrating

compliance with the criteria is the same as that previously described for self-luminous products

in Section 2.14.1 (i.e., a probability of failure is low” if there is not more than one failure per

year for each 10,000 exempt units distributed, and negligible” if there is not more than one

failure per year for each 1 million exempt units distributed).

The class exemption” for self-luminous products (see Section 2.14.1) set out the concept of

safety criteria (dose limits for individuals) that consumer products containing tritium (3

H), 85Kr, or

147Pm must meet under conditions of handling, storage, use, and disposal. The class

exemption” for gas and aerosol detectors carried that concept one step further by not specifying

the byproduct material that may be used in the product. Instead, it charged the manufacturer or

importer (license applicant) with exercising care in the selection of the type and quantity of the

radionuclide and in the design and construction of containment and shielding features used to

meet the safety criteria. Both class exemptions” provide for consideration of the probability

and consequences of failure of containment, shielding, or other safety features of the product. 

The safety criteria for gas and aerosol detectors are similar to the safety criteria for selfluminous products containing 3

H, 85Kr, and 241Am, except only a single set of criteria with

somewhat different dose limits is specified for normal handling, storage, use, and disposal of

gas and aerosol detectors.

Therefore, the safety criteria for gas and aerosol detectors provide: (1) radiation dose limits for

individuals from normal handling, storage, use, and disposal of these products, and

(2) radiation dose limits for individuals in conjunction with approximate associated probabilities

of occurrence for accidents. The safety criteria do not include any limits on the collective

radiation dose to members of the public.

2.15.2 Description of Items

The most commonly used products that have been distributed for use under this class

exemption” are smoke detectors containing 241Am and chemical detectors containing either

241Am or 63Ni. The radionuclide 63Ni has also been used in smoke detectors to some degree. 

The major decay mode of 241Am is alpha-particle emission, and the major decay mode of 63Ni is

beta-particle emission, thereby limiting external exposure during routine use of these devices. 

The photons emitted during the decay of 241Am have energies of only 69 keV or less, and only

very low-energy photons from bremsstrahlung are produced during the decay of 63Ni.
2–217

2.15.2.1 Smoke Detectors

For more than a decade, residential deaths from fire in the United States have declined by an

average of 4% per year (Smoke Detectors, 1994). This can be attributed to the increased use

of smoke detectors in U.S. homes during the late 1970s and early 1980s. The two most

common types of smoke detectors are optical smoke detectors (OSDs), also referred to as

photoelectric smoke detectors, and ionization chamber smoke detectors (ICSDs). OSDs work

best with smoke from smoldering fires that contain large particles, while ICSDs work best with

smoke from fast-burning fires that contain smaller particles with diameters of about 0.01 to

1 m (Sacks, 1986). Some fire prevention experts recommend installation of both OSDs and

ICSDs in the home.

An ICSD consists of an ionization chamber, electronic circuitry, a power source that is usually a

battery, an alarm mechanism, and an outer case. The ionization chamber is the main

component. It consists of a source of ionizing radiation, usually 241Am, positioned between two

oppositely charged electrodes. Alpha particles emitted during radioactive decay of the 241Am

interact with neutral air molecules flowing through the chamber and convert them to positive

ions by removal of electrons. The removed electrons then form negative ions by attachment to

other neutral molecules. The resulting positive and negative ions are attracted toward the

electrodes, causing a small, reasonably steady current between the electrodes. The electronic

circuitry monitors this current and, if the current drops below a preset level, triggers an audible

alarm.

Under normal conditions, there is an equilibrium between ion production in the chamber and ion

removal from the chamber by either striking an electrode, ion recombination, or being carried

out of the chamber by air flow. However, if the air entering the chamber contains combustion

products or smoke particles that are more massive than the air molecules, this equilibrium will

be disturbed. The smoke particles will also become charged by alpha-particle interactions,

electron capture, and combination with existing ions. The charged smoke particles move

toward the electrodes more slowly because they are more massive than the ionized air

molecules. This allows an excess of charged smoke particles to be swept out of the chamber

by air flow before reaching the electrodes. The net effect is a reduction in the current flowing

between the electrodes. When the current decreases below a preset level, the alarm is

activated to signal the very early stages of a fire.

Between 1971 and 1986, Nuclear Regulatory Commission (NRC) records (NRC, Unpublished

Reports, 1989) indicate that 124,000 ICSDs containing a total of 44 gigabecquerel (GBq)

(1.2 curie (Ci)) of 63Ni and 92 million ICSDs containing a total of 12 TBq (320 Ci) of 241Am were

sold in the United States. In 1973, only 250,000 ICSDs were sold, with most of the units going

into public and commercial buildings (NCRP 95). These units contained a total of 480 GBq

(13 Ci) of 241Am, with an average of approximately 1.9 MBq (50 Ci) per unit. From 1973 to

1978, the sales of ICSDs increased dramatically (O’Donnell et al., 1981), and the number of

homes with ICSDs increased from 10 to 77% (Hall, 1989). In 1978, approximately 14 million

ICSDs were sold, with most of the units going into homes. These units contained a total of

1.7 TBq (46 Ci) of 241Am, with an average of 0.11 MBq (3 Ci) per unit (NUREG/CR–1156). 

It is currently estimated that more than 80% of U.S., homes have 1 or more ICSDs and the total

number of ICSDs in residential use is approximately 100 million. An ICSD should work for

about 10 years, after which it should be discarded and replaced by a new smoke detector
2–218

(Smoke Detectors, 1994). Thus, the annual distribution of ICSDs in the United States currently

totals about 10 million units. The average 241Am activity per unit is approximately 37 kBq

(1 Ci), and the total number of curies of 241Am distributed annually in this product is

approximately 0.37 TBq (10 Ci).

The specific design features of the smoke detectors vary among manufacturers, but the

following features are common to all designs (O’Donnell et al., 1981). First, the source is

normally cut or punched from a composite material that consists of gold and 241Am that is hotforged onto a 0.2-mm-thick silver backing and covered by a 0.001- to 0.002-mm-thick gold foil

to form a sealed source, typically 3 to 5 mm in diameter. Next, the sealed source is either

pressed and crimped into a recess or electron beam welded within the ionization chamber. 

Finally, the ionization chamber is also sealed, except for screen-covered openings to allow gas

and smoke particulates to flow into the ICSD. The detector designs require deliberate

tampering in order to remove the 241Am source.

2.15.2.2 Chemical Detectors

Chemical detectors are used to monitor for harmful or toxic gases and a variety of vapors. 

They may be used by industry to monitor for leakage of gases such as sulfur hexafluoride,

refrigerants, and products of combustion; by the military to monitor for chemical warfare agents

such as nerve and blister gases; by hospitals to monitor for waste anesthetic gases; and by

airports to monitor for vapors from explosives or narcotic substances. The chemical detectors

may be designed for use as either a fixed monitor or a portable instrument for field

measurements. Typical chemical detectors use sources containing a total of either 0.37 GBq

(10 mCi) of 63Ni per detector or 5.9 MBq (0.16 mCi) of 241Am per detector.

A chemical detector consists typically of a detector cell, electronic circuitry, a power source, an

air pump, a heater, and an outside case. The detector cells contain a radioactive source which

is usually coiled into a cylindrical shape with the radioactive side inward. The technology used

in making the 241Am sources is similar to that used in making the 241Am sources for smoke

detectors (see Section 2.15.2.1). The 63Ni sources are made by electroplating the nickel onto

a metallic foil, which can then be formed into a cylindrical source. Tamper-resistant screws are

used to restrict human access to the radioactive sources in the chemical detectors.

Air is drawn into the detector through air flow tubes via the internal pump, heated as it is drawn

past the detector cell, and expelled as exhaust. The air flow tubes are too small in diameter to

permit human access, and bends in them block any direct radiation path from the sources. As

the heated air passes over the radioactive sources, electrons and positive ions are formed,

which cause a current in the detector cell. The free electrons are readily captured by many

gases and vapors, causing a reduction of the current in the detection cell. If the current drops

below a present level, a visual and audible alarm is given to alert the operator or others in the

vicinity of the chemical detector.

Chemical detectors are relatively new products whose total distribution is not accurately known

at present and whose distribution will probably increase quite rapidly in the future. Thus, it is

assumed here that 10,000 chemical detectors containing a total of 59 GBq (1.6 Ci) of 241Am and

10,000 chemical detectors containing a total of 3.7 TBq (100 Ci) of 63Ni are distributed annually,

and the effective lifetime of the detectors is approximately 10 years.
2–219

2.15.3 Summary of Previous Analyses and Assessments

Assessments of the potential radiological impacts on the public from smoke detectors

containing 241Am sources have been performed by Belanger et al. (NUREG/CR–1156) and

O’Donnell et al. (1981), and the results of these assessments have been summarized in NCRP

95. Belanger et al. also evaluated the potential radiological impacts on the public from smoke

detectors containing 63Ni and 226Ra sources. However, the manufacture of smoke detectors

containing 226Ra sources ceased following the development of 241Am sources for smoke

detectors, and only a relatively small number of smoke detectors containing 63Ni sources have

been distributed since the 1970s.

For smoke detectors containing 241Am sources, NCRP 95 estimated a total collective dose of

7.8 person-Sv (780 person-rem) based on the results of Belanger et al. (NUREG/CR–1156). 

This estimate assumes (1) an annual distribution and disposal of 10 million smoke detectors

containing 0.11 MBq (3 Ci) of 241Am each and (2) an effective lifetime of 10 years for

the smoke detectors during routine home usage. For these conditions, the collective dose

was distributed as follows: manufacture and distribution—0.7 person-Sv (70 person-rem), use

and maintenance—7 person-Sv (700 person-rem), and disposal—0.05 person-Sv (5 personrem). The total collective dose estimated by O’Donnell et al. (1981) was 2.3 person-Sv

(230 person-rem), or a factor of about 3 less than the value of 7.8 person-Sv (780 person-rem)

adopted by the NCRP (1987) based on the results of Belanger et al. (1979). This difference

resulted mostly from assumptions used in estimating individual and collective doses from direct

external exposures to the smoke detectors during routine home usage. The individual and

collective doses based on the external dose-rate calculations of O’Donnell et al. (1981) were

judged to be the more reliable for the following reasons.

Belanger et al. assumed that the dose rates at 1 meter from a bare 241Am foil source were

3.1 picocoulomb (pC)/kg-h (12 nanoroentgen (nR)/h), or 0.11 nSv/h (11 nrem/h), per 37 kBq

(1Ci) of 241Am and the dose rates at 1 meter from an 241Am foil source inside an assembled

smoke detector were 74 pSv/h (7.4 nrem/h), per 37 kBq (1 Ci) of 241Am. The latter value,

used in their assessment of individual and collective doses from routine use of smoke detectors

in homes, appears to have been based on the maximum measured value reported by

manufacturers. It seems to provide an overly conservative estimate of the external dose from a

smoke detector when compared with the calculated value used by O’Donnell et al. (1981). 

However, the value assumed for a bare 241Am foil source is in good agreement with a value

recently reported by Schmitt-Hannig et al. (1995). They reported that measurements on a bare

241Am foil source with a 4 m gold window indicate a dose rate at 1 meter of 0.09 nSv/h

(9 nrem/h) per 37 kBq (1 C) of 241Am. 

Schmitt-Hannig et al. also noted that calculated values that do not account for air or

self-absorption give overly conservative dose rates from a bare 241Am foil source. The 241Am

calculations of Unger and Trubey (1981, also see Table 2.1.2 of Section 2.1), for example, give

unreasonably high dose rates because the calculations ignore the self-absorption of the very

low-energy photons from 241Am in a bare foil source (and an assembled smoke detector). 

Photons with energies of only 14, 26, 33, 59, and 69 keV are emitted in the radioactive decay of

241Am with intensities of 42.7, 2.4, 0.11, 35.9, and 0.18%, respectively (Kocher, 1981).
2–220

O’Donnell et al. (1981) carefully modeled the foil source in an assembled smoke detector as a

5-mm-diameter by 0.002-mm-thick cylinder of a gold plus 241Am matrix covered by a

0.0015-mm-thick gold window. The matrix contained 0.11 MBq (3 Ci) of 241Am. The foils were

also assumed to be enclosed in a 0.254-cm-thick iron housing within the assembled smoke

detector. The dose equivalent rate at 1 meter from the assembled smoke detector was

calculated to be 39 pSv/h (3.9 nrem/h), or 13 pSv/h (1.3 nrem/h) per 37 kBq (1 Ci) of 241Am. 

This value was within the range of dose equivalent rates of 7 to 58 pSv/h (0.7 to 5.8 nrem/h)

per 37 kBq (1 Ci) of 241Am measured by manufacturers (based on information from the NRC’s

licensing files (NRC, Unpublished Reports, 1989)) and was a factor of 5.7 lower than the value

of 74 pSv/h (7.4 nrem/h) per 37 kBq (1 Ci) of 241Am assumed by Belanger et al. O’Donnell

et al. also modeled arrays of detectors (cartons and pallets) during shipment as a composite

material having an effective density of 1.4 g/cm3

. This composite was a homogeneous mixture

of 241Am, plastic, cellulose, iron, carbon, and lead chosen to approximate the exposure rates

from arrays of smoke detectors in cartons and on pallets during distribution and transport.

For chemical detectors containing 241Am and 63Ni, there are no known previously published

analyses of the potential radiological impacts on members of the public. However, licensees

would have supplied analyses to the NRC in their applications as required by 10 CFR 32.26. 

2.15.4 Present Exemption Analysis for Smoke Detectors

Table 2.15.1 presents the results of potential radiation doses to the public from an assumed

annual distribution of 10 million smoke detectors containing 37 kBq (1 Ci) of 241Am each. The

smoke detectors are assumed to have an effective lifetime of 10 years during routine home

use. If external exposure is the only significant mode of exposure, this analysis will simply use

a third of the estimates of effective dose equivalent (EDE) made by O’Donnell et al. (1981). 

This scales the results of their analysis for smoke detector activities of 0.11 MBq (3 Ci) to

smoke detector activities of 37 kBq (1 Ci) used in the current analysis. If internal exposure

is a significant mode of exposure, then 50-year committed EDEs are calculated here using dose

conversion factors from Table 2.1.2 in Section 2.1 of this report. Internal exposure by inhalation

or ingestion of 241Am from smoke detectors may occur following accidents or misuse and

disposal at landfill or incinerator sites.

2.15.4.1 Distribution and Transport

Based on the work of O’Donnell et al., the individual EDE could be 2×105

 mSv (0.002 mrem) to

stockhandlers who handle large numbers of cartons of packaged smoke detectors in wholesale

warehouses and who work near the stored cartons. Individual doses would be less for other

warehouse and truck terminal workers, store clerks and shoppers, truck drivers, and members

of the public along truck routes. The total collective EDE from 1 year’s distribution of 10 million

smoke detectors containing 37 kBq (1 Ci) of 241Am each could be 0.04 person-Sv

(4 person-rem), due almost entirely to exposure to store customers.

2.15.4.2 Routine Use

Based on the work of O’Donnell et al., it is assumed that two ICSDs were installed in each

home, 10% in bedrooms and 90% in hallways. Because the ICSDs may be installed in a home

in a variety of different ways, the following representative exposure scenario was chosen to
2–221

indicate potential radiation doses from routine use of smoke detectors: one smoke detector

installed in a hallway in the sleeping area of the home and the other installed in the master

bedroom.

Table 2.15.2 lists the exposure conditions and radiation doses from use of two 37-kBq (1-Ci)

ICSDs, one in the hallway in the sleeping area of the home and the other in the master

bedroom. If the ICSDs have a 10-year lifetime, this set of detectors could deliver the estimated

EDEs for 10 years. Purchase, installation, removal, and disposal occur only once during the

10-year lifetime of an ICSD, and the doses from these exposures were found to be insignificant

compared to those from routine use (O’Donnell et al., 1981).

A resident who purchases, installs, and maintains two smoke detectors in the home, sleeps

8 h/day, and spends 4 h/day at other activities in the home could receive an annual EDE of

1×105

 mSv (0.001 mrem). Essentially all of the dose is from the single detector located in the

master bedroom. Other family members could receive annual EDEs of less than 1×105

 mSv

(<0.001 mrem).

The collective EDE from routine home use of 10 million ICSDs has been estimated based on

the above doses for one detector in the hallway and one in the master bedroom. The

collective EDE would total approximately 1 person-Sv (100 person-rem) for the 5 million homes

(see Table 2.15.3).

Major commercial users of smoke detectors are motels and hotels (Bill, 1990). If a hotel maid

is exposed to a smoke detector for 2000 h/yr at an average distance of 3 meters from a smoke

detector unit, not necessarily the same unit, containing 37 kBq (1 Ci) of 241Am, then the

maximum individual annual EDE would be less than 1×105

 mSv (<0.001 mrem).

2.15.4.3 Disposal

The potential individual and collective doses from disposal in landfills and by incineration are

estimated using the generic methodology in Appendix A.2 of this report. In applying this

methodology, it is assumed that the 241Am sources remain intact during waste collection and

burial at landfills. Thus, the pathway for inhalation is not operative in the case of off-site

members of the public near landfills, and the pathways for ingestion or inhalation are not

operative in the case of workers at landfills or waste collectors at both landfills and incinerators. 

The average activity of 241Am in each detector is assumed to be 37 kBq (1 Ci) and the total

activity of 241Am in the 10 million smoke detectors disposed of annually is assumed to be

0.37 TBq (10 Ci).

For disposal at landfills, the annual EDE equivalent would be about 0.002 mSv (0.2 mrem) to

waste collectors. The annual individual doses to workers at landfills, off-site members of the

public near the landfills, and future on-site residents at landfills would be substantially less. The

total collective dose to the public would be about 0.3 person-Sv (30 person-rem), due almost

entirely to exposure to future on-site residents at landfills for 1000 years after loss of

institutional controls over the sites. If exposure to future on-site residents is not considered, the

collective EDE would be 0.008 person-Sv (0.8 person-rem), due mainly to exposure to waste

collectors at landfills.
2–222

For disposal at incinerators, the annual EDE would be 0.01 mSv (1 mrem) to waste collectors. 

The individual doses to workers at incinerators and off-site members of the public are

substantially less. The total collective dose to the public would be 0.005 person-Sv

(0.5 person-rem), due mainly to exposure to off-site members of the public to airborne releases

from incinerators.

2.15.4.4 Accidents and Misuse

In the case of accidents, the following is considered: (1) a residential fire involving 2 smoke

detectors, (2) a transportation fire involving a typical shipment of 7,200 smoke detectors

(O’Donnell et al., 1981), and (3) a manufacturer’s warehouse fire involving 36,000 smoke

detectors (i.e., five typical shipments). Also, a release factor of 0.01% is assumed for the 241Am

source foils inside the smoke detectors (NUREG/CR–0403; Hall and Hunt, NUREG/CP–0001;

Niemeyer, 1969). Based on these assumptions and the generic accident methodology of

Appendix A.1, the individual doses can be summarized as follows:

For a person escaping from a residential fire or a neighborhood hero trying to rescue a

person from a residential fire, the individual EDE from two smoke detectors containing a

total of 74 kBq (2 Ci) of 241Am could be 4×105

 mSv (0.004 mrem).

For a firefighter wearing a respirator at a residential fire, the individual EDE from two

smoke detectors containing a total of 74 kBq (2 Ci) of 241Am would be less than

1×105

 mSv (<0.001 mrem)

For a firefighter wearing a respirator at a transportation fire, the individual EDE from

7,200 smoke detectors containing a total of 266 MBq (7.2 mCi) could be 0.003 mSv

(0.3 mrem).

For a firefighter wearing a respirator at a warehouse fire, the individual EDE from 36,000

smoke detectors containing a total of 1.3 GBq (36 mCi) could be 0.003 mSv (0.3 mrem).

For the cleanup following a residential fire, the individual EDE could be about 5×105

 mSv

(0.005 mrem) for an individual who works for 8 hours and does not wear a respirator, and for

fire inspectors who do not wear respirators and spend about 25% of their time inspecting mostly

residential fires (62.5 working days), the annual individual EDE could be about 0.003 mSv

(0.3 mrem) or approximately the same radiation dose as that to a firefighter who fights a single

transportation or warehouse fire.

In the case of misuse, the following scenarios are considered: (1) exposure to a teacher who

removes an 241Am source from a smoke detector for use in classroom demonstrations about

radioactivity, and (2) exposure to a person who removes an 241Am source from a smoke

detector and subsequently swallows it by accident. To estimate the potential radiation dose to

the teacher, it is assumed that the teacher stores the 241Am source in a convenient location in a

classroom and is exposed at an average distance of about 1 meter from the source for

1000 h/yr. The EDE rate at 1 meter from a 37-kBq (1-Ci) source of 241Am without regard for

shielding by other materials is about 0.09 nSv/h (9 nrem/h) (see Section 2.15.3) and the annual

EDE to the teacher could be 9×105

 mSv (0.009 mrem). Also, the teacher handles the 241Am

source for 10 h/yr during classroom demonstrations at a nominal 1 cm distance (due to the size

of the disk upon which the source is mounted) then the dose equivalent to a small area of skin
2–223

on the hand is about 0.3 mSv (30 mrem), based on a calculation using photon fluence and

tissue (skin) energy absorption. Assuming a 10 cm2 exposed skin area out of a total skin area

of 1.8×104

 cm2 (ICRP 26) and a skin weighting factor of 0.01 (ICRP 60), the calculated EDE

would be less than 1×105

 mSv (< 0.001 mrem).

The inadvertent ingestion of an 241Am smoke detector source by a consumer is not considered

a realistic scenario, due to the source being contained within a protective housing and mounted

on a larger metal or plastic disk. However, ingestion of an 241Am smoke detector source has

occurred during the manufacturing process. For this situation, the potential radiation to a

person who swallows an 241Am source is based on data from an actual case history reported by

Rundo et al. (1977). From these data, it was determined that an 241Am source foil lost 15.6% of

its original activity in the GI tract. If the 15.6% is applied as the quantity available for uptake

and the f1 factor of 1×103

 assumed by Eckerman et al. (EPA–520/1–88–020) is used, then the

calculated individual EDE is about 6 mSv (600 mrem) to a person who swallows an 241Am

source foil with an activity of 37 kBq (1 Ci). It needs to be noted that the 241Am source foils in

smoke detectors are inaccessible under normal conditions of distribution, use, and disposal,

and a person would have to deliberately destroy a smoke detector to remove the source foil for

other purposes. Thus, the removal of an 241Am source from a smoke detector should be

considered as a potential accident having a negligible” probability of failure of the safety

features (i.e., less than one failure per year per 1 million units distributed) (see Section 2.15.1).

2.15.4.5 Safety Criteria 

Table 2.15.4 compares the results of the current dose assessment for smoke detectors

containing 241Am sources with the safety criteria for gas and aerosol detectors set forth in

10 CFR 32.27 (see Section 2.15.1). First, the limiting 10 CFR 32.27 values for dose to the

whole body are compared with the maximum estimates of individual EDE, and second, the

limiting 10 CFR 32.27 values for dose to other organs are compared with the maximum

estimates of dose to red marrow (i.e., an active blood-forming organ) and bone surfaces

(i.e., the endosteal cells on the bone surfaces). The dose equivalent to the red marrow of the

skeleton and the endosteal cells on the bone surfaces of the skeleton are approximately 1.5

and 20 times greater, respectively, than the EDE to the whole body (EPA–520/1–88–020). The

endosteal cells provide an important function as the bone-forming and bone-resorbing cells of

the skeleton (Matthews, 1980). It is important to note that the maximum individual doses to the

whole body (or EDE) and to the red marrow or bone surfaces (or organ doses) are less than the

organ dose limits set forth in the safety criteria of 10 CFR 32.27 if, and only if, the swallowing of

an 241Am source from a smoke detector is considered as a potential accident having a 

negligible” incidence of failure of the safety features (i.e., less than 1 failure per year per 1

million exempt units distributed) (see Section 2.15.1).

2.15.5 Present Exemption Analysis for Chemical Detectors

Table 2.15.5 presents results of the current assessment of potential radiation dose to the public

from chemical detectors containing either 0.37 GBq (10 mCi) of 63Ni each or 5.9 MBq

(0.16 mCi) of 241Am each. As noted previously, chemical detectors are relatively new products

whose total distribution is not accurately known at present and whose distribution will

undoubtedly increase very rapidly in the future. Thus, it is assumed here that 10,000 sources

containing a total of 59 GBq (1.6 Ci) of 241Am and 10,000 sources containing a total of 3.7 TBq
2–224

(100 Ci) of 63Ni could be distributed annually in these detectors, which have an effective lifetime

of about 10 years.

In the case of 241Am, the dose equivalent rate at 10 cm from an unshielded 5.9-MBq (0.16-Ci)

source is about 0.15 Sv/h (150 rem/h) (see Section 2.15.2), and the dose equivalent rate at

10 cm from a partially shielded 5.9-MBq (0.16-mCi) source is about 300 nSv/h, (30 rem/h)

according to the chemical director manufacturer. The manufacturer also determined that the

dose equivalent rates at 30 cm and 100 cm from the partially shielded source in a chemical

detector were about 40 nSv/h (4 rem/h) and 4 nSv/h (0.4 rem/h), respectively. In the case of

63Ni, the dose equivalent rate from an unshielded 0.37-GBq (10-mCi) source due to

bremsstrahlung is about 5 nSv/h (0.5 rem/h) at 30 cm and 0.5 nSv/h (0.05 rem/h) at 100 cm,

using CONDOS II (Computer Codes, O’Donnell et al., 1981) and correcting for over estimation

as discussed in Appendix A.4. For the current analysis it is assumed that partial shielding of

the 63Ni source by its protective housing, the external case of the detector, and various other

detector components will reduce the dose equivalent rates due to the very low-energy

bremsstrahlung from the 63Ni source to essentially zero (0) for routine use and handling,

distribution and transport, and disposal.

During normal use and handling of the chemical detectors, it is assumed that inhalation or

ingestion of the 63Ni and 241Am does not occur, and the only significant mode of exposure is

external irradiation of the body. Ingestion and inhalation may occur, however, during disposal

of the chemical detectors or during accidents and misuse.

2.15.5.1 Distribution and Transport

To estimate the potential radiation doses to the public from the distribution and transport of

chemical detectors containing sources of 241Am, the generic methodology developed in

Appendix A.3 is used. In applying this methodology, it is assumed that the chemical detectors

are shipped primarily by a parcel-delivery service, and that a local driver in a large delivery truck

picks up the chemical detectors and takes them to a local terminal for shipment to other local

terminals for delivery to customers. It is assumed further that (1) each chemical detector

passes through an average of four regional terminals before reaching its final destination, (2)

the radiation exposure to workers in both local and regional terminals are the same as those

estimated for workers in a large warehouse, and (3) the local parcel-delivery driver could pick

up an average of 40 chemical detectors per day (250 day/yr) from the same manufacturer. 

Based on these assumptions and the generic methodology in Appendix A.3 for the 241Am

sources, the annual individual EDE to the local parcel-delivery driver could be 0.02 mSv

(2 mrem). Individual doses to other truck drivers, workers in truck terminals, and members of

the public along truck routes would be significantly less. The collective EDEs could be

3×104

 person-Sv (0.03 person-rem). These dose estimates are very conservative because the

generic methodology does not account for the partial shielding of the 241Am sources within the

chemical detectors.

If partial shielding of the sources is considered (see Section 2.15.5), then the above individual

and collective doses from 1 year’s distribution of 10,000 chemical detectors containing 59 GBq

(1.6 Ci) of 241Am each would be a factor of 5 less, or about 0.004 mSv (0.4 mrem) and

6×105

 person-Sv (0.006 person-rem), respectively.
2–225

2.15.5.2 Routine Use

To estimate potential radiation doses during routine use of the chemical detectors, it is

assumed that 50% of the instruments are portable devices and the other 50% are fixed devices. 

In the case of the portable devices, it is assumed that an operator carries the instrument for

500 h/yr at an average distance of 30 cm from the body and that the operator is also exposed

for 500 h/yr at an average distance of 1 meter while using and transporting the instrument. In

case of the fixed devices, it is assumed that two individuals are exposed for 1000 h/yr at an

average distance of 1 meter from each of the instruments. 

For the portable instruments containing 241Am sources (see Section 2.15.5), it is estimated that

the annual individual EDE to the operator could be (4 rem/h) × (500 h/yr) + (0.4 rem/h) ×

(500 h/yr), or 0.02 mSv (2 mrem). The collective EDE to all operators over the 10-year effective

lifetime of the 5000 portable instruments could be 1 person-Sv (100 person-rem). For the fixed

instruments containing 241Am sources, it is estimated that the annual individual EDE could be

(0.4 rem/h) × (1000 h/yr), or 0.004 mSv (0.4 mrem). The collective EDE from exposure to the

5000 fixed instruments over the 10-year effective lifetime of the instruments could be

0.4 person-Sv (40 person-rem).

For both portable and fixed chemical detectors containing 241Am sources, the estimated total

collective EDE during routine use over the 10-year effective lifetime of the instruments is

1 person-Sv (100 person-rem).

2.15.5.3 Disposal 

To estimate the potential radiation doses from landfill disposal and incineration of the 241Am

sources in chemical detectors, the analysis uses the generic disposal methodology in Appendix

A.2 and the same assumptions as used for smoke detectors in Section 2.15.4.3. Sources are

assumed to remain intact during waste collection (i.e., no ingestion or inhalation pathway). 

For the disposal of 1 year’s distribution of 10,000 chemical detectors containing a total of

59 GBq (1.6 Ci) of 241Am, we can summarize the individual and collective doses as follows:

At landfills, the annual EDE to a waste collector could be 3×104

 mSv (0.03 mrem). 

Individual doses are less to workers at landfills and off-site members of the public. The

total collective dose to all waste collectors, workers at landfills, and members of the

public could be 0.05 person-Sv (5 person-rem), due almost entirely to exposure to future 

on-site residents for 1000 years after loss of institutional controls over the sites. If

exposure to future on-site residents is not considered, the collective EDE would be

0.001 person-Sv (0.1 person-rem), due mainly to exposure to waste collectors at

landfills.

At incinerators, the annual EDE to a waste collector could be 0.002 mSv (0.2 mrem). 

Individual doses are significantly less to workers at incinerators and off-site members of

the public. The collective EDE to all waste collectors, workers at incinerators, and

members of the public could be 8×104

 person-Sv (0.08 person-rem), due mainly to

exposure to off-site members of the public to airborne releases from the incinerators.
2–226

For the disposal of chemical detectors containing 63Ni, doses are essentially zero (0), since

there is no photon component for 63Ni decay and the sources are assumed to remain intact

(i.e., no ingestion or inhalation pathway).

2.15.5.4 Accidents and Misuse

In the case of accidents, the following is considered: (1) a transportation fire involving a typical

shipment of 40 chemical detectors (see Section 2.15.5.1), and (2) a manufacturer’s warehouse

fire involving 200 chemical detectors (i.e., five typical shipments). Also, a release factor of

0.01% is assumed for the 63Ni and 241Am sources inside the chemical detectors. Based on

these assumptions and the generic accident methodology of Appendix A.1, the individual doses

can be summarized as follows:

• For a firefighter wearing a respirator at a transportation fire, the individual EDE from 40

chemical detectors containing a total of 15 GBq (0.4 Ci) of 63Ni would be less than

1×105

 mSv (<0.001 mrem) and the individual EDE from 40 chemical detectors

containing a total of 0.24 GBq (6.4 mCi) of 241Am could be 0.003 mSv (0.3 mrem).

For a worker who is involved in the cleanup following a transportation fire and does not

wear a respirator, the individual EDE from the 40 chemical detectors containing 63Ni

could be 1×105

 mSv (0.001 mrem), and the individual EDE from the 40 chemical

detectors containing 241Am could be 0.3 mSv (30 mrem).

For a firefighter wearing a respirator at a warehouse fire, the individual EDE from 200

chemical detectors containing a total of 74 GBq (2 Ci) of 63Ni would be less than

1×105

 mSv (<0.001 mrem), and the individual EDE from 200 chemical detectors

containing a total of 1.2 GBq (32 mCi) of 241Am could be 0.002 mSv (0.2 mrem).

For a worker who is involved in the cleanup following a warehouse fire and does not

wear a respirator, the individual EDE from the 200 chemical detectors containing 63Ni

could be 6×105

 mSv (0.006 mrem), and the individual EDE from the 200 chemical

detectors containing 241Am could be 0.2 mSv (20 mrem).

In the case of misuse, the exposure to a person who finds and vandalizes a chemical detector

containing a 5.9-MBq (0.16-mCi) source of 241Am is considered. It is assumed that the person

deliberately destroys the chemical detector out of curiosity and handles the 241Am source for

30 minutes before discarding both the source and other detector parts. Assuming that the hand

exposure is at a nominal 1 cm distance (due to the size of the disk upon which the source is

mounted) then the dose equivalent to a small area of skin on the hand is about 2 mSv

(200 mrem), based on a calculation using photon fluence and tissue (skin) energy absorption. 

Assuming a 10 cm2

 exposed skin area out of a total skin area of 1.8×104 cm2 (ICRP 26) and a

skin weighting factor of 0.01 (ICRP 60), the calculated EDE would be 1×105

 mSv

(0.001 mrem). For potential ingestion of material due to the handling of a 0.37-GBq (10-mCi)

63Ni source or a 5.9-MBq (0.16-mCi) 241Am source, the generic accident methodology

developed in Appendix A.2 is applied for spills of radioactive material in the form of a powder. It

is assumed, first, that 10% of the material on the source is deposited on the body and, second,

that 0.1% of this deposited material is ingested before it is removed from the body by washing. 

Thus, the EDE could be 0.006 mSv (0.6 mrem) from the ingestion of 63Ni and 0.6 mSv

(60 mrem) from ingestion of 241Am.
2–227

2.15.5.5 Safety Criteria

Tables 2.15.6 and 2.15.7 compare results of our present dose assessment for chemical

detectors containing 63Ni and 241Am sources with the safety criteria for gas and aerosol

detectors set forth in 10 CFR 32.27 (see Sect. 2.15.1). The comparisons for chemical

detectors containing 241Am sources are similar to those for smoke detectors in Table 2.15.4. 

The comparisons for chemical detectors containing 63Ni sources are selected to reflect the

different metabolic behavior of 63Ni in the body. In the case of 63Ni, the dose equivalents to all

organs of the body from inhalation are essentially the same as the EDE, but the dose

equivalent to the colon (i.e., large intestines) from ingestion is about 4 times greater than either

the dose equivalents to all other organs of the body or EDE (Private communication with Keith

F. Eckerman, Oak Ridge National Laboratory, Oak Ridge, TN, July 1997). It is important to

note from comparisons in Tables 2.15.6 and 2.15.7 that the maximum estimates of individual

dose to the whole body (or EDE) and dose equivalent to various organs of the body are less

than the dose limits set forth in 10 CFR 32.27. In the case of 241Am, however, the estimate of

dose equivalent to bone surfaces of a person who vandalizes a chemical detector and ingests

radioactive material due to contamination of the hands is very close to the dose limit for a

potential accident having a low” incidence of failure of the safety features (i.e., less than one

failure per year per 10,000 exempt units distributed) (see Section 2.15.1).

2.15.6 Summary

The results of the current assessment of potential radiation doses to the public from smoke

detectors containing 241Am are shown in Table 2.15.1. These results are based on a 10-year

effective lifetime for the smoke detectors and an annual distribution of 10 million smoke

detectors containing 37 kBq (1 Ci) of 241Am each. In Table 2.15.4 the results from the current

assessment for smoke detectors are compared further with the safety criteria for gas and

aerosol detectors. These comparisons show that the maximum estimates of individual doses

for the current analysis are within the applicable dose limits set forth in the safety criteria of

10 CFR 32.27.

An annual collective dose of 2 person-Sv (200 person-rem) from smoke detectors containing

241Am was estimated previously by O’Donnell et al. (1981). NCRP 95 estimated a collective

dose from smoke detectors containing an 241Am source of 7.8 person-Sv (780 person-rem)

based on the results of Belanger et al. (NUREG/CR–1156). These differences can be

attributed mostly to the different assumptions used in estimating both individual and collective

doses from exposure to smoke detectors during routine use in homes (see Section 2.15.3).

The collective dose estimated in this study for smoke detectors containing 241Am sources is

1 person-Sv (100 person-rem), or about half that estimated in the previous study by O’Donnell

et al. (1981). The difference can be attributed mostly to a decrease in the average 241Am

content of a smoke detector from about 0.11 MBq (3 Ci) in the early 1980s to about 37 kBq

(1 Ci) at present (see Section 2.15.2.1).

Also presented are the results from a completely new assessment of the potential radiation

doses to the public for an assumed annual distribution of 10,000 chemical detectors each

containing 0.37 GBq (10 mCi) of 63Ni each and 10,000 chemical detectors each containing

5.9 MBq (0.16 mCi) of 241Am each (see Table 2.15.5). The results of this assessment, which
2–228

are based on an effective lifetime of 10 years for the chemical detector units, suggest a total

collective dose of about 2 person-Sv (200 person-rem).

The results of the completely new assessment of potential radiation dose to the public from

chemical detectors containing 241Am or 63Ni sources are also compared with the safety criteria

for gas and aerosol detectors in Tables 2.15.6 and 2.15.7. These comparisons show that the

maximum estimates (for the new assessment) of individual dose to members of the public from

chemical detectors containing 63Ni and 241Am sources are less than the applicable dose limits

set forth in the safety criteria of 10 CFR 30.27.
2–229

Table 2.15.1 Potential Radiation Doses From Smoke Detectors Containing 241Am

Exposure Pathway

Individual Annual Effective

 Dose Equivalent

 (mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport 0.002c 4 

Routine use 0.002d 100

Disposal in ordinary trash

 Landfills

 Incinerators

0.2e

1f

30

0.5

Accidents and misuse

 Classroom use of source

 Fire

 Swallowing of source

0.009g

0.3h

600i

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Collective doses are based on an assumed annual distribution of 10 million smoke detectors

containing 37 kBq (1 Ci) of 241Am each and having a 10-year effective lifetime. c

 Dose estimate applies to stockhandlers exposed to multiple exempt units in wholesale

warehouses during distribution and transport; dose estimates are less for other warehouse and

truck terminal workers, truck drivers, store clerks and customers, and members of the public

along truck routes (see Section 2.15.4.1).

d

 Dose estimate applies to residents who are exposed to two smoke detectors during routine

home use for 12-24 hours per day; dose estimates are less for family members who are

exposed for less than 12 hours per day (see Section 2.15.4.2).

e

 Dose estimate applies to waste collectors exposed to multiple smoke detectors during

disposal at landfills; dose estimates are significantly less for workers at landfills, off-site

members of the public, and future on-site residents are significantly less (see Section 2.15.4.3).

f

 Dose estimate applies to waste collectors exposed to multiple smoke detectors during disposal

at incinerators; dose estimates are significantly less for other workers at incinerators and offsite members of public (see Section 2.15.4.3).

g

 Dose estimate applies to a teacher who removes the 241Am source from a smoke detector unit

and stores the unshielded source in a classroom (see Section 2.15.4.4).

h

 Dose estimate applies to a firefighter who is exposed to multiple smoke detector units during a

transportation or warehouse fire (see Section 2.15.4.4).

i

 Dose estimate applies to a person who removes a source from a smoke detector and

accidentally swallows it (see Section 2.15.4.4).
2–230

Table 2.15.2 Exposure Conditions and Radiation Doses From Routine Home Use

of Two Smoke Detectors Containing 0.037 MBq (1 Ci) of 241Am Each a

Exposed

Person

Exposure

Activity Sources

Duration of

Exposure

(h/yr)

Distance From

Source

(m)

Individual Dose

Over First Year

(mrem)b

Resident Purchase

Transport

Install

Maintain

Sleep

Other

 Total:

2 detectors

2 detectors

2 detectors

2 detectors

1 detector

1 detector

2 detectors

0.5

0.5

0.5

2

2,920

2,920

1,460

 0.3

1

1

1

2

6

6

<0.001

<0.001

<0.001

<0.001

0.001

<0.001

<0.001

0.001

Other family

 members

Sleep

Other

Total:

2 detectors

2 detectors

2,920

1,460

6

6

<0.001

<0.001

<0.001

a

 Adapted from O’Donnell et al. (1981).

b

 1 mrem = 0.01 mSv.
2–231

Table 2.15.3 Potential Radiation Doses From Routine Home Use of 10 Million 

Smoke Detectors Containing 0.037 MBq (1 Ci) Each of 241Am a

Exposed

Person

Individual Dose

Over First Year

(mrem)b

Individual Dose

Over 10 Years

(mrem)b

Number of

Exposed

Persons

Collective Dose

Over 10 Years

(person-rem)b

 Resident 0.001 0.01

 

 10,000,000

 

 100

 

a

 Scenario is for one detector in a hallway in the sleeping area of a home and one detector in

the master bedroom.

b 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.
2–232

Table 2.15.4 Comparison of Estimated Maximum Individual Doses From Smoke

Detectors Containing 37 kBq (1 Ci) of 241Am and Limiting Organ Doses 

From 10 CFR 32.27

Exposure Conditionsa Maximum Individual Doseb

(mrem, rem, or mrem/yr)c

Regulatory Limit for

Organ Dose

(mrem, rem, or mrem/yr)c

Effective dose equivalent

 Routine use and disposal of a

 single exempt unit

 Accidents involving a single 

 exempt unit

 Normal handling and storage of

 multiple exempt units

 Accidents involving multiple 

 exempt units

0.001mremd

600 mreme

0.002 mrem/yrh

0.3 mremi

5 mrem/yr

0.5 remf

 15 mremg

5 mrem/yr

0.5 remf

 15 mremg

Dose equivalent to red marrow

 Accidents involving a single

 exempt unit

 Accidents involving multiple

 exempt units

800 mremj

0.5 mremk

0.5 remf

 15 remg

0.5 mremf

 15 remg

Dose equivalent to bone surfaces

 Accidents involving a single

 exempt unit

 Accidents involving multiple

 exempt units

10 remj

6 mremk

1.5 reml

 50 remm

1.5 reml

 50 remm

See the following page for footnotes.
2–233

Footnotes to Table 2.15.4

a

 See Section 2.15.4 of this report.

b

 Maximum individual dose based on the typical quantity used (i.e., 37 kBg (1 Ci) of 241Am). 

However, the activity of the source could be increased until the regulatory limit for organ dose is

reached.

c

 1 mrem = 0.01 mSv; 1rem = 0.01 Sv.

d

 Dose estimate applies to spouse of a resident who is exposed to a single smoke detector

during routine home use (see Section 2.15.4.2).

e

 Dose estimate applies to a person who removes an 241Am source from a smoke detector and

subsequently swallows it (see Section 2.15.4.4).

f

 Limit on whole-body dose and dose to active blood-forming organs when probability of failure

of safety features in product is less than one failure per year for each 10,000 exempt units

distributed (see Section 2.15.1).

g

 Limit on whole-body dose and dose to active blood-forming organs when probability of failure

of safety features in product is less than one failure per year for each 1 million exempt units

distributed (see Section 2.15.1).

h

 Dose estimate applies to stockhandlers exposed to multiple smoke detectors at wholesale

warehouses (see Section 2.15.4.1).

i

 Dose estimate applies to a firefighter who is exposed to radioactive material from multiple

smoke detector units at a transportation or warehouse fire (see Section 2.15.4.4).

j

 Bone surface dose or red marrow dose to a person who removes an 241Am source from a

smoke detector and subsequently swallows it (see Section 2.15.4.4).

k

 Bone surface dose or red marrow dose to a firefighter exposed to radioactive material from

multiple smoke detector units at a transportation or warehouse fire (see Section 2.15.4.5).

l

 Limit on other organ” doses when probability of failure of safety features in product is less

than one failure per year for each 10,000 units distributed (see Section 2.15.1).

m Limit on other organ” doses when probability of failure of safety features in product is less

than one failure per year for each 1 million exempt units distributed (see Section 2.15.1).
2–234

Table 2.15.5 Potential Radiation Doses From Chemical Detectors

Containing 241Am

Exposure Pathway

Individual Annual Effective

 Dose Equivalent

(mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport 0.4c 0.006

Routine use 2d 100

Disposal

 Landfills

 Incinerators

0.03e

0.2f

5

0.08

Accidents and misuse

 Fire

 Vandalism of detector unit

30g

60h

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Collective doses are based on (1) an assumed annual distribution of 10,000 chemical

detectors containing 5.9 MBq (0.16 mCi) of 241Am each and (2) an assumed 10-year effective

lifetime for the detectors.

c

 Dose estimate applies to local parcel-delivery drivers exposed to multiple exempt units in

wholesale warehouses during distribution and transport; dose estimates are significantly less

for other truck drivers, workers in truck terminals, and members of public along truck routes

(see Section 2.15.5.1).

d

 Dose estimate applies to a user of a portable chemical detector containing an 241Am source;

dose estimates are less for a user of a fixed chemical detector containing an 241Am source (see

Section 2.15.5.2).

e

 Dose estimate applies to waste collectors exposed to multiple chemical detectors during

disposal at landfills; dose estimates are significantly less for workers at landfills, off-site

members of the public, and future on-site residents (see Section 2.15.5.3). 

f

 Dose estimate applies to waste collectors exposed to multiple chemical detectors during

disposal at incinerators; dose estimates are significantly less for other workers at incinerators

and off-site members of the public (see Section 2.15.5.3).

g

 Dose estimate applies to a worker who are involved in the cleanup following the transportation

fire involving multiple chemical detectors with 241Am sources (see Section 2.15.5.4).

h

 Dose estimate applies to ingestion of 241Am by a person who finds and destroys a chemical

detector containing an 241Am source, (see Section 2.15.4.4).
2–235

Table 2.15.6 Comparison of Estimated Maximum Individual Doses 

From Chemical Detectors Containing 5.9 MBq (0.16 mCi) of 241Am 

and Limiting Organ Doses From 10 CFR 32.27

Exposure Conditionsa Maximum Individual Doseb

(mrem, rem, or mrem/yr)c

Regulatory Limit for

Organ Dose

(mrem, rem, or mrem/yr)c

Effective dose equivalent

 Routine use and disposal of a

 single exempt unit

 Accidents involving a single 

 exempt unit

 Normal handling and storage of

 multiple exempt units

 Accidents involving multiple 

 exempt units

2 mrem/yrd

60 mremf

0.4 mrem/yri

30 mremk

5 mrem/yre

0.5 rem g

15 remh

5 mrem/yrj

0.5 rem g

15 remh

Dose equivalent to red marrow

 Accidents involving a single

 exempt unit

 Accidents involving multiple

 exempt units

90 mreml

40 mremm

0.5 remg

15 remh

0.5 mremg

 15 remh

Dose equivalent to bone surfaces

 Accidents involving a single

 exempt unit

 Accidents involving multiple

 exempt units

1 reml

500 mremm

1.5 remn

50 remo

1.5 remn

50 remo

See following page for footnotes.
2–236

Footnotes to Table 2.15.6

a

 See Section 2.15.4 of this report.

b

 Maximum individual dose based on the typical quantity used (i.e., 5.9 mBq (160 mCi) of

241Am). However, the activity of the source could be increased until the regulatory limit for

organ dose is reached.

c

 1 mrem = 0.01 mSv; 1 rem = 0.01 Sv.

d

 Dose estimate applies to an operator who is exposed to a single chemical detector unit during

routine use (see Section 2.15.5.2).

e

 Limit on whole-body dose and active blood-forming organs during normal use and handling of

a single exempt unit (see Section 2.15.1).

f

 Dose estimate applies to a person who vandalizes a chemical detector unit and ingests

radioactive material due to contamination of hands (see Section 2.15.5.4).

g

 Limit on whole-body dose and dose to active blood-forming organs when probability of failure

of safety features in product is less than one failure per year for each 10,000 exempt units

distributed (see Section 2.15.1).

h

 Limit on whole-body dose and dose to active blood-forming organs when probability of failure

of safety features in product is less than one failure per year for each 1 million exempt units

distributed (see Section 2.15.1).

i

 Dose estimate applies to local parcel-delivery driver who is exposed to multiple chemical

detector units during distribution and transport (see Section 2.15.5.1).

j

 Limit on whole-body dose and dose to active blood-forming organs during normal handling and

storage of multiple exempt units (see Section 2.15.1).

k

 Dose estimate applies to a worker who inhales radioactive material from multiple chemical

detector units during cleanup after a transportation fire (see Section 2.15.5.4).

l

 Bone surface dose or red marrow dose to a person who vandalizes a chemical detector unit

and ingests radioactive material due to contamination of hands (see Section 2.15.5.4)

m Bone surface dose or red marrow dose to worker who inhales radioactive material from

multiple chemical detector units during cleanup after a transportation fire (see Section 2.15.5.4).

n

 Limit on other organ” doses when probability of failure of safety features of the product is less

than one failure per year for each 10,000 exempted units distributed (see Section 2.15.1).

o

 Limit on other organ” doses when probability of failure of safety features of the product is less

than one failure per year for each 1 million exempt units distributed (see Section 2.15.1).
2–237

Table 2.15.7 Comparison of Estimated Maximum Individual Doses 

From Chemical Detectors Containing 0.37 GBq (10 mCi) of 63Ni 

and Limiting Organ Doses From 10 CFR 32.27

Exposure Conditionsa Maximum Individual 

Doseb (mrem or rem)c

Regulatory Limit for

Organ Dose

(mrem/yr or rem)c

Effective dose equivalent

 Routine use and disposal of a

 single exempt unit

 Accidents involving a single 

 exempt unit

 Normal handling and storage of

 multiple exempt units

 Accidents involving multiple 

 exempt units

0d

0.6 mremf

0i

0.01 mremk

5 mrem/yre

0.5 remg

15 remh

5 mrem/yrj

0.5 remf

15 remg

Dose equivalent to critical organl

 Accidents involving a single 

 exempt unit

 Accidents involving multiple 

 exempt units

 3 mremm

14 remp

1.5 remn

50 remo

1.5 remn

50 remo

See following page for footnotes.
2–238

Footnotes to Table 2.15.7

a

 See Section 2.15.5 of this report.

b

 Maximum individual dose based on the typical quantity used (i.e., 0.37 GBq (10 mCi) of 63Ni). 

However, the activity of the source could be increased until the regulatory limit for organ dose is

reached.

c

 1 mrem = 0.01 mSv; 1 rem = 0.01 Sv.

d

 Dose estimate applies to an operator who is exposed to a single chemical detector unit during

routine use. As discussed in Section 2.15.4.2, EDE during normal use and handling is

essentially zero (0).

e

 Limit on whole-body dose during normal use and disposal of a single exempt unit (see

Section 2.15.1).

f

 Dose estimate applies to a person who vandalizes a chemical detector unit and ingests source

material due to contamination of hands (see Section 2.15.5.4).

g

 Limit on whole-body dose when probability of failure of safety features in product is less than

one failure per year for each 10,000 exempt units distributed (see Section 2.15.1).

h

 Limit on whole-body dose when probability of failure of safety features in product is less than

one failure per year for each 1 million exempt units distributed (see Section 2.15.1).

i

 Dose estimate applies to a local parcel-delivery driver exposed to multiple chemical detector

units during distribution and transport. As discussed in Section 2.15.5.1, dose during routine

use and handling is essentially zero (0).

j

 Limit on whole-body dose during normal handling and storage of multiple exempt units (see

Section 2.15.1).

k

 Dose estimate applies to a worker who inhales radioactive material from multiple chemical

detector units during cleanup after a transportation fire (see Section 2.15.5.4).

l

 The colon is the critical organ for accident involving a single unit and the lung is the critical

organ for an accident involving multiple units.

m Colon dose to a person who vandalizes a chemical detector and ingests radioactive material

due to contamination of hands (see Section 2.15.5.4).

n

 Limit on other organ” doses when probability of failure of safety features in product is less

than one failure per year for each 10,000 exempt units distributed (see Section 2.15.1).

o

 Limit on other organ” doses when probability of failure of safety features in product is less

than one failure per year for each one million exempt units distributed (see Section 2.15.1).

p

 Colon dose to a worker who inhales radioactive material from multiple exempt units during

cleanup after a transportation fire (see Section 2.15.5.4).
3–1

3 EXEMPTIONS FOR SOURCE MATERIAL

3.1 Properties of Source Material

3.1.1 Introduction

This subsection provides an introduction to the following subsections of Section 3, that present

the results of assessments of radiological impacts on the public from products or materials

containing exempted amounts of source material, i.e., uranium or thorium. Information

presented in this section includes the definition of source material (see Section 3.1.2),

radioactive decay data for the naturally occurring isotopes of uranium and thorium and their

radioactive decay products (see Section 3.1.3), and dosimetry data used in estimating dose

from external and internal exposure for the radionuclides of concern in source material (see

Section 3.1.4).

3.1.2 Definition of Source Material

As defined in 10 CFR 40.4 and used in this report, the term source material” means:

(1) Uranium or thorium, or any combination thereof, in any physical or chemical form or

(2) ores which contain by weight one-twentieth of one percent (0.05%) or more of:

(i) Uranium, (ii) thorium, or (iii) any combination thereof. Source material does not

include special nuclear material.

The term special nuclear material” is defined in Section 2.1.2 of this report. In essence, source

material is the raw material from which nuclear fuel is made, and it includes uranium or thorium

in their natural isotopic abundances.

In most of the exempted products or materials containing source material discussed in the

following subsections of Section 3, the uranium or thorium is initially in a chemically separated

form. However, both uranium and thorium decay to short-lived radionuclides that are important

in estimating dose at times shortly after chemical separation of these elements. The short-lived

decay products are included in all dose assessments for chemically separated materials.

In addition, in a few cases (e.g., the exemption for unrefined and unprocessed ore containing

source material discussed in Section 3.3), the exempted materials contain uranium or thorium

that has not been chemically separated. In these cases, all decay products would be present,

including the long-lived decay products of uranium when this is the source material of concern. 

However, the regulatory jurisdiction over the decay products can vary depending on the

particular situation. From the definition of source material given above, any ore containing

0.05% or more by weight of uranium and thorium is source material in its entirety, whereas only

the uranium and thorium are source material in other materials containing uranium or thorium. 

Other naturally occurring radioactive materials that would be present, including the decay

products of uranium and thorium, normally fall under the jurisdiction of the Environmental

Protection Agency and the States. However, particularly when the decay products are present

in connection with an activity involving source material that is licensed by the Nuclear

Regulatory Commission (NRC). Therefore, in the assessments for materials containing less
3–2

than 0.05% by weight of source material in Section 3.2 and for rare earth materials containing

no more than 0.25% by weight of source material in Section 3.9, which are concerned with

nonchemically separated source material other than ore, doses resulting from exposure to the

source material and its radiologically significant decay products are identified separately. 

Information on the total dose from exposure to the source material and its decay products also

is included.

3.1.3 Decay Data for Source Materials

This section presents radioactive decay data that are important in estimating dose from

exposure to uranium or thorium and their radioactive decay products.

First, Table 3.1.1 lists the mass and activity abundances of chemically separated uranium and

thorium. For uranium, these data are given for naturally occurring and depleted uranium,

because both types of uranium are used in exempted products or materials. For naturally

occurring uranium, the activity abundances of the different isotopes are based on their known

mass abundances and half-lives and the assumption that 238U and 234U occur in equal activity

abundances. As defined in 10 CFR 40.4, depleted uranium is any source material uranium in

which the mass abundance of 235U is less than the value for naturally occurring uranium. 

Depleted uranium also has a lower mass abundance of 234U. The abundances of the different

isotopes in depleted uranium are somewhat variable. The values in Table 3.1.1 are intended to

be representative of byproduct residues from uranium enrichment, i.e., the production of

material from natural uranium with a mass abundance of 235U of about 3% for use as special

nuclear material, because this has been the most important source of depleted uranium used in

exempted products or materials. 

Thorium at the time of chemical separation is assumed to consist of equal activity abundances

of 232Th and 228Th, and the 228Th has a negligible mass abundance due to its short half-life. 

Justification for not considering other isotopes of thorium, principally 230Th, that also may be

present in chemically separated thorium is discussed in Section 3.1.5.

Then, radioactive decay data for the naturally occurring isotopes of uranium and thorium and

their radioactive decay products are given in Tables 3.1.2 and 3.1.3. These data include:

The half-life of each isotope of uranium or thorium.

The specific activity of each isotope of uranium or thorium, which is defined as the

activity per unit mass.

All decay products of the isotopes of uranium or thorium that are produced in at least

0.1% of the decays of the parent radionuclide, their half-lives, and the branching fraction

in the decay of the parent radionuclide.

Identification of the principal decay modes for each radionuclide (i.e., alpha, beta, or

beta and gamma).

Table 3.1.2 also identifies the short-lived radioactive decay products of the uranium isotopes

that would be present in significant quantities within a few days or months after chemical
3–3

separation. All other decay products would be present in significant quantities in chemically

separated uranium only after very long times (see Table 3.1.4). For thorium, all radioactive

decay products listed in Table 3.1.3 would be present in significant quantities within a few days

or years after chemical separation (see Table 3.1.5). When uranium and thorium have not

been chemically separated, all decay products presumably would be present and in activity

equilibrium.

Finally, data on the activity of the longer lived radioactive decay products of uranium and

thorium relative to the initial activity of the parent radionuclide as a function of time after

chemical separation are given in Tables 3.1.4 and 3.1.5. Data for the decay chains of uranium

in Table 3.1.4 are shown out to 1 million years only to indicate the very long times required for

the decay products to reach activity equilibrium with the parent radionuclides in chemically

separated natural and depleted uranium. The long-lived decay products of uranium generally

are unimportant in chemically separated uranium at times less than 100 years. The shorter

lived decay products listed in Table 3.1.2 or 3.1.3 that are not shown in Table 3.1.4 or 3.1.5

would always be in activity equilibrium with their immediate longer lived precursors. Decay

chains of the isotopes of uranium and thorium are described in the following paragraphs.

For chemically separated uranium, each naturally occurring isotope has a long-lived decay

product that occurs relatively early in the decay chain (234U in the decay of 238U, 231Pa in the

decay of 235U, and 230Th in the decay of 234U). Therefore, only the activities of the immediate

short-lived decay products identified in Table 3.1.2 (234Th, 234mPa, and 234Pa in the decay of 238U

and 231Th in the decay of 235U) would be important during normal distribution, use, and disposal

of exempted products or materials containing chemically separated uranium. Because of the

very long time periods required for significant buildup of activity for the long-lived decay

products of the naturally occurring isotopes of uranium, as shown in Table 3.1.4, these decay

products and their short-lived decay products generally can be neglected in dose assessments

for exempted products or materials containing chemically separated uranium. The long-lived

decay products of uranium and their short-lived decay products may need to be considered in

dose assessments for exempted products or materials containing uranium that has not been

chemically separated, because all decay products would be in activity equilibrium with the

parent isotopes of uranium. However, as indicated in Section 3.1.2, the regulatory jurisdiction

over the decay products can vary depending on the particular situation.

The decay chains of 232Th and 228Th, which are assumed to occur in equal activities in

chemically separated uranium, include only relatively short-lived radionuclides. Therefore, in

contrast to the situation for uranium described above, the contributions from all decay products

must be taken into account in dose assessments for exempted products or materials containing

thorium, regardless of whether the thorium has been chemically separated. For chemically

separated thorium that initially contains equal activities of 232Th and 228Th, the data in

Table 3.1.5 show that the activity of 228Th and its short-lived decay products decreases with

time for the first few years, due to the decay of the initial activity of 228Th. However, the activity

of 228Th then increases with time after a few years, due to its buildup in the decay of 232Th, until

activity equilibrium in the decay chain is achieved within 50 years.

Dose assessments presented in this report for distribution and use of exempted products or

materials containing thorium presented in this report usually assume that the thorium has been

aged for 20 years after chemical separation. At this time, the data in Table 3.1.5 show that the

activity of all decay products of thorium has achieved about 90% of activity equilibrium. The
3–4

assumption that the thorium has been aged for 20 years provides conservative estimates of

dose for times less than 20 years, and it does not significantly underestimate doses for times

beyond 20 years. For thorium that has not been chemically separated, all decay products

would be in activity equilibrium with the 232Th. Dose assessments for disposal operations also

assume that the thorium has been aged for 20 years, but the assessments of dose after

disposal in landfills assume that the decay products are in activity equilibrium with the 232Th.

3.1.4 Dosimetry Data for Source Materials

Data that can be used to estimate external and internal doses from exposure to uranium or

thorium and their decay products are listed in Tables 3.1.6 and 3.1.7. These data include:

The specific gamma-ray dose constant, which is defined as the dose-equivalent rate per

unit activity at a distance of 1 meter from an unshielded point source in air and which is

a reasonable approximation to the effective dose equivalent (EDE) rate per unit activity

for radionuclides that emit high-energy photons.

The external dose coefficient for submersion in an atmospheric cloud, which is defined

as the external EDE rate per unit concentration in a uniformly contaminated, semiinfinite volume of air. The skin dose component has been included using a 0.01

weighting factor.

The internal dose coefficient for ingestion, which is defined as the 50-year committed

EDE per unit activity intake by ingestion.

The internal dose coefficient for inhalation, which is defined as the 50-year committed

EDE per unit activity intake by inhalation.

The specific gamma-ray dose constant and external dose coefficient for air submersion are

listed in Tables 3.1.6 and 3.1.7 only if a radionuclide emits photons of sufficient energy and

intensity that external exposure could be of concern in assessing dose from exposure to

uranium or thorium and its decay products. Thus, these data are not listed for radionuclides

that are not photon emitters or that emit only very low-energy photons. Similarly, the dose

coefficients for ingestion or inhalation are listed for a decay product only if the dose from

internal exposure would be significant compared with the dose from other radionuclides that

would be present in chemically separated or unseparated materials. As noted in Table 3.1.2,

only the 234Th, 234mPa, and 234Pa decay products of 238U and the 231Th decay product of 235U

would be present in chemically separated uranium. The dosimetry data for each decay product

take into account the branching fraction in the decay of the parent radionuclide given in

Table 3.1.2 or 3.1.3.

For some radionuclides, dose coefficients for ingestion are listed for more than one value of the

gastrointestinal-tract absorption fraction. Dose coefficients for inhalation are listed for more

than one lung clearance class. Absent specific information on the chemical form of uranium or

thorium and their decay products in a particular product or material, the largest dose coefficient

for ingestion or inhalation generally is used in assessing dose. If the chemical form of a

radionuclide in a particular product or material is known, the appropriate dose coefficients for

ingestion and inhalation can be selected based on the assignments given in Table 3.1.8.
3–5

For 222Rn and 220Rn, the dose coefficient for inhalation is the EDE rate from inhalation of radon

and its short-lived decay products per unit concentration of radon in air. In each case, values

are given for both indoor and outdoor exposures. The 219Rn decay product of 235U is sufficiently

short-lived that inhalation exposure generally is not of concern for this radionuclide. The

inhalation dose coefficients for 222Rn and 220Rn assume that their short-lived decay products are

in activity equilibrium in air. Thus, these values generally are conservative for both indoor and

outdoor exposures, because they do not consider the effects of ventilation and deposition. 

Recommended corrections to the dose coefficients for radon to account for the extent of activity

equilibrium of the decay products in air also are discussed in Tables 3.1.6 and 3.1.7.

The dosimetry data in Tables 3.1.6 and 3.1.7 are used, when appropriate, in the dose

assessments for source material in the remainder of Section 3. The dose coefficients for

ingestion and inhalation are used in all assessments of internal exposure. Because the dose

coefficients for ingestion or inhalation are nearly the same for 238U, 235U, and 234U, the dose from

internal exposure to chemically separated uranium does not depend significantly on whether the

uranium is naturally occurring or depleted or on the particular mass abundances of the different

isotopes in depleted uranium. The external dose coefficient for air submersion also is used

whenever this exposure pathway is considered. The external dose coefficients provide

conservative estimates of dose for submersion in a finite atmospheric cloud.

However, the specific gamma-ray dose constant, which provides an indication of the potential

importance of external exposure, is used to estimate external dose only when exposure to an

unshielded point source is an appropriate assumption, i.e., when the dimensions of the source

are small compared with the distance between the source and receptor locations, and the

radionuclides of concern emit photons with energies above about 100 keV. For finite sources

that cannot be represented as a point and for radionuclides that emit only lower energy photons

(e.g., 238U, 234U, and 232Th), external dose rates normally are calculated using the CONDOS-II

(Computer Codes, O’Donnell et al., 1981) or MicroShield (Computer Codes, Grove

Engineering, 1996) computer codes for the appropriate source geometry and amount of

shielding between the source and receptor locations.

In addition, even for small sources that can be represented as a point, the external dose from

238U, 234U, and 232Th generally can be neglected because, for these low-energy photon emitters

(Kocher, 1981), the self-shielding provided by the uranium or thorium in the source would

reduce the dose rate by a large factor compared with the values given in Tables 3.1.6 and 3.1.7

(Unger and Trubey, 1981). Therefore, external exposure to 238U, 234U, and 232Th is well

approximated by assuming that the dose is due only to any photon-emitting decay products that

would be present.

3.1.5 Isotopic Composition of Chemically Separated Thorium

As discussed in Section 3.1.3 and summarized in Table 3.1.1, all assessments in this report

assume that thorium consists of equal activity abundances of 232Th and 228Th at the time of

chemical separation and that no other isotopes of thorium are present. However, thoriumbearing ores also contain varying amounts of uranium (Albert, 1966), and the thorium decay

products of the naturally occurring isotopes of uranium would be present in chemically

separated thorium. The particular isotopes of thorium that would be present, in addition to 232Th
3–6

and 228Th, include 234Th and 230Th produced in the decay of 238U, and 231Th and 227Th produced

in the decay of 235U (see Table 3.1.2).

The amounts of the thorium isotopes other than 232Th and 228Th that would be present in

chemically separated thorium depend on the particular ore. However, based on the following

considerations, the 234Th decay product of 238U and the 231Th and 227Th decay products of 235U

would not be present in significant activities in any thorium that had been aged for a few months

after chemical separation, regardless of their initial activities relative to the activities of 232Th and

228Th. First, the activities of these decay products decrease rapidly with time, because of their

short half-lives (see Table 3.1.2). Second, the external dose from these decay products,

including the contributions from any of their short-lived decay products, is less than the

contribution from 232Th and 228Th with their decay products present. Finally, the internal dose

from ingestion or inhalation of these decay products is much less than the contributions from

232Th and 228Th (see Tables 3.1.6 and 3.1.7).

Therefore, 230Th is the only decay product of the naturally occurring isotopes of uranium of

potential concern in chemically separated thorium. Indeed, there was a report that 230Th

accounted for most of the alpha activity in welding rods containing thorium (see Section 3.6),

although the presence of significant quantities of 230Th in chemically separated thorium has not

been noted in any other studies.

Available information on the most commonly used sources of thorium indicates that 230Th

should not be radiologically significant in chemically separated thorium compared with 232Th and

228Th. Information supporting this conclusion is summarized below.

Most of the thorium used commercially has been extracted from the mineral monazite (Hedrick,

1991; NCRP 118). Other ores containing greater amounts of uranium have not been significant

sources of thorium. The amount of uranium and, thus, 230Th in monazite is somewhat variable

(NCRP 118), but Albert (1966) reported that monazite typically contains an activity of 230Th

equal to 11% of the activity of 232Th and 228Th. Based on this activity, the contribution to dose

from 230Th relative to the contribution from 232Th and 228Th and their decay products can be

estimated as follows.

Because 230Th emits only low intensities of low-energy photons (Kocher, 1981), this isotope

would not contribute significantly to external dose for chemically separated thorium, especially

when the rapid buildup of the photon-emitting decay products of 228Th is considered (see

Tables 3.1.5 and 3.1.7). For ingestion, the dose from 230Th would be only about 2% of the total

dose from all thorium isotopes in freshly separated materials, based on the dose coefficients

given in Tables 3.1.6 and 3.1.7. The contribution would decrease to about 1.4% in aged

materials in which the decay products of 232Th and 228Th are in activity equilibrium. For

inhalation of either Class W or Class Y materials, the dose from 230Th also would be only about

2% of the total dose in freshly separated or aged materials. Thus, the contributions to internal

dose from 230Th in monazite would be insignificant.

The isotope 230Th decays to 226Ra and its short-lived decay products, and these decay products

are radiologically more significant than the parent isotope (see Table 3.1.6). However, the

decay products of 230Th do not occur in significant quantities in chemically separated thorium for

hundreds of years or more (see Table 3.1.4). For example, the contribution to internal dose

from 230Th would approach 10%, which is still a relatively insignificant amount, for those times
3–7

after 1000 years following chemical separation. Therefore, the buildup of 226Ra and its decay

products would not be of concern during the normal life cycle of exempted products or materials

containing thorium extracted from monazite, or for long times after disposal.

Based on the analysis described above, the possible presence of 230Th in chemically separated

thorium is not considered in the assessments of exempted products or materials containing

source material in the following subsections of Section 3.
3–8

Table 3.1.1 Mass and Activity Abundances of Naturally Occurring Isotopes in Chemically 

Separated Uranium and Thorium

Element Isotope

Half-lifea

(yr)

Mass

Abundance

Activity

Abundance

Uranium (natural) 238U 4.468×109 99.2745%b 48.83%c

235U 7.038×108 0.720% 2.34%

234U 2.445×105 0.0055% 48.83%

Uranium (depleted) 238U 4.468×109 99.75%d 90.1%e

235U 7.038×108 0.25% 1.5%

234U 2.445×105 0.0005% 8.4%

Thorium 232Th 1.405×1010 100% 50%f

228Th 1.9132 50%

a

 Values from Kocher (1981).

b

 Mass abundances for isotopes in natural uranium from Parrington et al. (1996).

c

 Activity abundances for isotopes in natural uranium are based on known mass abundances

and half-lives and assumption that 238U and 234U occur in equal activity abundances.

d

 Mass abundances for isotopes in depleted uranium obtained as byproduct residues from

uranium enrichment reported by Rich et al. (1988).

e

 Activity abundances for isotopes in depleted uranium are based on assumed mass

abundances and known half-lives.

f

 Activity abundances for isotopes in natural thorium are based on assumption of equal activity

abundances of 232Th and 228Th at time of chemical separation and assumption that no other

isotopes of thorium produced in decay of naturally occurring isotopes of uranium are present

(see Section 3.1.5).
3–9

Table 3.1.2 Decay Data for Naturally Occurring Isotopes of Uranium 

and Their Radioactive Decay Products

Isotope

Decay

Product Half-Life a

Specific

Activity b

(curie(Ci)/g)c

Branching

Fractiond

Principal

Decay Modes

238U 4.468×109

 yr 3.35×107 Alpha

234The 24.10 day 1.0 Beta/gamma

234mPae 1.17 min 1.0 Beta/gamma

234Pae 6.70 h 0.0016 Beta/gamma

234U 2.445×105

 yr 1.0 Alpha

230Th 7.7×104

 yr 1.0 Alpha

226Ra 1,600 yr 1.0 Alpha

222Rn 3.8235 day 1.0 Alpha

218Po 3.05 min 1.0 Alpha

214Pb 26.8 min 0.9998 Beta/gamma

214Bi 19.9 min 1.0 Beta/gamma

214Po 163.7 s 0.99979 Alpha

210Pb 22.26 yr 1.0 Beta

210Bi 5.013 day 1.0 Beta

210Po 138.378 day 1.0 Alpha

See end of table for footnotes.
3–10

Table 3.1.2 Decay Data for Naturally Occurring Isotopes of Uranium 

and Their Radioactive Decay Products (continued)

Isotope

Decay

Product Half-Lifea

Specific

Activityb

(Ci/g)c

Branching

Fractiond

Principal

Decay Modes

235U 7.038×108

 yr 2.16×106 Alpha/gamma

231The 25.52 h 1.0 Beta/gamma

231Pa 3.276×104

 yr 1.0 Alpha/gamma

227Ac 21.773 yr 1.0 Beta/alpha

227Th 18.718 day 0.9862 Alpha/gamma

223Fr 21.8 min 0.0138 Beta/gamma

223Ra 11.434 day 1.0 Alpha/gamma

219Rn 3.96 s 1.0 Alpha/gamma

215Po 1.778 ms 1.0 Alpha

211Pb 36.1 min 1.0 Beta/gamma

211Bi 2.13 min 1.0 Alpha/gamma

211Po 0.516 s 0.00273 Alpha/gamma

207Tl 4.77 min 0.99727 Beta/gamma

234U f g

 2.445×105 yr 6.24×103 Alpha

a

 Values from Kocher (1981).

b

 Specific activity of isotope, defined as activity per unit mass.

c

 1 Ci/g = 0.037 TBq/g.

d

 Number of atoms of decay product per decay of parent radionuclide.

e

 Short-lived radioactive decay product that would be present in significant quantities within

short time after chemical separation of uranium.

f

 Entries for radioactive decay products are given following entry for 234U decay product of 238U.

g

 No decay products would be present in significant quantities for many years after chemical

separation of uranium.
3–11

Table 3.1.3 Decay Data for Naturally Occurring Isotopes of Thorium 

and Their Radioactive Decay Products

Isotope

Decay

Product Half-Lifea

Specific

Activityb

(Ci/g)c

Branching

Fractiond

Principal

Decay Modes

232Th e 1.405×1010 yr 1.09×107 Alpha

228Ra 5.75 yr 1.0 Beta

228Ac 6.13 h 1.0 Beta/gamma

228Th 1.9132 yr 1.0 Alpha

224Ra 3.62 day 1.0 Alpha/gamma

220Rn 55.61 s 1.0 Alpha

216Po 0.146 s 1.0 Alpha

212Pb 10.643 h 1.0 Beta/gamma

212Bi 60.55 min 1.0 Beta/gamma

212Po 0.298 s 0.6407 Alpha

208Tl 3.053 min 0.3593 Beta/gamma

228Thf e 1.9132 yr 8.19×102 Alpha

a

 Values from Kocher (1981).

b

 Specific activity of isotope, defined as activity per unit mass.

c

 1 Ci/g = 0.037 TBq/g.

d

 Number of atoms of decay product per decay of parent radionuclide.

e

 All radioactive decay products would be present in significant quantities within short time after

chemical separation of thorium.

f

 Entries for radioactive decay products are given following entry for 228Th decay product of

232Th.
3–12

Table 3.1.4 Activity of Radioactive Decay Products of Uranium vs. Time

After Chemical Separation of Uranium a

Time After Chemical Separation of Uranium

(yr)

Radionuclideb 102 103 104 105 106

238U chain

 238U 1.0 1.0 1.0 1.0 1.0

 234U 2.8×104 2.8×103 2.8×102 0.25 0.94

 230Th 1.3×107 1.3×105 1.2×103 8.8×102 0.91

 226Ra 1.8×109 1.7×106 7.9×104 8.4×102 0.91

 210Pb 8.5×1010 1.5×106 7.9×104 8.4×102 0.91

235U chain

 235U 1.0 1.0 1.0 1.0 1.0

 231Pa 2.1×103 2.1×102 0.19 0.88 1.0

 227Ac 1.5×103 2.0×102 0.19 0.88 1.0

234U chain

 234U 1.0 1.0 0.97 0.75 5.9×102

 230Th 9.0×104 9.0×103 8.5×102 0.51 8.6×102

 226Ra 1.9×105 1.7×103 6.6×102 0.50 8.6×102

 210Pb 1.1×105 1.6×103 6.6×102 0.50 8.6×102

238U + 234Uc 1.0 1.0 1.0 1.0 1.0

 238U 0.50 0.50 0.50 0.50 0.50

 234U 0.50 0.50 0.50 0.50 0.50

 230Th 4.5×104 4.5×103 4.3×102 0.30 0.50

 226Ra 9.6×106 8.5×104 3.3×102 0.29 0.50

 210Pb 5.4×106 8.0×104 3.3×102 0.29 0.50

a

 Activities of each radionuclide are normalized to initial activity of unity for parent radionuclide

or isotopes of uranium at time of chemical separation.

b

 At any time after chemical separation of uranium, all other shorter lived decay products listed

in Table 3.1.2 would be in activity equilibrium with their immediate longer lived precursors,

taking into account the decay branching fraction.

c

 Activities of the two uranium isotopes are assumed to be equal at time of chemical separation,

as would be the case for naturally occurring uranium.
3–13

Table 3.1.5 Activity of Radioactive Decay Products of Thorium vs. Time

After Chemical Separation of Thorium a

Time After Chemical Separation of Thorium

(yr)

Radionuclideb 0.5 1 5 10 20

232Th chain

 232Th 1.0 1.0 1.0 1.0 1.0

 228Ra 5.8×102 0.11 0.45 0.70 0.91

 228Th 5.0×103 1.9×102 0.26 0.56 0.87

 224Ra 4.8×103 1.8×102 0.26 0.56 0.87

228Th chain

 228Th 0.83 0.70 0.16 2.7×102 7.1×104

 224Ra 0.84 0.70 0.16 2.7×102 7.1×104

232Th + 228Thc 0.92 0.86 0.71 0.80 0.93

 232Th 0.50 0.50 0.50 0.50 0.50

 228Ra 2.9×102 5.7×102 0.23 0.35 0.46

 228Th 0.42 0.36 0.21 0.30 0.43

 224Ra 0.42 0.36 0.21 0.30 0.43

a

 Activities of each radionuclide are normalized to initial activity of unity for parent radionuclide

or isotopes of thorium at time of chemical separation.

b

 At any time after chemical separation of thorium, all other shorter lived decay products listed in

Table 3.1.3 would be in activity equilibrium with their immediate longer lived precursors, taking

into account decay branching fraction.

c

 Activities of the two thorium isotopes are assumed to be equal at time of chemical separation,

as would be the case for naturally occurring thorium.
3–14

Table 3.1.6 Dosimetry Data for Naturally Occurring Isotopes of Uranium

and Their Radioactive Decay Products

Dose Coefficient

Radionuclidea

Γb

(rem/h per Ci)

Air Submersionc

(rem-m3/Ci-yr)

Ingestiond, e

(rem/Ci)

Inhalationd, f

(rem/Ci)

238U 6.5×108 4.3×104 2.6×101

 (0.05)

2.4×102

 (0.002)

2.5 D

7.0 W

1.2×102

 Y

 234Thg 7.5×108 4.0×102 1.4×102

 234mPag 1.0×108 1.5×101

 234Pag 3.2×109 1.8×102

 234U 2.8×101 (0.05)

2.6×101

 (0.002)

2.7 D

7.9 W

1.3×102

 Y

 230Th 5.5×101 3.3×102

 W

2.6×102

 Y

 226Ra 1.2×108 3.7×102 1.3 8.6 W

 222Rn 3.2×102 h

 214Pb 3.2×107 1.4

 214Bi 8.4×107 9.1

 210Pb 5.4 1.4×101

 D

 210Po 1.9 9.4 D

8.6 W

See end of table for footnotes.
3–15

Table 3.1.6 Dosimetry Data for Naturally Occurring Isotopes of Uranium

and Their Radioactive Decay Products (continued)

Dose Coefficient

Radionuclidea

Γb

(rem/h per Ci)

Air Submersionc

(rem-m3/Ci-yr)

Ingestiond, e

(rem/Ci)

Inhalationd, f

(rem/Ci)

 235U 3.4×107 8.5×101 2.7×101

 (0.05)

2.7×102

 (0.002)

2.5 D

7.3 W

1.2×102

 Y

 231Thg 5.5×107 6.4×102

 231Pa 3.7×107 2.0×101 1.1×101 1.3×103 W

8.6×102

 Y

 227Ac 1.4×101 6.7×103

 D

1.7×103

 W

1.3×103

 Y

 227Th 4.2×107 5.7×101

 223Ra 3.3×107 7.2×101 6.6×101

 219Rn 5.3×108 3.2×101

 211Pb 3.6×108 3.3×101

 211Bi 4.7×108 2.6×101

 207Tl 1.3×109 5.5×102

 234Ui 7.8×108 9.4×104 2.8×101

 (0.05)

2.6×102

 (0.002)

2.7 D

7.9 W

1.3×102

 Y

See following page for footnotes.
3–16

Footnotes to Table 3.1.6

a

 Indented entries are radioactive decay products of parent uranium isotopes. Dosimetry data

for decay products are listed only if they are significant compared with data for parent uranium

isotope or other precursor radionuclides for uranium that is not chemically separated. 

Dosimetry data for decay products take into account branching fraction in decay of parent

radionuclide given in Table 3.1.2.

b

 Specific gamma-ray dose constant obtained from Unger and Trubey (1981) gives doseequivalent rate per unit activity at a distance of 1 meter from an unshielded point source in air. 

1 rem/h per Ci = 270 millisieverts (mSv)/h per MBq. c

 Values for external exposure from submersion in uniformly contaminated, semi-infinite

atmospheric cloud obtained from EPA 402–R–93–081. The skin dose component has been

included using a 0.01 weighting factor. 1 rem-m3

/Ci-yr = 8.57×1015 Sv-m3/Bq-s.

d

 Values for internal exposure obtained from EPA–520/1–88–020. 1 rem/Ci = 270 mSv/MBq. e

 If more than one value is given, entry in parentheses is corresponding gastrointestinal-tract

absorption fraction. Assigned absorption fraction for different chemical forms of element is

given in Table 3.1.8.

f

 Assumed lung clearance class is denoted by D for days, W for weeks, or Y for years. 

Assigned clearance class for different chemical forms of element is given in Table 3.1.8.

g

 Short-lived decay product that would be present in chemically separated uranium; all other

decay products would be present only in uranium that is not chemically separated.

h

 Value obtained from ICRP 50 (see references) gives effective dose-equivalent rate from

inhalation of 222Rn and its short-lived decay products per unit concentration of 222Rn in indoor

air, in units of rem-m3

/Ci-yr, assuming activity equilibrium in air between radon and its decay

products. For 222Rn in outdoor air, value is increased by factor of 1.4. Representative mean

values of activity equilibrium factor for 222Rn decay products in indoor and outdoor air are 0.45

and 0.7, respectively (ICRP 50). 1 rem-m3

/Ci-yr = 8.57×1015 Sv-m3/Bq-s.

i

 Entries for radioactive decay products are given following entry for 234U decay product of 238U.
3–17

Table 3.1.7 Dosimetry Data for Naturally Occurring Isotopes of Thorium

and Their Radioactive Decay Products

Dose Coefficient

Radionuclidea

Γb

(rem/h per Ci)

Air Submersionc

(rem-m3/Ci-yr)

Ingestiond

(rem/Ci)

Inhalationd, e

(rem/Ci)

 232Th 6.8×108 1.1×103 2.7 1.6×103

 W

1.2×103

 Y

 228Ra 1.4

 228Ac 8.4×107 5.7

 228Th 7.9×108 1.1×102 4.0×101 2.5×102

 W

3.4×102

 Y

 224Ra 1.1×108 5.6×102 3.7×101

 220Rn 1.6×103 f

 

 212Pb

2.7×107 8.2×101

 212Bi 1.9×107 1.1

 208Tl 6.1×107 7.6

 228Thg 7.9×108 1.1×102 4.0×101 2.5×102

 W

3.4×102

 Y

a

 Only 232Th and 228Th are assumed to be present in significant amounts in naturally occurring

thorium (see Section 3.1.5). Indented entries are decay radioactive products of parent thorium

isotopes. All decay products would be present in chemically separated thorium within a short

time after separation (see Table 3.1.5), but dosimetry data for decay products are listed only if

they are significant compared with data for parent thorium isotope. Dosimetry data for decay

products take into account branching fraction in decay of parent radionuclides given in

Table 3.1.2.

b

 Specific gamma-ray dose constant obtained from Unger and Trubey (1981) gives doseequivalent rate per unit activity at a distance of 1 meter from an unshielded point source in air. 

1 rem/h per Ci = 270 mSv/h per MBq. c

 Values for external exposure from submersion in uniformly contaminated, semi-infinite

atmospheric cloud obtained from EPA 402–R–93–081. The skin dose component has been

included using a 0.01 weighting factor. 1 rem-m3

/Ci-yr = 8.57×1015 Sv-m3/Bq-s.

d

 Values for internal exposure obtained from EPA–520/1–88–020. 1 rem/Ci = 270 mSv/MBq. e

 Assumed lung clearance class is denoted by D for days, W for weeks, or Y for years. 

Assigned clearance class for different chemical forms of element is given in Table 3.1.8.

f

 Value obtained from ICRP 50 (see references) gives effective dose-equivalent rate from

inhalation of 220Rn and its short-lived decay products per unit concentration of 220Rn in indoor or

outdoor air, in units of rem-m3

/Ci-yr, assuming activity equilibrium in air between radon and its

decay products. Theoretical estimates of activity equilibrium factor for 220Rn decay products in

indoor air are 0.02 to 0.1 (ICRP 50). 1 rem-m3

/Ci-yr = 8.57×1015 Sv-m3/Bq-s.

g

 Entries for radioactive decay products are given following entry for 228Th decay product of

232Th.
3–18

Table 3.1.8 Gastrointestinal-Tract Absorption Fractions (f1) and Lung Clearance Classes

for Chemical Compounds of Uranium, Thorium, and Their Decay Products a

Ingestion Inhalation

Element Compound f1 Compound f1/Class

U (Uranium) Hexavalent

Insoluble forms

 0.05

 0.002

UF6, UO2F2, and

 UO2(NO3)2

UO3, UF4, and UCl4

UO2 and U3O8

0.05 D

 0.05 W

 0.002 Y

Th (Thorium) All forms 2×104 Oxides and hydroxides

All others

2×104 Y

 2×104

 W

Pa (Protactinium) All forms 1×103 Oxides and hydroxides

All others

 1×103 Y

 1×103

 W

Ac (Actinium) All forms 1×103 Oxides and hydroxides

Halides and nitrates

All others

 1×103

 Y

 1×103

 W

 1×103

 D

Au (Gold) All forms 0.1 Oxides and hydroxides

Halides and nitrates

All others

 0.1 Y

 0.1 W

 0.1 D

Po (Polonium) All forms 0.1 Oxides, hydroxides and

 nitrates

All others

 0.1 W

 0.1 D

a

 Assignments of gastrointestinal-tract absorption fractions and lung clearance classes obtained

from EPA–520/1–88–020. For elements not listed in this table, dose coefficients for ingestion

and inhalation in Tables 3.1.6 and 3.1.7 apply to all chemical forms.
3–19

3.2 Chemical Mixture, Compound, Solution, Alloy Containing

<0.05 Percent by Weight of Source Material

3.2.1 Introduction

In 10 CFR 40.13(a), any person is exempt from the requirements for a license, to the extent the

such person receives, possesses, uses, transfers, or delivers source material in any chemical

mixture, compound, solution, or alloy in which the source material is by weight less than 0.05%

of the mixture, compound, solution, or alloy. This exemption was proposed on September 7,

1960 (25 FR 8619), and issued as a final rule on January 14, 1961 (26 FR 284). 

However, previously the definition of source material in 10 CFR Part 40 excluded all uranium

and thorium in materials less than 0.05% by weight of uranium and thorium. In the original

10 CFR Part 40, published in 1947, source material was defined as any material, except

fissionable material, which contains by weight one-twentieth of 1% (0.05%) or more of

(1) uranium, (2) thorium, or (3) any combination thereof.” The rulemaking addressed in the

aforementioned Federal Register notices constituted an overall revision of 10 CFR Part 40. In

that rulemaking, the definition of source material was changed to be consistent with the

definition in the Atomic Energy Act of 1954 to its current wording: Source material means

(1) Uranium or thorium, or any combination thereof, in any physical or chemical form or (2) ores

which contain by weight one-twentieth of 1% (0.05%) or more of (i) uranium, (ii) thorium or

(iii) any combination thereof. Source material does not include special nuclear material.” 

Because of the change in the definition of source material, the exemption for materials other

than ore that are less than 0.05% of source material was added, with the net effect of no

change to the licensing program. The record does not reflect a full consideration of the health

and safety significance of this concentration of source material. Applying the same

concentration limit, as used for ore in the definition of source material, to source material in

other forms was administratively convenient.

In the Atomic Energy Act of 1946, source material was defined as uranium, thorium, or any

other material which is determined by the Atomic Energy Commission (AEC), with the approval

of the President, to be peculiarly essential to the production of fissionable materials; but

includes ores only if they contain one or more of the foregoing materials in such concentrations

as the AEC may by regulation determine from time to time.” In the Atomic Energy Act of 1954,

the definition was modified somewhat, but not substantively. Also in the 1946 Act, Section

5(b)(2), it was stated that licenses shall not be required for quantities of source materials that, in

the opinion of the AEC, are unimportant. Section 62 of the Atomic Energy Act of 1954 contains

the same statement. It appears that the Commission selected the 0.05% value primarily on the

basis of that concentration of source material that was considered strategically important for the

production of special nuclear material.

The exemption refers only to the weight percent of uranium and thorium, 0.05%, which equates

to 500 ppm. To provide a comparison of weight percent to activity concentrations for uranium

and thorium, two major types of materials are considered--natural and processed. Natural

uranium consists of 238U (99.27% by mass), 235U (0.72% by mass), and 234U (0.0054% by

mass). Natural and processed thorium consists of 232Th (nearly 100% by mass and 50% by

activity) and 228Th (50% by activity). Depleted uranium is an example of chemically processed

uranium; it differs from natural uranium by containing less 235U (0.2 to 0.25% by mass) and 234U
3–20

(0.0005% by mass). Table 3.2.1 summarizes the radionuclide concentrations associated with

0.05% weight of uranium and thorium. Refer to Section 3.1 for a more detailed discussion on

the properties of source material.

3.2.2 Description of Materials

Since uranium and thorium are ubiquitous in the environment, there is a vast range of products

and materials that may contain low concentrations of source material and thus be covered by

this exemption. Only a few of these involve purposeful introduction of the source material. 

Uranium and thorium exist naturally in soil, and trace quantities appear in many products and

materials, including the human body. Based on a literature review and discussions with

knowledgeable individuals, it appears that some of the specific materials and products covered

by this exemption that may have significant concentrations of source material are:

dental products containing uranium;

ophthalmic glass; 

mineral-derived products from minerals such as zirconium, hafnium, and titanium and

large-volume metals, such as copper and iron, coal, and phosphates; and 

waste-derived products from materials such as coal ash, phosphate slag, and water

treatment sludge.

The following subsections contain brief descriptions of individual products, such as dental

products containing uranium and ophthalmic glass, and categories of products, such as

mineral-derived and waste-derived products, which are covered by this exemption.

3.2.2.1 Dental Products

The practice of adding uranium to dental ceramics in the United States began during the early

1900s and continued until the early 1980s. It was discovered that a small amount of uranium

contributed a natural color and fluorescence to dentures. Restoration of natural appearance is

one of the major reasons for using prostheses. Other substances have been found to imitate

these characteristics over a broad range of daylight and artificial lighting conditions. American

manufacturers have used depleted uranium in their porcelain dental products, whereas, at least

in certain Japanese products, natural uranium has been used (Phone call, D. L. Thompson,

Center for Devices and Radiological Health, U.S. Food and Drug Administration, May 1995).

In 1979, the American Dental Association (ADA) developed Specification No. 52, which

provided standards for the uranium content in dental porcelain and porcelain teeth (ANSI/ADA

Spec. No. 52). This standard established a 40% reduction in the permissible amount of

uranium that could be used in porcelain teeth (0.03% by weight). Assuming depleted uranium

is used, 0.03% by weight is equivalent to about 3.7 becquerels (Bq)/g (100 picocuries (pCi)/g). 

However, the fluorescent behavior of dental ceramics was still not identical to natural teeth in all

cases and the use of uranium in ceramics was considered to be undesirable. Dental porcelain

manufacturers, while adhering to this standard, were also looking for alternative materials to

replace uranium. In the early 1980s, the last of the U.S. major manufacturers had phased out
3–21

uranium in dental ceramics (Phone call, D. L. Thompson, Center for Devices and Radiological

Health, U.S. Food and Drug Administration, May 1995). According to ANSI/ADA Specification

No. 69, which became effective May 1992, the manufacturer must submit a certification of

assurance verifying that fluorescing agents that will increase the radioactivity of the ceramic

have not been added (ANSI/ADA, Spec. No. 69). In addition, Specification No. 52 had been

withdrawn (ANSI/ADA, Spec. No. 69).

Full dentures have an average life of only 5 to 11 years since the oral cavity may change over

time (Mazurat, 1992). The predicted service life for removable partial dentures is 8 to 10 years

(Mazurat, 1992). For denture wearers 65 years and over, the age of dentures worn is typically

10 years or more. In one study it was found that 20% of denture wearers had worn dentures for

20 years (Mazurat, 1992). Typically, old dentures are either thrown into the trash or retained in

the home (Phone call, A. Kayes, President of the Prosthodontics Society, May 1995).

Crowns and inlays are considered permanent dentifrices since they are replaced infrequently. 

The predicted service life for fixed restorations is 8 to 11 years, though there are instances in

which these restorations may be left in place for greater periods of time. Therefore, though

uranium has not been used in dental products since the early 1980s, some individuals may still

have these crowns.

3.2.2.2 Ophthalmic Lenses

The manufacture of ophthalmic glass, which is glass used in eyeglasses or sunglasses,

frequently used mixtures of rare earths and zirconium oxides, which contain natural uranium

and thorium. These lenses should not be confused with intentionally thoriated glass lenses

used in special instruments. Refer to Section 3.19 for information on thorium in finished optical

lenses.

In 1975, the Optical Manufacturers Association (OMA) voluntarily issued an ophthalmic glass

radiological standard (OMA, 1975). The purpose of this standard was to establish a uniform

maximum level for radioactive emissions from ophthalmic glass. Use of the standard is

voluntary. However, it appears that U.S. ophthalmic glass manufacturers comply with this

standard (Phone call, W. Price, Product Assurance, Corning, Inc., Corning, NY, May 1995).

The use of glass lenses in eyeglasses has decreased considerably over the years as plastic

lenses have become more popular. It is estimated that glass has only 5-8% of the U.S.

eyeglass market (Optical Advisor (OA), 1999). Based on this estimate, there may be about

12 million glass-containing eyeglass wearers (Vision Council of America (VCA), 1999). The

average lifetime of a pair of eyeglasses ranges from 2 to 4 years (Phone call, T. Loomis,

Manager, Product Assurance, Optical Products Department, Corning Inc., Corning, NY,

May 1995; phone call, W. Price, Product Assurance, Corning, Inc., Corning, NY, May 1995). 

Typically, eyeglasses are stored in the home or are disposed in a sanitary landfill and could go

to municipal incinerators. Eyeglasses, both frames and lenses, have been redistributed to

individuals in the United States and other parts of the world by service organizations such as

the Lions Club (Phone call, W. Price, Product Assurance, Corning, Inc., Corning, NY,

May 1995).
3–22

3.2.2.3 Mineral-Derived Products

Mineral-derived products can encompass a wide range of minerals that contain naturally

occurring radioactive material, including uranium and thorium. Table 3.2.2 summarizes the

uranium and thorium content found in selected mineral resources, products, and process

wastes. Table 3.2.3 lists a few common building materials and associated uranium and thorium

concentrations. As one may note, most of these products typically contain less than 0.05% by

weight of uranium or thorium.

Products that are derived from minerals are numerous. Mineral-derived products range from

mineral concentrates, which may only be physically processed, to products that have been

highly refined, e.g., chemically processed and purified. These mineral-derived products may

then be used in other consumer products. Examples of mineral-derived intermediate and final

products are shown in Table 3.2.4.

Mineral extraction may include a number of processes, as illustrated in Table 3.2.5. Some of

the processes may carry uranium and thorium into the product, or it may concentrate within the

product or the waste. How radionuclides are incorporated into products and wastes are

determined by a combination of factors, including uranium and thorium levels in ore, process

chemistry, process temperature, and process and collection efficiency (Hendricks, 1987). The

heavy mineral extraction industry provides an excellent example of how thorium and uranium

can concentrate in wastes based on processing method, as shown in Figure 3.2.1.

Five major categories of minerals from which mineral-derived products result are: (1) rare

earths; (2) the special application metals, such as zirconium (Zr), hafnium (Hf), and titanium

(Ti); (3) the large volume-metals, such as copper and iron; (4) coal; and (5) phosphates. The

rare earth industry is not described in this section since it is covered by a separate exemption in

10 CFR 40.13(c)(1)(vi); an assessment of that exemption is contained in Section 3.9. Heavy

minerals, such as zirconium and titanium, and phosphates appear to typically exhibit elevated

concentrations as compared to the other mineral resources. Therefore, the radiological impacts

associated with these mineral resources are highlighted in this assessment.

3.2.2.4 Waste-Derived Products

3.2.2.4.1 Coal Ash

Coal contains naturally occurring uranium and thorium as well as their radioactive decay

products. The radioactivity of coal is known to vary more than two orders of magnitude

depending upon the type of coal and the region from which it has been mined. Utility and

industrial boilers are estimated to generate 61 million metric tons of coal ash per year

(Environmental Protection Agency (EPA) RAE–9232/1–2). Coal ash consists of about 74% fly

ash, 20% bottom ash, and about 6% boiler slag. Uranium and its progeny tend to stay with the

ash when the coal is burned. Concentrations of uranium in coal and ash vary widely, but some

indications of mean values are given in Table 3.2.6.

Of the total amount of coal ash generated, nearly 20 million metric tons are used in a variety of

applications instead of being sent to disposal facilities. Coal ash is primarily being used as an

additive in concrete, as a structural fill, and for road construction. Fly ash has been used to

replace cement in concrete in the United States since 1910. Typically, fly ash is substituted for
3–23

10 to 30% of the cement (EPA, RAE–9232/1–2). Since the early 1970s, all three types of coal

ash have been used in construction projects, including industrial parks, housing developments,

roadbed embankments, and soil stabilization (EPA, RAE–9232/1–2).

3.2.2.4.2 Phosphate Slag

Uranium in phosphate ores in the United States ranges in concentration from 20 to 200 ppm,

while thorium occurs at ambient background concentrations of between 1 to 5 ppm 

(EPA 402–R–92–002). Phosphate slag is the principal waste byproduct generated from the

production of elemental phosphorus. Phosphogypsum is the principal waste byproduct

generated during the phosphoric acid production process. Some of the impurities contained in

the phosphate slag and phosphogypsum include uranium and thorium and their radioactive

decay products, which become concentrated in the waste. Table 3.2.7 summarizes typical

radionuclide concentrations found in phosphate ore, phosphate slag, and phosphogypsum. As

noted, uranium and thorium concentrations are higher in phosphate slag as compared to

phosphogypsum. Because of the high temperatures involved in the thermal process, slag is a

vitrified material that binds nonvolatile radionuclides.

Phosphate slag wastes have been used in a number of different applications (EPA,

RAE–9232/1–2). Phosphate slag is used as an aggregate in asphalt manufacturing. 

Phosphate slag has been used extensively in highway construction for many years in Idaho,

Montana, and Tennessee. The material is used as crushed base and crushed aggregate for

asphalt and in casting material for highway structures. It has also been used as an aggregate

for Portland cement and concrete in making construction blocks, driveways, patios, and

drainage ditches. Phosphate slag is used as railroad ballast and as stabilization material for

stockyards. In Florida, where use of phosphate slag in habitable structures has not been

prohibited, slag has been used on roofing shingles, in septic tank fields, and in manufacture of

rockwool insulation.

3.2.2.4.3 Water Treatment Sludge

Water supply systems generate an estimated yearly total of 3.1 million metric tons of waste,

including sludge and other waste forms (EPA, RAE–9232/1–2). Of this, it is estimated that 700

water utilities generate 300,000 metric tons of naturally occurring radioactive material waste

yearly, including sludge, spent resin, and charcoal beds. Most of this waste is disposed in

landfills and lagoons or applied to agricultural fields (EPA, RAE–9232/1–2). 

Federal and State regulations allow radioactive material to be discharged into sanitary sewage

systems with specified concentrations, providing the material is readily soluble or dispersible

biological material. In many cases, naturally occurring radioactive materials are not readily

soluble and may accumulate in the sewer lines. Likely contributors of radionuclides that may

reconcentrate in the sanitary sewage systems are the users of zircon sand, zircon flour, and

thorium oxide (Conference of Radiation Control Program Directors (CRCPD), 1994). These

users include foundries, casting facilities, producers of zirconium or refractories, and ceramic

and welding rod manufacturers (CRCPD, 1994). In addition, naturally occurring radioactive

material, including uranium and thorium, has been found to reconcentrate in sewage sludge. 

Therefore, it poses potential problems with the use and distribution of sludge and its

byproducts, such as fertilizer and compost.
3–24

3.2.2.4.4 Oil and Gas Production and Process Waste

The presence of naturally occurring radioactive materials in oil and gas production and

processing wastes has been recognized since the 1930s (Smith et al., 1995). The primary

radionuclides of concern are 226Ra and 228Ra. The uranium and thorium radionuclides, present

in the subsurface formation, are largely immobile and remain in the subsurface (Smith et al.,

1995). The more soluble radium can become mobilized in the formation water and be

transported to the surface in the produced water stream. Since uranium and thorium are not

the radionuclides of concern, the radiological impacts associated with oil and gas production

and process waste are not evaluated in this assessment.

3.2.3 Summary of Previous Analyses and Assessments

When this exemption was proposed and finalized, there was no accompanying information on

analyses of radiological impacts on members of the public from the use of the materials. The

proposed rule states only that possession and use of these materials can be conducted

without any unreasonable hazard to life,” but the meaning of unreasonable hazard” is not

quantified.

In previous analyses for this exemption, dose assessments were performed only for consumer

products distributed for use under this exemption: specifically, dental products (e.g., dentures

and crowns) containing uranium and for ophthalmic lenses (e.g., eyeglasses) containing natural

thorium. For the current assessment, information was also compiled on mineral-derived

products, with a particular emphasis on zircon-derived products, and on waste-derived

products, with an emphasis on coal ash, phosphate slag, and sewer sludge. A general

description of the aforementioned products follows.

3.2.3.1 Dental Products

Summarized in Table 3.2.8 are the results from a number of studies that evaluated the uranium

concentrations in dental products and potential doses to denture and crown wearers. Buckley

et al. (NUREG/CR–1775) conducted a comprehensive dose assessment associated with the

manufacture, distribution, and use of dental products containing uranium. A brief description of

the dental product dose assessment done by Buckley et al. (NUREG/CR–1775) is provided

below. 

Doses from routine use of dental products containing 0.05% by weight of uranium were

estimated by Buckley et al. (NUREG/CR–1775). The routine exposure scenarios for exempt

uses of these products included exposures to distributors and workers in dental laboratories

and dental offices and exposures to wearers of dentures and crowns. The exposure scenarios

involving accidents or misuse of dental products include exposures to firefighters from uranium

released in a warehouse fire, and exposures to dental technicians from uranium dust generated

during tooth construction. Doses were not estimated for disposal of dental products, since few

false teeth would be placed in landfills and there should be little release of uranium from the

porcelain material. 

For wearers of dental restorations, the gums, inner cheeks, tongue, and lips may be exposed to

alpha, beta, and gamma radiations. Dose estimates made by Buckley et al.
3–25

(NUREG/CR–1775), from alpha and beta particles from teeth containing 0.05% by weight of

uranium are summarized below:

Annual dose equivalents from alpha particles at the surface of the teeth range from 1 to

about 4 sieverts (Sv) (100 to 400 rem). However, since the maximum range of the

alpha particles in tissue is 30 m, essentially the entire dose to tissues of the mouth

would be received by superficial cells on the surface and the dose to radiosensitive cells

in the basal layer of the tissues would be zero (0).

The annual dose equivalent to the basal layer of tissues from beta particles would be

about 14 mSv (1.4 rem).

The dose estimates for distribution workers and denture wearers from gamma irradiation are

summarized as follows: 

The annual whole-body dose equivalent to workers in distribution facilities from photon

irradiation is about 3×105

 mSv (0.003 mrem).

For wearers of full dentures, the annual dose equivalent to the whole body from external

exposure was 0.003 Sv (0.0003 mrem). For wearers of a few crowns, the estimated

annual dose equivalent from external exposure was 0.0005 Sv (0.05 rem). For all

wearers (65 million), the annual collective dose equivalent was 0.1 person-Sv

(10 person-rem). 

The dose estimates obtained by Buckley et al. (NUREG/CR–1775) for the accident and misuse

scenarios are summarized as follows:

For a fire in a warehouse, the maximum dose commitment was 5 mSv (0.5 rem) to the

lungs, and the whole-body dose to firefighters could be as high as 0.06 mSv (6 mrem). 

Conservative assumptions were made about the number of teeth in the warehouse and

the amount of uranium that would be released in a fire. No respiratory protection

equipment was assumed. 

For a dental technician working in a room in which uranium powder is lost during tooth

construction, and using conservative assumptions for the concentration of uranium in

dust in air and for exposure time, the whole-body dose was estimated to be 0.02 Sv

(2 rem).

Buckley et al. (NUREG/CR–1775) stated that the doses from alpha and beta particles to dental

wearers were conservative, but not greatly overestimated. Thompson (1976) stated that it was

possible that the absorption by saliva, pellicle, air and/or calculus may be sufficient in a portion

of the population to prevent any alpha from ever reaching the soft tissue. As with alpha dose

rates, the dose rate due to beta particles may be overestimated because of the assumption of

intimate and continuous contact between teeth and tissues. However, particle-attenuating

factors would not yield as significant a reduction for high-energy betas as for alpha particles

(Thompson, 1976).
3–26

3.2.3.2 Ophthalmic Lenses

Ophthalmic glass is used to manufacture lenses for eyeglasses and eyepieces. Buckley et al.

(NUREG/CR–1775) also analyzed radiological impacts on the members of the public from

eyeglasses containing thorium. The routine exposure scenarios evaluated were (1) exposure

during distribution by wholesale optical labs and dispensers, (2) exposure to wearers of

eyeglasses, (3) exposure to workers during collection of solid waste containing discarded

eyeglasses, and (4) exposure to the public following disposal of eyeglasses in a landfill or

incineration. However, only doses to eyeglass wearers were estimated. In the other scenarios,

doses were not estimated, because of the complexities in defining the exposure scenarios or to

the presumption that doses would be very low based on similar scenarios for other consumer

products, or because the dose estimates were unreasonably conservative, particularly for

disposal of eyeglasses in landfills.

For an individual wearing eyeglasses containing 0.05% by weight of thorium, dose analyses

were done for exposure to the eye, primarily from irradiation by alpha particles, and for wholebody exposures from photons. The dose estimates reported by Buckley et al.

(NUREG/CR–1775) and in NCRP 95 assume that an individual wears eyeglasses for 16 h/day

and that 55% of the population wears glasses. These dose estimates are summarized as

follows:

The annual dose equivalent to the cornea from alpha particles is about 0.04 Sv (4 rem). 

Assuming a tissue weighting factor of 10–4, the annual individual dose was estimated to

be 0.004 mSv (0.4 mrem).

The annual dose equivalent to the whole body of the individual from photons is about

2×104

 mSv (0.02 mrem).

In NCRP 95, it was estimated that 50 million people in the United States wear eyeglasses with

glass lenses. Assuming an annual individual effective dose equivalent (EDE) of 0.004 mSv

(0.4 mrem) from alpha particles and a whole-body dose equivalent of 2×104

 mSv (0.02 mrem),

the annual collective EDE to the U.S. population would be about 200 person-Sv

(20,000 person-rem). 

NCRP 95 also evaluated doses to the cornea of the eye from irradiation by alpha particles

during use of an instrument eyepiece containing 0.05% by weight of thorium. By assuming that

the instrument was used by an individual for 20 h/wk, the estimated annual dose equivalent to

the cornea from alpha particles is 0.03 Sv (3 rem).

The Yaniv study (AEC, 1974) concluded that radiation dose rates from ophthalmic glass could

be reduced significantly with better quality control of the rare earth and zirconium oxides. 

Another problem revealed in this study was that the observed radiation is not directly related to

the source material content of the glass, but due to the widely varying parent-progeny

equilibrium conditions. The radiation emissions are mainly due to the short-lived decay

products of 232Th and 238U, which can be present in glass even after the parent radionuclides

are removed. Therefore, control of source material was not sufficient to eliminate radioactive

material from glass (NCRP 95).
3–27

3.2.3.3 Mineral-Derived Products

This subsection is further divided into three sections. The first section describes studies that

estimated radionuclide air concentrations and doses due to industrial uses of mineral sand

products. The second section provides direct irradiation measurement results from the use of a

cellophane tape dispenser that contains zircon sand as ballast. The final section describes

studies that estimated doses and risks attributed to the disposal of mineral sand wastes, with an

emphasis on zircon sand wastes.

3.2.3.3.1 Industrial Worker Exposures

Four articles evaluated exposures associated with the handling of zircon and ilmenite (FeTiO2)

products. They are from Shimko (1996), Lischinsky et al. (1991), Wallace and Leach (1987),

and Boothe et al. (1980). In Shimko (1996), air sampling was performed at two plants that

process zircon sands. The first plant (Plant A) blends zircon flour, which is ground zircon sand,

with sodium carbonates, and pelletizes the mixture to produce a zirconium carbonate. At the

second plant, Plant B, zircon sand is ground in a ball mill to produce zircon flour. The zircon

flour is bagged and resold, primarily as a refractory and to make welding rods. Both plants

adhere to Occupational Safety and Health Administration (OSHA) regulations for ionizing

radiation (29 CFR 1910.96). Neither plant has an Nuclear Regulatory Commission (NRC)

license. The uranium and thorium concentrations in zircon sands are given as 286 to 343 ppm

and 116 to 157 ppm, respectively (Shimko, 1996). Measured air concentration data from plant

B is provided in Table 3.2.9. 

The EDEs due to inhalation were estimated. At Plant A, where the worker continuously

performs the same tasks, the annual EDE was estimated to be 3 mSv (300 mrem), without any

respiratory protection. At Plant B, the worker spends less than 10 h/wk processing zircon. 

Therefore, the worker’s weekly exposure was less than 8% of the derived air concentration

(DAC). However, if this worker processed material for 2000 h/yr, the estimated annual EDE

could be 17 mSv (1700 mrem), without respiratory equipment. At both plants the workers wore

National Institute for Occupational Safety and Health (NIOSH)-approved respirators for

zirconium compounds. 

During decommissioning activities of a former NRC-licensed ceramic manufacturing facility,

Lischinsky and Vigliani (1991) found concentrations of 0.053% uranium and thorium in

zirconium oxide that was marginally over the 0.05% weight criterion. The unlicenced zirconium

oxide powders were used in the manufacture of an electronic component.

Several screening measurements of other zirconium sands from different vendors showed

levels of radioactivity from very low trace quantities to about the same levels found by

Lischinsky and Vigliani (1991). The worst case air concentration of 0.03 Bq/m3

 (0.8 pCi/m3)

was estimated using the OSHA zirconium compound air concentration limit of 5 mg/m3

multiplied by the uranium concentration of 5.4 Bq/g (150 pCi/g) in the zirconium sand. The

worst case air concentration was considered to be about 15% of the unrestricted maximum

permissible concentration (MPC). Using the same approach for 232 Th, the estimated air

concentration was 2.2×103 Bq/m3

 (0.06 pCi/m3) in air, or roughly 6% of the unrestricted MPC

(Lischinsky and Vigliani, 1991).
3–28

Surface contamination measurements of the zircon sand were in the range of 250 to

333 Bq/100 cm2 (15,000 to 20,000 dpm/100 cm2), which is three times greater than the average

fixed beta-gamma limit stipulated in the NRC’s surface contamination guides (NRC,

Memoranda, Guidelines, 1982; Lischinsky and Vigliani, 1991). Furthermore, for dispersed

radioactivity, a uranium activity concentration of approximately 5.4 Bq/g (147 pCi/g) was 15

times greater than the unrestricted release guideline of 0.37 Bq/g (10 pCi/g) for natural

uranium. It was noted that the alpha contamination of approximately 7 to 8 Bq/100 cm2 (400 to

500 dpm/100 cm2

) was considered to be below the NRC average fixed surface contamination

limits (Lischinsky and Vigliani, 1991). 

In Wallace and Leach (1987), air sampling was performed to evaluate the potential exposures

during sand-blasting operations. Ilmenite is used extensively in the sand-blasting industry. The

ilmenite product used in Australia normally contains other minerals such as monazite and

zircon. In Australia, thorium concentrations in ilmenite range from 50 to 500 ppm and uranium

concentrations range from <10 to 30 ppm (Koperski, 1993). Particle sizing of the high-volume

dust samples indicated that 90% of the radioactivity was in the 0.45 to 5 micron particle size

ranges. Using the Y solubility and 1 micron activity median aerodynamic diameter, the

estimated annual EDE from inhalation to a sand operator, without protective respiratory

equipment, was 3 mSv (300 mrem). 

In Boothe et al. (1980), external radiation and radon levels were measured from zircon sand. At

contact the measured external exposure rate was 41 nanocoulomb (nC)/kg-h

(158 microroentgen (R)/h). Above the zircon sand at 0.9 meter, measured exposure rates

decreased to 18% of the on pile-readings. External measurements on three other comparable

piles of zircon used by foundries were within ± 20% of the survey reading of 44 nC/kg-h

(170 R/h).

The radon emanation rate measured above a stockpile of zircon sand was 0.074 Bq/m2

-s

(2 pCi/m2

-s). According to Boothe et al. (1980), the radon emanation rate from uranium mill

tailings is about 70 times higher than that from zircon sands. As explained by Boothe et al.

(1980), this low emanation rate is because radium in zircon occurs interstitially in the ZrSiO4

crystal, and radon is trapped within the crystalline lattice.

Early findings of the Oregon Radiation Control Section indicated that zircon sands used in

Oregon foundries originated from Australian placer mines and were stockpiled in large

quantities, in both bulk and bagged form (Boothe et al., 1980). Measurements made of a single 

40-kg bag of zircon sand were 44 nC/kg-h (170 R/h) at the surface and 3.91 nC/kg-h 

(15 R/h) at 0.9 meter above the bag. Measurements made of 40-kg bags of sand stacked

upon a pallet the size of 1.2 m × 1.2 m × 1.2 m gave exposure rates of 77 nC/kg-h (300 R/h)

at the surface and 21 nC/kg-h (80 R/h) at 0.9 m. Further, measurements made over 2×106

 kg

of bulk sand gave exposure rates of 52 nC/kg-h (200 R/h) at a height of 0.9 m above the

surface. All of the above readings were uncorrected for background, which ranged from 2 to

2.6 nC/kg-h (8 to 10 R/h). 

Another area evaluated by Boothe et al. (1980) was the possible exposure through inhalation

during the use of zircon sand as flour (200-mesh or 74-micron size). The size commonly used

is 200-mesh. The mesh number indicates only the screen size through which all particles will

pass, not the particle size spread. Sand of this nature is used in the mulling or mixing of the

mold formula by workers directly handling the mold by the use of the scoop or dump method. 
3–29

Air monitoring in a high-use area within one large carbon steel casting plant and at the

breathing zone level of five mullers at the plant showed no significant zircon flour dust

concentrations, with one exception. However, in Boothe et al. (1980), no air concentrations

were provided. Based on the limited information gathered, it could not be said that an

inhalation problem did not exist during the mold-making process. Further study was suggested.

An article written by Hipkin and Paynter (1991) summarized typical activity concentrations of

many materials that have a natural radioactivity content and estimated annual individual EDEs

due to external radiation and dust inhalation, as shown in Tables 3.2.10 and 3.2.11,

respectively. Though there was no detailed information in the article that indicated how the

activity concentrations were derived or how the EDEs were estimated, it was stated that realistic

assumptions were made when estimating doses and exposure durations. 

3.2.3.3.2 Public Exposure

In a survey of common cellophane tape dispensers used in Oregon, it was found that the

majority of tape dispensers used zircon or monazite sand for weight (Boothe et al., 1980). The

distinction was made due to the gamma-ray spectrum exhibited; zircon exhibits a predominantly

226Ra spectrum, as compared to a thorium-decay series spectrum given off by monazite. For

the zircon sand weight, the maximum exposure rate at the surface of the tape dispenser was

between 5 and 10 nC/kg-h (20 and 40 R/h). The exposure rate at 0.6 m was about

77 pC/kg-h (0.3 R/h) above background (Boothe et al., 1980; CRCPD, 1981).

3.2.3.3.3 Waste Disposal

In Oregon, there are 14 foundries using zircon as basic sand, flour, slurry, or paint (Boothe 

et al., 1980). As of 1980, these foundries used and disposed of about 4170 metric tons of

zircon each year. There is some recycling of zircon, and prior to disposal zircon is diluted with

silica sand and other materials. It was estimated there were 54,400 metric tons of foundry

zircon in landfills in the metropolitan area of Portland, Oregon. The radiological impact of zircon

sand is greatly limited by the low radon emanation of the sand. The maximum external

radiation level in landfills containing zircon sands was measured at 36 nC/kg-h (140 R/h). The

average level at such landfills appeared to be about 5 nC/kg-h (20 R/h). The Oregon

Radiation Control Section found no cases where houses or other structures were constructed

over zircon sand wastes. As noted in Boothe et al. (1980), this may not be true in more

industrialized parts of the country where there are more and larger foundries.

The largest volume of radioactive waste generated is sludge from the zirconium-hafnium

separation process (Boothe et al., 1980). The sludge is considered to be a good soil

conditioner. Approximately 10,900 metric tons of sludge were used on two agricultural fields

near Albany, Oregon, in 1975 and 1978. The gross external irradiation exposure rates over a

60-acre field ranged from 3.6 to 23 nC/kg-h (14 to 90 R/h). The average exposure rate was

8 nC/kg-h (30 R/h). Background exposure rates ranged from 2 to 2.6 nC/kg-h (8 to 10 R/h). 

The maximum exposure rate of the 24-acre field was 5 nC/kg-h (20 R/h). The radon

emanation rate over the 60-acre field was measured as 0.44 Bq/m2

-s (12 pCi/m2-s).

Special application metal waste is the category used by the EPA (RAE–9232/1–2) to describe

tailings and residues associated with zircon, hafnium, titanium, and tin processing. The
3–30

uranium and thorium concentrations used in the dose assessment for special application

wastes are given in Table 3.2.12. The special application metal processing waste site is

assumed to consist of commingled waste rock, overburden, and a tailing pile of 10 acres. The

pile was 200 meters by 200 meters with a height of 20 meters, and has a 1-meter-thick soil

cover. The pile contains 1.6 million metric tons of tailings, overburden, and wastes. The

generic site is assumed to be located in a rural area in Texas and underlain with an aquifer. 

The population density is assumed to be 65 persons per square mile (EPA, RAE–9232/1–2). 

Utilization and disposal of waste materials from these special application metals are varied. It is

assumed that the waste materials have negligible reuse. 

Summarized in Table 3.2.13 are the estimated annual EDEs for workers at the Texas disposal

site. For workers at the disposal site, the primary exposure pathways are direct irradiation and

dust inhalation. Only for the on-site office worker were indoor radon risks evaluated. It was

assumed the office was located on top of the closed waste pile. Summarized in Table 3.2.14

are the estimated annual individual EDEs to members of the public from storage and disposal

of special application metal processing wastes. Collective EDEs from storage or disposal of

these wastes are shown in Table 3.2.15.

3.2.3.4 Waste-Derived Products

The use of coal ash and phosphate slag in building and roadway construction materials and the

disposal of these wastes, as well as water treatment sludge, are described in the following

subsections.

3.2.3.4.1 Coal Ash

3.2.3.4.1.1 Product Use

Potential doses due to use of fly ash as an aggregate in concrete or cement were estimated for

the public by EPA (RAE–9232/1–2). It was assumed that 2% of the 6 million metric tons of fly

ash is used in various cement and concrete applications. EPA (RAE–9232/1–2) assumed the

generic dwelling area was 143 m2

 and that about 34 m3 of concrete were used in the dwelling. 

The thickness of the concrete was assumed to be 0.13 meter. The estimated annual EDE due

to direct irradiation from building materials was estimated to be 0.17 mSv (17 mrem). If 7,500

dwellings for 30,000 persons are constructed from 1 year’s reuse of coal ash, the resulting

annual collective EDE would be about 5.1 person-Sv (510 person-rem).

3.2.3.4.1.2 Coal Ash Disposal

The EPA (Cont. No. 68–02–4375) conducted a study to estimate the potential doses and risks

associated with environmental releases from coal and coal ash piles at utility and industrial

facilities with coal-fired boilers. The purpose of the study was to provide background

information to consider exempting coal and coal ash piles from the Comprehensive

Environmental Response, Compensation and Liability Act (CERCLA) reportable quantity

notification requirements. For direct irradiation and airborne exposures, the potential doses and

risks were analyzed for both the on-site worker and nearby resident. The 238U and 232Th

concentrations used by EPA (Cont. No. 68–02–4375) were 0.16 Bq/g (4.3 pCi/g) and 0.13 Bq/g

(3.5 pCi/g), respectively. The worker was assumed to be exposed only to an active disposal

cell with an effective surface area of 1000 m2

. The external radiation exposure rates were
3–31

estimated to be 2 nC/kg-h (8 R/h) and 5 nC/kg-h (20 R/h) for standing next to and standing

on the pile, respectively. In addition, potential doses from exposure to contaminated

groundwater and to surface water runoff were evaluated for the nearby resident. The estimated

annual EDEs for the coal ash pile worker and on-site resident are summarized in Table 3.2.16.

In 1993 the EPA conducted a dose assessment (EPA, RAE–9232/1–2) that evaluated the

potential doses and risks associated with ash disposal. The 238U, 235U, and 234U concentrations

in the coal ash were assumed to be 0.12 Bq/g (3.3 pCi/g), 0.007 Bq/g (0.2 pCi/g), and

0.12 Bq/g (3.3 pCi/g), respectively (see Table 3.2.12). The generic ash impoundment was

assumed to be 25,000 m2

 with a depth of 5 meters. The disposal site was located in the

Northeast and contained a total of 1.3 million metric tons of ash materials. The annual

individual EDE to workers and to on-site and off-site public receptors are summarized in Tables

3.2.13 and 3.2.14, respectively. Annual collective EDEs are summarized in Table 3.2.15.

3.2.3.4.2 Phosphate Slag

3.2.3.4.2.1 Product Use

During 1986 and 1987, the EPA (EPA/520/6-90/008) conducted a study to evaluate direct

irradiation exposures and attendant risks to the populations of Pocatello and Soda Springs,

Idaho, from the use of phosphate slag in the construction of roads and buildings. Gamma

measurements were made using both aerial surveys and ground surveys. These communities

each have phosphate ore processing facilities that have operated or are operating. Radioactive

slags from these plants have been used as an additive in materials for paving streets and

constructing building foundations. In Pocatello, the annual average and maximum individual

EDEs are about 0.14 mSv (14 mrem) and 1.5 mSv (150 mrem), respectively. In Soda Springs,

the annual average and maximum individual EDEs are about 0.5 mSv (50 mrem) and 2 mSv

(200 mrem), respectively (EPA/520/6-90/008). As shown in Table 3.2.17, the annual collective

dose equivalents are 8 person-Sv (800 person-rem) and 2 person-Sv (200 person-rem) for

Pocatello and Soda Springs, respectively (EPA/520/6-90/008).

3.2.3.4.2.2 Phosphate Slag Disposal

The EPA performed an assessment (RAE–9232/1–2) in which 238U, 235U, and 234U

concentrations in phosphate slag were assumed to be 0.9 Bq/g (25 pCi/g), 0.05 Bq/g 

(1.3 pCi/g), and 0.9 Bq/g (24 pCi/g), respectively (Table 3.2.12). The generic disposal pile is

assumed to contain 24 million metric tons of slag. This size is slightly larger than the largest

pile at an active facility. Assuming a square surface and a height of 20 meters for the slag pile,

the disposal site has a surface area of 630,000 m2

 (156 acres). The annual individual EDE to

workers and to on-site and off-site members of the public are summarized in Tables 3.2.13 and 

3.2.14, respectively. Annual collective EDEs are summarized in Table 3.2.15. 

3.2.3.4.3 Water Treatment Sludge

The EPA assumed that water treatment sludge is disposed in a sanitary landfill

(RAE–9232/1–2). The volume of sludge sent to the landfill over more than 20 years has a

volume of 7,500 m3

 when diluted by the other landfill wastes. The 238U, 235U, and 234U

concentrations are assumed to be 0.004 Bq/g (0.1 pCi/g), 3×105 Bq/g (7×104

 pCi/g), and 
3–32

0.004 Bq/g (0.1 pCi/g), respectively (Table 3.2.12). The size of the facility is assumed to be

400 meters by 400 meters and there is a 0.3-meter layer of topsoil placed over the fill area. 

The annual individual EDEs to workers and to on-site and off-site public receptors are

summarized in Tables 3.2.13 and 3.2.14, respectively. Annual collective EDEs to the public are

summarized in Table 3.2.15. 

3.2.4 Present Exemption Analysis

In the present exemption analyses, annual individual and collective EDEs are estimated for

truck drivers transporting zircon flour. To estimate EDEs for routine worker and public

exposures from uranium in dental products, the results from previous analyses are used in the

current assessment. Annual EDEs are calculated for industrial workers handling bulk zircon

flour and involved in sand-blasting operations. Mineral-derived products are of particular

interest for this exemption since, as is the case for zircon-derived products, the uranium and

thorium concentrations can be close to the 0.05% weight criterion. In the current assessment,

doses to other members of the public from the use of ophthalmic glass containing zirconium

oxides, tape dispensers containing zircon sand as ballast, and phosphate slag used to pave

streets and in building construction are based on previous analyses. Individual and collective

doses for the disposal of zircon sand are estimated in the current assessment and the results

are compared to previous evaluations. The accident scenarios evaluated include a warehouse

fire that contains zircon flour and a fire that occurs during zircon flour transport. 

3.2.4.1 Transport

In the transportation component of this current assessment, it is assumed that a single truck

driver is transporting 48 pallets of 0.05% by weight of uranium and thorium in zircon flour

(8.5 Bq/g (230 pCi/g) natural uranium in equilibrium with short-lived progeny and 0.65 Bq/g

(17.5 pCi/g) 232Th in equilibrium with its progeny). The primary exposure pathway to the truck

driver is direct irradiation. It is assumed there are fifty 41-kg bags per pallet. The dimensions

of the product on a pallet are 1.2 m × 1.2 m × 0.6 m (height). The pallets are assumed to be

stacked four high, six deep, and two wide. The distance between the truck driver and the load

is 0.9 meter and the thickness of the truck body is assumed to be 0.6 cm of steel. Using

MicroShield (Computer Codes, Grove Engineering, 1996), the estimated EDE rate is

1×104

 mSv/h (0.01 mrem/h). If the driver takes 24 hours to transport a load, the EDE per trip

is 0.002 mSv (0.2 mrem). If this driver makes 25 trips per year with the same type of load, the

annual EDE is estimated to be 0.06 mSv (6 mrem). Further, if 1000 truck drivers make 25 trips

annually, the annual collective EDE is estimated to be 0.06 person-Sv (6 person-rem).

3.2.4.2 Routine Exposures

In this subsection, the EDEs from routine worker and public exposures to uranium in dental

products, zircon and ilmenite products, and phosphate slags are summarized. For dental

products and phosphate slags, the results from previous analyses, as described in

Section 3.2.3, are used in the current assessment. As mentioned earlier, mineral-derived

products are of particular interest for this exemption since, as is the case for zircon-derived

products, the uranium and thorium concentrations can be close to the 0.05% weight criterion.
3–33

3.2.4.2.1 Dental Products

NCRP 95 estimated that the annual dose equivalent to the basal mucosa of a dental prostheses

wearer from beta particles was 5 mSv (0.5 rem) at 0.02% by weight of uranium. On the basis

of a weighting factor of 0.01 for the human skin, and assuming that irradiation of the basal

mucosa is equivalent to irradiation of 1% of the skin, the EDE was estimated to be 5×104

 mSv

(0.05 mrem) (NCRP 95). NCRP 95 assumed 45 million people were wearing dental prostheses

containing uranium. As the average life of the prosthetic device is not expected to exceed 11

years and 14 years have passed since that report, it is reasonable to assume that only 1 million

people are still wearing some type of dental prostheses with uranium. The resulting annual

collective dose equivalent to the U.S. population from beta particles would be conservatively

0.5 person-Sv (50 person-rem).

Buckley et al. (NUREG/CR–1775) assumed that the annual EDE to an individual due to external

irradiation is 0.03 Sv (0.003 mrem). Assuming that the same number of porcelain dental

prostheses containing uranium (approximately 1 million) are currently worn in the United States,

the conservative collective EDE from gamma irradiation is about 0.03 person-Sv

(3 person-rem). This is only 6% of the skin EDE.

Therefore, the annual individual EDE from wearing a porcelain dental prosthesis containing the

average weight % of uranium is 5×104

 mSv (0.05 mrem). The annual collective EDE due to

beta particles and external irradiation is conservatively estimated to be 0.5 person-Sv

(50 person-rem). This dose is expected to decrease significantly over time as porcelain without

uranium and other types of materials used for dental prostheses replace the old porcelain

containing uranium. At the maximum allowed weight %, the annual EDE would be two and

one-half times the dose for the dentures of average weight %.

3.2.4.2.2 Mineral-Derived Products

In this assessment, exposures to zircon-derived products are used to represent the upper

bound of potential doses from the broad category of mineral-derived products. Zircon-derived

products were selected since the uranium and thorium concentrations can be at or close to the

0.05% weight exemption criterion. For the worker exposure scenario, ilmenite as a sand

abrasive product is also evaluated.

Two major routes of exposure appear to dominate: dust inhalation and external irradiation. In

most situations, dust inhalation is the dominant exposure pathway, especially if the particulate

mass loading in air is high. The radioactive content of the dust is predominantly due to the

monazite or bastnasite content, which was not completely removed from the other minerals

during the separation process (Koperski, 1993). Other sources of radioactivity in the minerals

are due to adsorption of radionuclides onto minerals and incorporation of thorium and uranium

into the mineral lattice during the crystallization process (Koperski, 1993).

The radon emanation rate measured above a stockpile of zircon sand is about 0.074 Bq/m2

-s

(2 pCi/m2

-s) (Boothe et al., 1980). As explained by Boothe et al., this low emanation rate is

because radium in zircon occurs interstitially in the ZrSiO4 crystal, and radon is trapped within

the crystalline lattice. Radon emanation rate measurements over dry land have been reported

to range from 2×104

 to 0.05 Bq/m2-s (5×103 to 1.4 pCi/m2-s) (NCRP 103). The average radon

emanation rate in background soils is about 0.02 Bq/m2

-s (0.5 pCi/m 2-s) (NCRP 103). 
3–34

Therefore, the radon emanation rate associated with zircon sand is marginally higher than

expected in normal background soil.

3.2.4.2.2.1 Worker Exposures

Typically, mineral-derived products that have not been highly processed, such as zircon flour

and ilmenite used in sand-blasting operations, are more likely to contain elevated

concentrations of uranium and thorium as compared to higher purity products (Phone call,

J. B. Hedrick, Thorium and Rare Earth Specialist, Bureau of Mines, U.S. Department of the

Interior, 1995). As mentioned earlier, the two primary exposure pathways appear to be external

irradiation and dust inhalation.

MicroShield (Computer Codes, Grove Engineering, 1996) was used to estimate the EDE rate at

1 meter from 20 pallets of zircon flour containing natural uranium (8.5 Bq/g (230 pCi/g)) in

activity equilibrium with its short-lived progeny and 232Th (0.65 Bq/g (17.5 pCi/g)) in equilibrium

with its progeny. Assuming no shielding, the calculated EDE rate was about 2×104

 mSv/h

(0.02 mrem/h). If the exposure duration was 2000 h/yr, at this distance, an unlikely event, the

annual EDE from direct irradiation is about 0.4 mSv (40 mrem). The direct gamma-exposure

rate measured by Shimko (1996) at about 0.9 meter from a stack of bags was 10 nC/kg-h 

(0.04 mR/h). In Booth et al. (1980), the direct irradiation exposure rate, correcting for

background, measured at a distance of 0.9 meter from a pallet of zircon flour was about 

18 nC/kg-h (0.07 mR/h). The differences in the measured exposure rates may be from

variations in actual thorium and uranium concentrations in the zircon product and the possible

contribution of uranium long-lived decay products.

Shimko (1996) evaluated the worker exposure from the processing of zircon sands in Plant A

and Plant B. The respirable dust concentration, 1.4 mg/m3

, was measured at Plant A, which

blends zircon flour with sodium carbonate to produce zircon carbonate. For this assessment, it

is assumed that the product concentration is the same as the airborne zircon flour dust, i.e.,

natural uranium (8.5 Bq/g (230 pCi/g)) in equilibrium with its short-lived progeny and 232Th 

(0.65 Bq/g (17.5 pCi/g)) in equilibrium with its progeny. With an exposure duration of

2000 hours, the annual EDE is estimated to be 2 mSv (200 mrem). This estimate is potentially

low by a factor of about 2 if consideration is given to the presence of the other uranium decay

products. Radionuclide air concentrations were measured in Plant B where zircon sand is

ground to zircon flour, bagged, and resold. Using measured area (near bagger) airborne

radionuclide concentrations (as shown in Table 3.2.9) and assuming an exposure duration of

2000 hours, the estimated annual EDE was about 7 mSv (700 mrem). Using personal sampler

data (bagger operator) and assuming an exposure duration of 2000 hours, the estimated

annual EDE was about 40 mSv (4 rem). In Shimko (1996), the estimated annual EDE

associated with Plant A and Plant B (bagger operator) activities was estimated to be about

3 mSv (300 mrem) and 17 mSv (1700 mrem), respectively, based on calculations using DAC

values from 10 CFR 20, including consideration of a 10 micron particle size for uranium. As

mentioned in Shimko (1996), dust respirators were worn by the workers in both Plant A and

Plant B and the actual exposure duration for the bagger operator (Plant B) was less than that

used to estimate the annual EDE. 

In Wallace and Leach (1987), the mean dust concentration for sand-blasting operations was

10.4 mg/m3

. Radioactivity measured from the high-volume dust samples indicated that 90% of

the particles were in the size range of 0.45 to 5 microns, the same particle size ranges collected
3–35

by personal air samplers. The maximum measured area dust concentration during

sand-blasting operations was 40.3 mg/m3. However, sand blasting operators usually work from

an upwind location so the dust is blown away from the work station.

In ilmenite, an abrasive product used in sand blasting, the average 238U concentration is

0.092 Bq/g (2.5 pCi/g) and the 232Th concentration is 0.2 Bq/g (5.5 pCi/g) (Wallace and Leach,

1987). Assuming the average radionuclide concentrations (including 238 U, 234U, and 235U in

equilibrium with short-lived progeny and 232Th in equilibrium with its progeny) and mean

respirable dust concentration (10.4 mg/m3

), the annual EDE is estimated to be 3 mSv

(300 mrem). This estimated annual EDE is very similar to that determined by Wallace and

Leach (1987). Depending on the type of dust respirator used regularly, the annual EDE could

be reduced by a factor of 5 to 10 (NIOSH, 1990). It should not be assumed that respirators are

always worn by workers handling zircon flour or similar mineral sand products. 

The annual EDE from both external irradiation and dust inhalation to an industrial worker

handling bulk zircon product is estimated to range from about 3 mSv (300 mrem) to 7 mSv

(700 mrem) assuming no respiratory protection. For the bagger operator, the annual EDE was

estimated to be about 35 mSv (3500 mrem); however, the actual exposure duration may be

less than that used to estimate the EDE (Shimko, 1996). The annual EDE due to the use of

ilmenite in sand blasting is estimated to be about 3 mSv (300 mrem). Ilmenite is not the only

product that can be used as an abrasive; zircon has been used in the past and low silicatecontaining minerals may also be used. As shown, the dominant exposure pathway is dust

inhalation when a dispersible product is used. 

The handling of bulk mineral-derived products full-time should be considered an upper bound to

the potential doses that may occur during the use of products that contain 0.05% by weight of

source material. It is unknown how many workers may be handling bulk zircon or other

mineral-derived products with uranium and thorium concentrations nearing (or exceeding) the

0.05% weight exemption limit. However, if it is assumed that 1000 workers handle similar types

of bulk zircon products annually, the annual collective EDE could range from 3 person-Sv

(300 person-rem) to 7 person-Sv (700 person-rem).

3.2.4.2.2.2 Public Exposure

Public exposures to zircon-derived products include the use of ophthalmic glass and exposure

to zircon sand ballast for tape dispensers. Due to the potential use and distribution of such

products in the public and the elevated concentrations of thorium and uranium, these products

are highlighted. Undoubtably many other mineral-derived products are disseminated.

Ophthalmic Glass. It is estimated that the number of people wearing eyeglasses in the United

States is 145 million (VCA, 1999). Currently, it has been estimated that only 8%, or about 12

million people, wear eyeglasses that have glass lenses (OA, 1999). Assuming the same annual

individual EDE of 0.004 mSv (0.4 mrem) from alpha particles and the annual whole-body dose

equivalent of 2×104

 mSv (0.02 mrem), as described by Buckley et al. (NUREG/CR–1775) and

in NCRP 95, the annual collective EDE to the U.S. population that currently wears glass-lensed

eyewear would be about 50 person-Sv (5,000 person-rem).

Tape Dispenser Ballast. In Boothe et al. (1980) and CRCPD (1981), the measured exposure

rate at 0.6 meter from tape dispensers was about 77 pC/kg-h (0.3 R/h) above background. 
3–36

Assuming an individual spent 80% of a 40-hour work week seated within 0.6 meter of a tape

dispenser, the annual EDE is estimated to be 0.005 mSv (0.5 mrem). Based on available

information, only one to two known manufacturers used zirconium sands as tape dispenser

ballast (CRCPD, 1981). The manufacturer of the model tested no longer uses radioactive

sands for ballast (CRCPD, 1981). Assuming 10,000 people use such tape dispensers (limited

distribution), the estimated collective EDE is estimated to be 0.05 person-Sv (5 person-rem).

3.2.4.2.3 Phosphate Slag

The results from the EPA (EPA/520/6-90/008) study that evaluated gamma exposures and

attendant risks to the populations of Pocatello and Soda Springs, Idaho, from the use of

phosphate slag are used in the current analysis. The average annual individual and collective

EDEs attributed to paving streets and constructing building foundations in the communities of

Soda Springs and Pocatello, Idaho, are summarized in Table 3.2.17. The collective EDE is

based on a 10-year useful lifetime. Idaho and Montana have both prohibited the use of

phosphate slag in structures but continue to permit its use in highway and other construction

(CRCPD, 1981).

3.2.4.3 Disposal

The disposal of zircon sand in municipal waste landfills is the primary focus of the current

assessment. The generic disposal methodology, as described in Appendix A.2, is used to

estimate annual individual and collective EDEs to landfill workers and other members of the

public from the disposal of zircon sand. For the waste collectors, it was assumed that the waste

zircon sand was transported in an off-loader with 15.3 m3

 capacity to the landfill and the sourceto-receptor distance was 1 meter. The trailer dimensions are 6.7 m × 2.4 m × 1.4 m and the

weight capacity is 10 metric tons (Phone call, Vicki Esponge, Customer Service, Waste

Management Incorporated, Knoxville, TN, May 1997).

In Oregon, foundries disposed of about 4180 metric tons of zircon each year (Boothe et al.,

1980). For this analysis, it is assumed this amount of zircon is disposed in each State per year,

resulting in a total amount of 2.1×105

 metric tons of zircon disposed annually in the United

States. It is also assumed the zircon sand is uniformly distributed among 3,500 operating U.S.

municipal landfills, which results in about 60 metric tons of zircon sand disposed of annually in a

single landfill. The zircon sand contains 0.05% by weight of source material (natural uranium 

8.5 Bq/g (230 pCi/g) in equilibrium with its short-lived progeny and 232Th 0.65 Bq/g (17.5 pCi/g)

in equilibrium with its progeny). The estimated annual individual and collective EDEs for

collectors, landfill operators, off-site receptors, and future on-site receptors from disposal of

zircon sand are summarized in Table 3.2.18.

There are major differences between the assumptions and methods used in the present

assessment as compared to those used in EPA RAE–9232/1–2). In the current assessment,

232Th and its decay products and only 238U, 234U, and 235U and their short-lived decay products

are taken into account, whereas in EPA (RAE–9232/1–2), 226Ra and its decay products are also

considered. The present assessment evaluates the disposal of zircon sand product disposed in

municipal landfills, whereas EPA (RAE–9232/1–2) evaluates the disposal of the tailings and

residues associated with the processing of zircon sands. Other differences include radionuclide

concentrations, amount of commingled waste, waste site characteristics, impacted populations,

and certain exposure assessment parameter values.
3–37

3.2.4.4 Accidents

In this evaluation, the warehouse and transportation accident scenarios were selected to

represent upper bound doses associated with a product that contains 0.05% by weight of

uranium and thorium. It is difficult to image a fire causing a significant fraction of the zircon

flour to become airborne; however, firefighting activities could disperse the material and

cleanup workers/investigators could encounter contamination. It is assumed 1000 zircon flour

41-kg bags are stored in a warehouse or are being transported. The uranium concentration is

about 8.5 Bq/g (30 pCi/g) and the thorium concentration is about 0.65 Bq/g (17.5 pCi/g). Using

the generic accident methodology described in Appendix A.1, the EDE to a firefighter

combating a warehouse fire is 0.004 mSv (0.4 mrem) and the EDE to a firefighter combating a

transportation fire is estimated to be 0.03 mSv (3 mrem). EDEs for workers involved in cleanup

from the warehouse and transportation accidents were estimated to be 0.03 mSv (3 mrem) and

0.3 mSv (30 mrem), respectively. Dose estimates obtained by Buckley et al.

(NUREG/CR–1775) for the accident and misuse scenarios of dental products are used in the

present analyses. These dose estimates are summarized in Section 3.2.3.1.

3.2.5 Summary 

This exemption limits the involvement of the NRC in the regulation of source material. It

exempts from licensing many naturally occurring and technologically enhanced sources of

uranium and thorium considered nonnuclear. This exemption potentially can cover a wide

range of products and materials. In this analysis, the types of product evaluated were divided

primarily into two major categories: mineral-derived products and waste-derived products. 

Uranium in dental products did not fit into either category. Therefore, it was evaluated

separately. Zircon products were used to represent the upper bound of doses that could be

attributed to mineral-derived products. Phosphate slag in building and road construction

materials were used to evaluate the doses from waste-derived products. Evaluation of doses

from landfill disposal focused on disposal of zircon sands in municipal waste landfills.

In this analysis, estimates of individual and collective doses to members of the public from

routine use of consumer products were based primarily on available published information. 

Doses from transport, routine uses, disposal, and accidents involving zircon flour and sand

were estimated for this analysis. The estimated doses are intended to provide reasonable

upper bounds for current practices. The results of this current assessment are summarized in

Table 3.2.19. 

Based on this analysis, the following general conclusions about radiological impacts on the

public associated with this exemption can be obtained:

Uranium is no longer used in U.S.-produced porcelain dental products. It is considered

very unlikely that uranium will be used for this application in the future.

The primary exposure pathways for the public appear to be due to dust inhalation and

direct irradiation. Radon emanation from zircon products and byproducts appears to be

very low. As explained by Boothe et al. (1980) and Koperski (1993), the low emanation

rate is because radium in zircon occurs interstitially and the radon is trapped within the
3–38

crystalline lattice. Chemical processing of mineral sands can alter this configuration and

may result in radon release.

Doses to industrial workers involved in routine bulk product (e.g., zircon flour) handling

operations are greater than those from transport or disposal. It is unknown how many

U.S. industrial workers may be involved with the handling of bulk mineral-derived

products and whether respiratory protection is routinely used. Industrial workers not

handling bulk (dispersible) products routinely are expected to have lower doses than

those working in dusty environments.

The use of ophthalmic glass in eyeglasses is declining as plastic lenses are replacing

glass. Tape dispensers that used zircon sand as ballast are no longer manufactured. It

is unknown whether there are other products that may result in higher doses to the

public.
10 Hewson and Upton, 1996

3–39

Figure 3.2.1 Heavy Mineral Concentrate Process Schematic 10
3–40

Table 3.2.1 Source Material Concentrations Associated With 0.05 Percent Weight 

of Source a,b

Radionuclide

Specific Activity

(Ci/g)c

Radionuclide Concentration

(500 ppm)

(pCi/g) (Bq/g)

Natural uranium

 (238U+ 235U+ 234U)

6.8×107 340 13

Depleted uranium

 (238U+ 235U+ 234U)

3.6×107 180 7

238U 3.3×107 165 6

Natural/processed

 thorium

 (232Th + 228Th)

2.2×107 110 4

232Th 1.1×107 55 2

a

 Conversions: Natural uranium at 250 Bq/g (6800 pCi/g) per % weight to 0.025 Bq/g

(0.68 pCi/g) per ppm. Depleted uranium at 130 Bq/g (3600 pCi/g) per % weight to 0.013 Bq/g

(0.36 pCi/g) per ppm. Natural and processed thorium at 40 Bq/g (1100 pCi/g) per % weight to

0.004 Bq/g (0.11 pCi/g) per ppm. 

b

 Conversion of weight % to radionuclide concentration does not include decay products;

however, decay products are included in the dose calculations.

c

 1 Ci/g = 0.037 TBq/g.
3–41

Table 3.2.2 Naturally Occurring Radioactivity Related to Mineral Resources a

Mineral Mineral or Waste Radioactivity

AT OR ABOVE THE 0.05 PERCENT WEIGHT CRITERIONb

Monazite Thorium 5–7% and uranium 0.1–0.3%

Niobium (columbium)-Tantalum Imported ores for production

Tin (thorium 0.4–4%)

pyrochlore ore (thorium oxide 1–3.3%,

uranium oxide 3.1%)

euxenite ore Canada (0.35–0.54% uranium,

2.5–4% thorium)

AT OR BELOW THE 0.05 PERCENT WEIGHT CRITERION

Aluminum Bauxite ore: 0.25 Bq/g (6.8 pCi/g) 238U,

0.3 Bq/g (8.2 pCi/g) 232Th, and

0.26-0.29 Bq/g (7–8 pCi/g) 228Th

Red mud slurry: 0.52 Bq/g (14 pCi/g) 238U

Coal U.S. average concentrations: 4.7 ppm thorium and

1.8 ppm uranium 

anthracite 5.4 ppm thorium and 1.5 ppm

uranium

bituminous 5 ppm thorium and 1.9 ppm

uranium

lignite 6.3 ppm thorium and 2.5 ppm uranium

Coal ash: 0.06–0.32 Bq/g (1.5–8.6 pCi/g)

238U, 0.01–0.28 Bq/g (0.4–7.5 pCi/g) 232Th

Fly ash: 0.2 Bq/g (5.4 pCi/g) 238U and

0.07 Bq/g (1.9 pCi/g) 232Th

Copper 1–100 ppm uranium in copper ores

36 pCi/g (1.3 Bq/g) 238U Arizona Miami

District, 0.016% mean (U3O8)

Phosphate 238U 20–200 ppm (0.26–2.5 Bq/g

(7–67 pCi/g))

232Th 1–5 ppm (0.003–0.02 Bq/g

(0.1–0.6 pCi/g))
Table 3.2.2 Naturally Occurring Radioactivity Related to Mineral Resources a (continued)

3–42

Mineral Mineral or Waste Radioactivity

AT OR BELOW THE 0.05 PERCENT WEIGHT CRITERION

Titanium-bearing mineral sands Rutile: thorium <50–350 ppm, uranium

<10–20 ppm

Leucoxene: thorium 80–700 ppm, uranium

20–50 ppm

Ilmenite: thorium 50–500 ppm, uranium

<10–30 ppm

Sludge ponds: 2.15 and 1.0 Bq/g (58 and

27 pCi/g) 238U and 234U, respectively

Zircon mineral sands Thorium 150–250 ppm 

Uranium 150–300 ppm

baddeleyite up to 7.4 Bq/g (200 pCi/g)

uranium radionuclides

a

 NCRP 118; EPA, RAE–9232/1–2; CRCPD, 1981; CRCPD, 1994; Koperski, 1993;

McBride et al., 1978; and Corbett, 1983.

b

 Conversions: Natural uranium at 250 Bq/g (6800 pCi/g) per % weight to 0.025 Bq/g

(0.68 pCi/g) per ppm. Depleted uranium at 130 Bq/g (3600 pCi/g) per % weight to 0.013 Bq/g

(0.36 pCi/g) per ppm. Natural and processed thorium at 40 Bq/g (1100 pCi/g) per % weight to

0.004 Bq/g (0.11 pCi/g) per ppm.
3–43

Table 3.2.3 Uranium and Thorium Concentrations in Common Building Materials a

Material

Uranium Concentration

pCi/g (ppm)b

Thorium Concentration

pCi/g (ppm)b

Granite 1.7 (4.7) 0.2 (2)

Cement 1.2 (3.4) 0.6 (5)

Byproduct gypsum 5 (13.7) 1.8 (16)

Dry wallboard 0.4 (1.0) 0.3 (3)

a

 NCRP 94.

b

 1 pCi/g = 0.037 Bq/g. Conversions: Natural uranium at 250 Bq/g (6800 pCi/g) per % weight to

0.025 Bq/g (0.68 pCi/g) per ppm. Depleted uranium at 130 Bq/g (3600 pCi/g) per % weight to

0.013 Bq/g (0.36 pCi/g) per ppm. Natural and processed thorium at 40 Bq/g (1100 pCi/g) per

% weight to 0.004 Bq/g (0.11 pCi/g) per ppm.
3–44

Table 3.2.4 Examples of Mineral-Derived Products

Mineral Products (Intermediate) Products (Final)

Bauxite and

 aluminum

Alumina-feedstock, aluminum

 reduction industry

Aluminum products

Rare earths Catalysts, ceramics, refractory and

 metallurgical processes, magnets,

 lighting, phosphors, glass and optics,

 electronics

Copper 40% building and construction industries

25% electrical and electronic industries

38% industrial machinery and equipment,

power industry, and transportation industry

Zinc 45% automobile industry-galvanized

 sheet metal

20% brass manufacturing

15% die casting

Titanium Titanium tetroxide: titanium

 dioxide + titanium sponge

Rutile: Titanium metal

Titanium dioxide—pigment in paper, paint,

 plastics, cosmetics, and ceramics

Titanium sponge—aircraft engines and

 airframes.

Titanium metal—aerospace industry

 aircraft frames, and jet engines

Zirconium

 and hafnium

Byproduct of mining and

 extracting titanium minerals,

 ilmenite and rutile

Sponge metal and crystal bar

Zircon: High production included

 refractory bricks and shapes,

 alumina-zirconia abrasives, foundry

 sands, and investment casting, milled

 and micronized zircon, zirconium

 chemicals, and zirconia

High-value products are cubic zirconia,

 technical zirconia ceramics, superalloy

 castings, zirconia textile refractories, and

 specialty chemicals

Zirconium (zircon)—foundry sands,

 refractory paints, and other refractory

 materials.

Hafnium—nuclear reactor control rods

Phosphate Elemental phosphorus,

phosphoric acid, phosphates

Fertilizers and soil conditioners, backfill and

 road-base materials, additives to concrete

 block, mine reclamation, and chemical

 feedstock, e.g., sulfur recovery
3–45

Table 3.2.5 Mineral Extraction Processing Methods

Process Description Byproduct

Beneficiation Crushed ores are concentrated

 to free valuable mineral and

 metal components from the

 matrix of less valuable rock

 (called gangue). Beneficiation

 processes can include physical

 and chemical separation

 techniques, such as gravity

 concentration, magnetic

 separation, electrostatic

 separation, flotation, ion

 exchange, solvent extraction,

 electroplating, precipitation,

 and amalgamation.

Nearly all beneficiation processes

 generate tailings, which can be

 considered waste material or

 can undergo further mineral

 extraction.

Milling Concentration of mineral ore

 further by physical or chemical

 processing.

Titanium: Chlorination in a

 fluidized-bed reactor-feedstock

 for production of titanium

 tetrachloride and titanium

 sponge.

Tailings.

Smelter and refining The mineral concentrates, such

 as copper concentrate, are

 refined in a smelter. The

 smelting stage may include

 roasting, smelting, and

 converting. Smelting may

 consist of a pyrometallurgical

 technique or electrolytic

 production technique. Other

 processes include plasma

 fusion and electric-arc

 techniques, e.g., zirconium

 oxide is produced from zircon.

Slag, bag house dust.
3–46

Table 3.2.6 Uranium in Coal and Ash a

Type of Coal

Coal-Uranium Concentration

g/g (ppm)

Ash-Uranium Concentration

g/g (ppm)

Anthracite 1.5 9

Eastern bituminous 1.9 38

Western bituminous 1.9 16

Lignite 2.3 22

a

 NCRP 77.
3–47

Table 3.2.7 Typical Radionuclide Concentrations in Phosphate Ore, 

Phosphate Slag, and Phosphogypsum

Radionuclide Concentrationa

(pCi/g)b

Material 238U 234U 235U 226Ra 230Th 232Th 228Th

Phosphate

 ore

3 3

3

31 5 27 36

Phosphate

 slag

25 24 1.3 35 32 0.8 0.8

Phospho-

 gypsum

6

6

0.3 33 13 0.3 1.4

Background

 soil

0.3 0.3 0.5 0.3 0.5 0.7

a

 EPA, RAE–9232/1–2.

b

 1 pCi/g = 0.037 Bq/g
3–48

Table 3.2.8 Uranium Concentrations in Dental Products and Estimated Doses

Product Concentrations

Annual Dose

Equivalent Comments Reference

Porcelain

 teeth

0.10% weight 6 Sv (600 rem)

0.028 Sv (2.8 rem)

Oral mucosa, α

Oral mucosa, β

O’Riordan

 and Hunt

 (1974)

0.05% weight

 (500 ppm)

4 Sv (400 rem/yr) Teeth surface Papastefanou

 (1987)

Dental

 products

0.044% (maximum

 concentration)

1.3 Sv (130 rem)

0.016 Sv (1.6 rem)

Oral mucosa, α

β dose rates

Thompson

 (1976)

0.02% weight average

 uranium

 concentration

 in U.S. dental

 porcelain 

7 mSv (0.7 rem)

5 mSv (0.5 rem)

Basal cell layer

 from β (200 m 

 depth) (U/K-40)

Basal cell layer

 from β particles

 (U)

Thompson

 (1976)

Dental

 porcelains

Shofu–Ace 5.2 ppm

Shofu–Real 342 ppm

G.C. Livdent 47 ppm

Trubyte Bioblend

 235 ppm

Higher and

 more uniform 

 concentration

 in U.S. product

 as compared

 to Japanese

Sairenji et al.

(1980)

Dental

 powders

345–1,090 ppm

 uranium powders

 used in superficial

 part of crown

1.6–2.7 ppm uranium

 used in core part and

 top of crown

Determination of

 15 elements

 used in dental

 porcelain

 powders sold

 in Japan

Noguchi et al.

 (1980)

Dental

 porcelains

0.05% weight 1–4 Sv

 (100–400 rem)

14 mSv (1.4 rem)

0.003 Sv

 (0.3 rem)

0.0005 Sv

(0.05 rem)

Teeth surface, α

Basal cell layer, β

Direct radiation,

 full dentures

Direct radiation,

 few crowns

NUREG/CR–1775
3–49

Table 3.2.9 Measured Air Concentrations at Plant B a

Air Concentrations

(Ci/cm3

)

b

Location 238U 235U 234U 232Th 230Th 228Th

Personnel

Area on west

 side of room

Area near bagger

1.3×1012

1.4×1013

2.7×1013

6.0×1013 1.9×1012

1.6×1013

2.7×1013

1.4×1013

2.3×1014

6.5×1014

2.0×1012

1.1×1013

2.2×1013

5.0×1013

3.5×1014

1.0×1013

a

 Shimko, 1996.

b

 1 Ci/cm3 = 3.7×104 Bq/cm3.
3–50

Table 3.2.10 Typical Activity Concentrations of Materials in Use a

Activity Concentrations

(pCi/g)b

Use 232Th 228Th 238U 230Th

Refractories

 Company 1

 Company 2

27

27

54

54

270

270

270

270

Zirconia 27 240 135 27

Glazes 11 135 27 540

Glass 81 135 27 27

Casting 14 14 14 15

Rare earth 270 270 0 0

Special alloys 190 190 160 160

Tin smelting

 Ore

 Fumes

27

 0

27

0

8

0

8

0

Titanium

 Ilmenite

 Rutile

 Chlorinator bed

27

5

270

27

5

270

14

5

140

14

5

140

Phosphate 0 0 41 41

a

 Hipkin and Paynter, 1991.

b

 1 pCi/g = 0.037 Bq/g.
3–51

Table 3.2.11 Individual Exposures: External Irradiation and Dust Inhalation 

Exposure Routes a

Use

Annual Effective

Dose Equivalent

Due to External

Irradiation

(mrem)b

Annual Effective Dose

Equivalent Due to Dust

Inhalation (no 

respiratory protection)

(mrem)b

Annual Effective Dose

Equivalent Due to Dust

Inhalation (respiratory

protection)

(mrem)b

Refractories 30 600 60

Zirconia 70 1,500 150

Glazes 70 400 40

Glass 30 500 50

Casting 0 10 <10

Special alloys 50 300 30

Titanium

 Ilmenite

 Rutile

 Chlorinator bed

 Scale

6

2

0

400

1,500

20

300

0

150

<1

30

0

Phosphate

 Rock

 Scale

4

400

30

0

<10

0

a

 Hipkin and Paynter, 1991.

b

 1 mrem = 0.01 mSv.
3–52

Table 3.2.12 Radionuclide Concentrations Used in Storage and Disposal Assessment a

Radionuclide Concentration

(pCi/g)c

Type of Waste 238U 234U 235U 232Th 228Th

Special application

 metalsb

43 43 2.2 22 22

Coal ash 3.3 3.3 0.2 2.1 3.2

Phosphate slag 25 24 1.3 0.8 0.8

Water treatment

 sludge

0.1 0.1 7×104 0.005 0.2

a

 EPA, RAE–9232/1–2.

b

 Special application metals include zirconium, hafnium, titanium, and tin.

c

 1 pCi/g = 0.037 Bq/g.
3–53

Table 3.2.13 Individual Effective Dose Equivalent to Workers From Storage

or Disposal of Wastes a

Annual Effective Dose Equivalent

(mrem)b

Exposure

Scenario

Special

Application

Metalsc Phosphate Slag Coal Ash

Water

Treatment

Sludges

Gamma irradiation 2×102 1×102 1×101 2

Dust inhalation 3 0.4 0.06 7×105

a

 EPA, RAE–9232/1–2.

b

 1 mrem = 0.01 mSv.

c

 Special application metals include zirconium, hafnium, titanium, and tin.
3–54

Table 3.2.14 Individual Effective Dose Equivalent to Members of the Public From 

Storage or Disposal of Wastes a

Annual Effective Dose Equivalent

(mrem)b

Exposure

Scenario

Special

Application

Metals

Phosphate

Slag Coal Ash

Water

Treatment

Sludges

ON-SITE INDIVIDUAL

Gamma irradiation 3×104 400 40 0.08

Dust inhalation 3 4×101 6×102 0.007

Total 3 400 40 0.09

OFF-SITE INDIVIDUAL

Gamma radiation 7×105 100 10 0.02

Dust inhalation 3×104 0.3 0.03 8×104

Drink-contaminated

 well water

0.008 5×104 0.04 4×106

Food-contaminated

 well water

0.01 4×104 0.1 6×106

Total 0.02 100 10 0.02

a

 EPA, RAE–9232/1–2.

b

 1 mrem = 0.01 mSv.
3–55

Table 3.2.15 Collective Effective Dose Equivalents From Storage or 

Disposal of Wastes a,b

Annual Collective Dose Equivalent

(person-rem)c

Exposure

Scenario

Special

Application

Metals

Phosphate

Slag Coal Ash

Water

Treatment

Sludges

Exposure to

 resuspended

 particulates

6×105 0.03 0.2 7×104

River water

 contaminated

 by groundwater

3×104 6×104 0.003 3×106

 Total 4×104 0.03 0.2 7×104

a

 EPA, RAE–9232/1–2.

b

 The population densities used in the estimation of the collective EDEs are as follows. For

special application metals, the population density is assumed to be 65 persons per square mile. 

The population density assumed for phosphate slag disposal is the average value for Idaho,

which is approximately 12 persons per square mile; for coal ash disposal, the population density

was based on the average population distribution of four Northeastern states: New Jersey,

Connecticut, Rhode Island, and Massachusetts. For water treatment sludges, the population

density is assumed to be 206 persons per square mile, which is based on the Illinois population

census.

c

 1 person-rem = 0.01 person-Sv.
3–56

Table 3.2.16 Estimated Doses From Exposures to a Coal Ash Pile and 

Coal Ash Product a

Exposure Parameter

Annual Effective 

Dose Equivalent

(mrem)b

Collective Effective 

Dose Equivalent

(person-rem)c

WORKER

Direct irradiation

 Standing on the pile

 Standing near the pile 

33

14

---

Particulate emissions 0.007 ---

Radon emissions 0.005 ---

Total 47 ---

NEARBY RESIDENT

Direct irradiation 0.6 ---

Particulate emissions 0.02 ---

Radon emissions 0.003 ---

Groundwater 0 ---

Surface water 0.7 ---

Total 1.3 ---

REUSE

Direct irradiation

 House foundation 17 510

a

 EPA, Cont. No. 68–02–4375.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 Annual collective effective dose equivalent from reuse is estimated assuming that 4 people

live in a house and 7,500 houses are constructed with concrete using 1 year’s production of

coal ash; therefore, 30,000 persons could be exposed.
3–57

Table 3.2.17 Gamma Irradiation Dose of Individuals and Populations of Pocatello 

and Soda Springs, ID a

Community

Average Annual 

Individual Dose

(mrem/yr)b

Maximum Annual 

Individual Dose

(mrem/yr)b

Annual Population

Dose

(person-rem)b

Pocatello 14 150 800

Soda Springs 50 200 200

a

 EPA/520/6-90/008.

b

 1 mrem/yr = 0.037 mSv/yr; 1 person-rem = 0.01 person-Sv.

Note: Radionuclide concentrations in Idaho phosphate slag (EPA, RAE–9232/1–2): 1.5 Bq/g 

(41 pCi/g) 238U, 1.48 Bq/g (40 pCi/g) 234U, 1.8 Bq/g (48 pCi/g) 226Ra, and 0.02 Bq/g (0.5 pCi/g)

232Th.
3–58

Table 3.2.18 Individual and Collective Effective Dose Equivalents Attributed

to the Disposal of Zircon Sands a

Exposure Scenario

Individual Annual

 Effective Dose Equivalentb

(mrem)d

Annual Collective

Effective Dose Equivalentc

(person-rem)d

Waste disposal worker

 Collectore

 Landfill operator 

5

2

20

30

Off-site resident

 Airborne releases

 Groundwater releases 

 

0.04

0.02

20

90

Future on-site residentf 10 1×104

a

 Natural uranium concentration is assumed to be about 8.5 Bq/g (230 pCi/g) and the 232Th

concentration is 0.65 Bq/g (17.5 pCi/g). The short-lived decay products of 238U and 235U are

assumed to be present and in activity equilibrium. The decay products of 232Th are assumed to

be present and in activity equilibrium. See Appendix A.2 for a description of the generic

disposal methodology used in this analysis.

b

 For the individual annual effective dose equivalent (EDE), it is assumed that there is about

518 MBq (14 mCi) of natural uranium and of 74 Bq (2 mCi) of 232Th are disposed in one landfill. 

It is assumed that 60 metric tons of zircon sand are disposed annually in one landfill. See

Section 3.2.4.3 for further information.

c

 For the collective EDE, it is assumed that 1850 GBq (50 Ci) of natural uranium and 259 GBq

(7 Ci) of 232Th are disposed annually in 3,500 landfills. The exposed populations are as follows:

3,500 waste collectors, 17,500 landfill operators, 35,000 future residents, and 700,000 people

that drink groundwater. It is assumed 500,000 people reside within 80 km of each landfill.

d

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

e

 It is assumed the waste is transported to the landfill in an off-load vehicle. The volume

capacity is 15 m3

 and the weight capacity is 10 metric tons.

f

 The annual individual and collective dose equivalent for the future on-site resident takes into

account exposure to 220Rn and its short-lived decay products during indoor residence. 

However, 222Rn is not taken into account since 238U short-lived decay products only are

considered.
3–59

Table 3.2.19 Summary of Effective Dose Equivalents According to Percent Weight

Activities and Receptors

Individual Annual

Effective Dose

Equivalent

(mrem)b

Collective Effective

Dose Equivalenta

(person-rem)b

Percent 

Weight 

TRANSPORT

Truck driver 6 5 0.05

ROUTINE USEc,d

Worker

 Zircon flour handling

 Sandblasting

Public

 Dental products

 Zircon-derived products

 - Ophthalmic glass

 - Tape dispenser

 Phosphate slag

 - Pavement and building

 construction

300–4,000

300

0.05

0.4

0.5

10–200

300–700

--

50

54,000

5

2,000–8,000

0.05

0.006

0.05

0.05

0.05

0.009

LANDFILL DISPOSAL—ZIRCON SANDe

Workers

 Collectors

 Landfill operators

Public

 Off-site resident

 - Airborne releases

 - Groundwater releases

 Future on-site resident

5

2

0.04

0.02

10

20

30

20

90

10,000

0.05

0.05

0.05

0.05

0.05

ACCIDENTf

Dental products

 Warehouse

 Dental technician

Zircon flour

 Warehouse fire

 - Firefighter

 - Cleanup worker

 Transportation fire

 - Firefighter

 - Cleanup worker

6

0.002

0.4

3

3

30

0.05

0.05

0.05

0.05

0.05

0.05

See following page for footnotes.
3–60

Footnotes to Table 3.2.19

a

 The population sizes used in estimation of the collective effective dose equivalent (EDE) are

as follows: 1,000 drivers, zircon flour handlers, and tape dispenser users were assumed. The

population assumed for dental products containing uranium was 1 million, and 12 million was

assumed for ophthalmic glass wearers. The collective EDE from phosphate slag use was

based on the population of Soda Springs and Pocatello, ID.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 In zircon flour and sand, the 238U concentration is estimated to be 8.5 Bq/g (230 pCi/g) and the

232Th concentration is estimated to be 0.65 Bq/g (17.5 pCi/g) , whereas in ilmenite, the abrasive

used in sand blasting, the 238U concentration is 93 Bq/g (2.5 pCi/g) and the 232Th concentration

is 0.2 Bq/g (5.5 pCi/g). Note that the respirable dust concentration for the zircon product

handling was 1.4 mg/m3

, whereas the average dust concentration for sand blasting was

10.4mg/m3

 and 90% of the dust was considered to be respirable particulate.

d

 Refer to Section 3.2.4.2 for further information on the estimation of annual individual and

collective dose equivalent attributed to routine operations.

e

 Refer to Section 3.2.4.3 for further information on the estimation of annual individual and

collective dose equivalent attributed to disposal. Also see Appendix A.2 for a description of the

generic disposal methodology.

f

 Refer to Section 3.2.4.4 for further information on the estimation of the individual dose

equivalent attributed to accidents. Also see Appendix A.1 for a description of the generic

accident methodology.
3–61

3.3 Unrefined and Unprocessed Ore Containing Source Material

3.3.1 Introduction

In 10 CFR 40.13 (b), any person is exempt from the requirements for a license, to the extent

that the person receives, possesses, uses, or transfers unrefined and unprocessed ore

containing source material. Unless authorized in a specific license, a person shall not refine or

process such ore. This exemption was proposed on September 7, 1960 (25 FR 8619), and

issued as a final rule on January 14, 1961 (26 FR 284). 

The Federal Register notices cited above do not contain any information on analyses of

radiological impacts on the miner or public from transfer of unrefined and unprocessed ore

containing source material. No indication was given that a dose analysis had been performed

to support the exemption.

Since source material pertains to both uranium and thorium, the transport and possession of

uranium and thorium ore are evaluated in the current assessment. This exemption may also

apply to the transfer and possession of other ores containing source material.

3.3.2 Description of Items

Ore is defined in the Bureau of Mines Dictionary as a natural mineral compound of the elements

of which at least one is a metal (Bureau of Mines, 1968). In 10 CFR 40.4, unrefined and

unprocessed ore is defined as ore in its natural form prior to any processing, such as grinding,

roasting or beneficiating, or refining.” Based on these definitions, this exemption applies to an

ore that contains a metal that has not been either physically or chemically altered.

The NRC licensing requirement for source material begins after removal from its place of

deposit in nature” (10 CFR 40.3). Therefore, NRC does not regulate mining of ores regardless

of their source material content. Ores containing <0.05% by weight uranium, thorium or any

combination thereof are not source material by definition (40 CFR 40.4). The unrefined and

unprocessed ore exemption, therefore, applies to such ore subsequent to the mining activity. 

Thus, abandoned mines are not subject to regulation by the NRC. Also, the exemption does

not apply to in-situ mining or uranium because the ore is processed” through chemical

treatment underground.

The exemption explicitly pertains to the receipt, possession, use, or transfer of unrefined ore

that contains source material. In the notice of proposed rulemaking (25 FR 8619), it was stated

that prior to the proposed regulation, miners are required to have a license to transfer source

material after it was mined. Under the proposed regulation . . . the possession and transfer of

unrefined and unprocessed ores containing source material would be exempted.” Based on

this text in the Federal Register (25 FR 8619) and discussions with knowledgeable individuals,

the original application of this exemption was for possession of uranium ore by miners and for

transfer to an ore-buying station or directly to the milling facility (Phone call, K. Weaver,

Radiation Programs, Colorado State Health Department, Denver, CO, October 1996; phone

call, C. Cain, Region IV, Nuclear Regulatory Commission (NRC), Arlington, TX, October 1996). 

After the ore is mined at either surface or underground mines, it may be stored on-site and then

transported to an ore-buying station or taken directly to a milling facility.
3–62

Once the ore is delivered to a buying station or mill, it is considered to be an integral part of the

milling operation and subject to controls to meet the NRC license requirements, including

10 CFR Part 20.

Currently, no underground or surface uranium mines are producing ore (Phone call, K.

Sweeney, National Mining Association, Washington DC, July 1997). Some of the open-pit

mines are being held on standby status, while others have closed permanently and either been

reclaimed or abandoned. The dominant type of uranium mining method is in situ mining, which

does not require the transport of ore from the mine to the mill. There are three operating in situ

mines, and it is projected there will be four additional sites in fiscal year 1998 (Phone call, D.

Gillan, Uranium Recovery Branch, NRC, Rockville, MD, July 1997). In situ mining involves the

injection of a leaching solution (lixiviant) into the uranium-bearing strata to extract uranium. In

this process, the uranium is extracted from the lixiviant by ion-exchange. The ion-exchange

resin is stripped of uranium, which is then precipitated to produce a yellow cake slurry.

Monazite, a rare earth and thorium phosphate mineral, is the primary source of thorium

(Hedrick, 1997). Domestic mine production of thorium-bearing monazite ceased at the end of

1994 as the world demand for thorium-bearing ores remained depressed (Hedrick, 1997). In

prior years, monazite had been recovered by dredging methods. Thorium production was a

byproduct during the processing for titanium and zirconium minerals and monazite was

recovered for its rare earth content (Hedrick, 1997). Mined mineral sands contain very low

concentrations of thorium, ranging from 0.0012 to 0.005% by weight. After the initial wet gravity

separation process, which occurs at the mining site, the mineral sand concentrate usually

contains less than 0.05% by weight of thorium (Hewson, 1990). The mineral sands are further

processed via physical processes (e.g., gravity and electromagnetic processes) to produce a

monazite product that ranges from 4 to 6% by weight of thorium. 

3.3.3 Summary of Previous Analyses and Assessments

The Federal Register notices cited above do not contain any information on analyses of

radiological impacts on the public from the use of unrefined and unprocessed ore containing

source material. Since the exemption was established, a number of studies have described the

emissions and doses associated with uranium mining and milling operations. In these studies,

off-site air concentrations and potential doses due to ore transport and storage activities were

usually grouped with other mine or mill operations. A number of studies have also been

conducted on worker exposures associated with the mineral sands processing industry in which

monazite is a byproduct. As with uranium mining and milling, workers involved explicitly in the

transport and storage of mineral sands are not identified and are grouped with other workers. 

However, relevant information from these reports is summarized below.

3.3.3.1 Uranium Mining and Milling Operations

In NUREG–0706, one of the accident scenarios evaluated was a truck accident transporting

uranium ore from the mine to the mill. It is assumed that the uranium ore is shipped to ore

stockpiles in 23-metric ton (MT) batches and the average distance from the mine to the mill is

50 km. The ore was assumed to contain 1% respirable dust by weight. In the accident

scenario, all of the dust was released and available for dispersal. An environmental release

factor of 9×103

 was used. Based on these assumptions, the quantity of respirable ore
3–63

released to the environment in the event of a truck accident was estimated to be 2.1 kilogram

(kg). The individual lung dose commitment at 500 meters was 1.3 mSv (130 mrem), and

0.14 mSv (14 mrem) at 2000 meters from the accident scene (NUREG–0706). 

The National Park Service abandoned mine land inventory shows 42 abandoned uranium mine

sites within units of the National Park System and 2 immediately adjacent to park boundaries

(Phone call, J. E. Burghardt, Geologic Resources Division, National Park Service, Lakewood,

CO, July 1997). The Orphan uranium mine, which produced high-grade uranium ore, is now

abandoned and its surface facilities are along the West Rim Trail near the Grand Canyon

Village (Burghardt, 1995). The surface facilities are connected to the major underground mine

by a 488-meter compartment vertical shaft. The shaft’s headframe, cages, hoists, as well as air

compressor building, numerous building foundations, and miscellaneous scrap, remain in the

surface yard (Burghardt, 1995). This yard is fenced on three sides and is open on its northeast

side abutting the Grand Canyon’s South Rim. The surface yard is rectangular in shape,

measuring approximately 131 m × 102 m.

Several gamma-radiation survey measurements have been conducted on the Orphan mine

surface yard, but only two preliminary surveys have been conducted outside of the fenced

enclosure. A gamma survey conducted by the Bureau of Land Management detected elevated

gamma values beyond the fence line (Burghardt, 1996). Elevated gamma levels were also

detected by a survey conducted by Burghardt (1995). Background gamma-exposure rate

measurements ranging from 8 to 10 nanocoulomb (nC)/kg-h (30 to 40 microroentgen (R)/h)

were detected within 15 to 30 meters east and south of the fenced enclosure. However, to the

west of the fenced enclosure, exposure rates ranged between 13 and 98 nC/kg-h (50 and

380 R/h) (gross exposure rate) and some quite elevated gamma levels (77–98 nC/kg-h

(300–380 R/h)) were detected in the immediate vicinity of the West Rim Trail. The highest

value measured was 320 nC/kg-h (1250 R/h), which was found 15 meters northwest (outside)

of the fenced enclosure, just 30 meters from the West Rim Trail. According to Burghardt

(1995), mine operations at some point must have extended beyond the present-day fenced

yard.

3.3.3.2 Monazite 

Numerous studies focus on the radiological impacts from the dry separation processing of

mineral sands, especially for workers involved in monazite processing. In studies of radiation

doses to Western Australian mineral sands industry workers involved in mining, wet

concentration (initial separation process), and transport were considered to be nondesignated

employees (Marshman and Hewson, 1994). Nondesignated workers are defined as those

employees working in situations who are estimated to receive annual effective dose equivalents

(EDEs) substantially less than 5 mSv (500 mrem) (Hewson, 1990). The focus of these studies

was on annual committed EDEs associated with designated workers rather than on

nondesignated workers.

In the mineral sands mined in Western Australia, the concentration of thorium in the mineral

sands or ore” and in the wet separation (primary) concentrate is estimated to be 0.005% by

weight of thorium and 0.03% by weight of thorium, respectively (Hewson, 1990). Typical

absorbed dose rates associated with extraction of the mineral sands range from 0.1 to

0.3 microgray (Gy)/h (10 to 30 rad/h). For the wet separation process, absorbed dose rates

range from 0.2 to 1 Gy/h (20 to 100 rad/h) (Hewson and Hartley, 1990). Airborne gross
3–64

alpha activity levels associated with mining and wet separation activities were less than

0.01 Bq/m3

(< 0.3 pCi/m3

) and 0.02 Bq/m3 (0.5 pCi/m3), respectively (Hewson and Hartley, 1990).

The major exposure pathways associated with mineral sands processing are inhalation of dust

and external radiation, with only small contributions from radon, thoron, and their daughter

products. Activity median aerodynamic diameters (AMADs) of airborne dusts in the dry

separation plants range from 2 to 12 m, with an overall average value of about 6 m (Mason

et al., 1988). Investigation into the mineralogy of airborne dust has found that monazite

concentrates preferentially in the dust (Hartley and Hewson, 1993). In areas where monazite

(6 to 7% by weight of thorium) product is stored, absorbed dose rates may reach 100 Gy/h

(10 mrad/h) or more, depending on the exposure geometry (Mason et al., 1988). In a monazite

storage area containing bulk monazite, the thoron progeny concentration near the breathing

zone (1.8 meters from the ground) was measured at 15 mWL (Mason et al., 1988). This is

nearly two orders of magnitude smaller than the derived air concentration for thoron progeny,

which is 1200 WL (Mason et al., 1988). 

3.3.4 Present Exemption Analysis

For the distribution and transportation scenario, doses associated with the transport of uranium

and thorium ore from the mine to the mill are calculated. The uranium and thorium ore

concentrations used for the transport and ore storage exposure scenarios are 0.2% by weight

of U3O8 (21 Bq/g 238U (560 pCi/g)) in equilibrium with its progeny and 0.03% by weight of 232Th

(1.4 Bq/g (37 pCi/g)) in equilibrium with its progeny, respectively. Off-site routine exposure

scenarios evaluated are possession of uranium and thorium ore samples by geologists and the

exhibition of ore samples by museums. For these scenarios, the uranium ore concentration

was increased to 1% by weight of U3O8 (104 Bq/g 238U (2800 pCi/g)) in equilibrium with its

short-lived progeny, since it is likely that geologists may be interested in high-grade mineral

products. 

Potential doses to park rangers and hikers walking by abandoned uranium mines on public

lands are also evaluated. In this scenario, the uranium ore concentration was assumed to be

0.2% by weight. The primary exposure pathway for ore transport and storage are airborne

particulates, direct irradiation, and to a lesser extent, radon. The primary exposure pathways

for possession of an ore sample in a geologist’s office and on display in a museum is direct

irradiation. Potential exposure pathways to park rangers and hikers from abandoned mines are

airborne particulates and direct irradiation. An accident involving the transport of uranium and

thorium ore from the mine to the mill is also evaluated.

3.3.4.1 Distribution and Transport

3.3.4.1.1 Uranium Ore Transport

The transport of uranium ore to uranium mills has been drastically reduced over the years. 

Since there are no underground or surface uranium mines currently producing ore, the following

assessment is hypothetical. In NUREG–0706, it was stated that uranium mills have the

capacity to process 8700 MT of ore per day, and typically maintain a 10-day backup supply of

on-site ore (NUREG–0706). However, the processing of about 270 and up to 2000 MT of ore
3–65

per day would be more likely (Phone call, J. Cain, Environmental Coordinator and Radiation

Safety Officer, Cotter Corporation, Canyon City, CO, October 1996; phone call, S. Landau,

Cotter Corporation, Lakewood, CO, July 1997; phone call, O. Paulson, Facility Supervisor,

Kennecott Energy, Rawlins, WY, July 1997). In the current assessment, it is assumed that

907 MT of ore (0.2% by weight of U3O8 (21 Bq/g 238U (560 pCi/g))) would be transported from a

uranium mine (160 km on public roads one way) to an operating mill per day and each

ore-hauling truck would have a 21-MT capacity. Under these circumstances, there could be

about 43 trucks per day or about 10,750 truck loads of ore transported annually from uranium

mines to a mill (250 day/yr). Under this assumption, one driver would transport two loads of ore

per day, resulting in about 500 trips per year. Ore trucks are covered and the ore is wetted

prior to transport. Primary exposure routes for truck drivers are direct irradiation and inhalation

during ore pickup and drop operations.

MicroShield (Computer Codes, Grove Engineering, 1996) was used to calculate the EDE rate

due to direct irradiation. The dump body dimensions used were 5 m × 2.3 m, the thickness of

the dump body was 0.3 cm iron, and the source-to-receptor distance was 1 m. The exposure

duration per trip was 2 hours (one way), which results in an approximate annual exposure

duration of 1000 hours in which a full load of ore is transported to the mill. Given these

conditions, the annual EDE is estimated to be 2 mSv (200 mrem). 

Fugitive dust emissions result when the ore truck driver picks up or drops a load of ore. A

respirable (<15 m) particulate emission factor of 0.005 kg/MT of uranium ore was used to

estimate the amount of airborne particulate from dropping operations (EPA, Aggregate

Handling, 1988). This particulate emission factor takes into account a wetting efficiency of 50%

and a mean wind speed of 5 m/s (Kennecott, 1994). Using a box model (Hanna et al., 1982)

and assuming an ore dropping area of 30 m2

 and a mixing height of 10 m, the estimated

respirable dust concentration is 900 g/m3

. Assuming that a driver is standing outside of the

truck cab and is exposed about 170 h/yr (10 minutes per pickup and 10 minutes per dropoff),

the estimated annual EDE due to dust inhalation is 0.6 mSv (60 mrem). In many cases, the

truck driver would not stand outside of the truck cab or the ore could be discharged into a

hopper underneath the truck (bottom load truck) and the driver would remain inside the truck

cab. The EDE due to radon is assumed to be negligible, since buildup of radon progeny would

be minimal in the highly ventilated space in a moving truck. The annual EDE to a truck driver

from direct irradiation and dust inhalation is estimated to be 3 mSv (300 mrem).

3.3.4.1.2 Thorium Ore Transport

In the original mineral sand, the thorium content is about 0.0013% by weight (13 ppm). When

the dredging method is used to mine mineral sands, a simultaneous gravity separation process

is used to separate the silica sand from the minerals. The discarded sand (about 97% of the

material excavated) is pumped back into the mining area. Three percent of the remaining

minerals are also separated on-site into mineral components. Of the remaining minerals

composite, about 0.6 to 0.8% is monazite and about 4.2% of the monazite is thorium (Phone

call, J. Raiser, RGC Minerals and Sands, Inc., Green Cove Springs, FL, July 1997). This

results in a thorium concentration of about 340 ppm (0.034% by weight, 1.4 Bq/g) (37 pCi/g) in

the initial gravity separated mineral ore concentrate. This is the same thorium concentration

value given by Hewson (1990). In 10 CFR 40.4, unprocessed, unrefined ore is defined as ore

in its natural form prior to any processing such as grinding, roasting or beneficiating, or

refining.” Due to the type of mineral sand extraction process, in which there is no change in
3–66

chemical or crystal matrix, it is assumed that the mineral product after this initial gravity

separation step is unprocessed, unrefined ore.

Although domestic mine production of thorium-bearing monazite ceased in 1994, for this

analysis it was assumed that 907 MT of monazite is processed per day. Twenty-one

MT-hauling-capacity trucks would be used to transport the mineral sand to the processing plant,

but the assumed distance to the plant is about 16 km (a one-way trip would take 0.5 hour at

32 km/h). Therefore, an exposure duration of about 1000 hours was assumed (one driver could

make about 5 trips per day, which results in about 1250 trips per year). Using the same dump

truck body dimensions and source-to-receptor distance for the driver as used in Section

3.3.4.1.1, the resulting annual EDE due to direct irradiation is about 0.2 mSv (20 mrem). Since

the mineral sands separation process is wet (and using a mean wind speed of 2 m/s), the

estimated respirable airborne particulate emission factor for the ore drop operation was

0.0007 kg/MT. Using the box model (Hanna et al., 1982) and the same area source

dimensions and mixing height as used for uranium ore drop operations, the estimated airborne

respirable dust concentration was 400 g/m3

. Assuming an exposure duration of 420 h/yr

(10 minutes per ore pickup or drop operation), the estimated annual EDE is 0.14 mSv

(14 mrem). The annual individual EDE due to both direct irradiation and dust inhalation to a

truck driver hauling mineral sands is about 0.34 mSv (34 mrem). 

3.3.4.2 Routine Exposure

The exemption explicitly pertains to receipt, possession, use, or transfer of unrefined and

unprocessed ore. Transport of ore was evaluated in Section 3.3.4.1. The exposure scenarios

evaluated are the possession of uranium and thorium ore samples in an office and the viewing

of ore specimens in a museum exhibit. Another exposure scenario evaluated is park rangers

and hikers walking on trails are near abandoned uranium mines. A summary of the estimated

individual and collective annual EDEs associated with these exposure scenarios is provided in

Table 3.3.4.

3.3.4.2.1 Sample Possession

Geologists, as well as other types of professionals (e.g., mining engineers) may keep ore

samples within their offices or laboratories (Tanner, 1990). Organizations such as the U.S.

Geological Survey (USGS) also have ore cores and other geological samples in controlled ore

storage areas (Phone call, R. Zielinski, USGS, Lakewood, CO, July 1997). Ore samples are

not usually kept in the desk, but rather within the office space (Phone call, R. Zielinski, USGS,

Lakewood, CO, July 1997). Assuming that a uranium ore sample (1% weight of U3O8, 111 Bq/g

(0.003 Ci/g) 238U) 15.2 cm in diameter was kept in an office 1.8 m from the worker for

2000 h/yr, the annual EDE due to direct irradiation is about 4×105

 mSv (0.004 mrem). If a

monazite sample (0.03% by weight of thorium) 15.2 cm in diameter was kept in an office 1.8 m

from the worker for 2000 h/yr, the annual EDE due to direct irradiation is about 9×105

 mSv

(0.009 mrem). It is unknown how many professionals (geologists or mining engineers) may

have ore samples in their possession (e.g., office or laboratory). However, if there are 1,000

individuals who have uranium and thorium ore samples, the annual collective EDEs were

estimated to be 4×105

 person-Sv and 9×105 person-Sv (0.004 person-rem and

0.009 person-rem), respectively. 
3–67

Extending this scenario to an amateur geologist with 100 uranium ore samples and 100

monazite samples on display in his home office, again with an exposure time of 2000 h/yr, the

annual EDE due to direct irradiation is 0.004 mSv (0.4 mrem) for the uranium ore 0.009 mSv

(0.9 mrem) for the thorium ore, or less than 0.01 mSv (<1 mrem) in both cases. However, if

there are 100 amateur geologists who have uranium and thorium ore samples in collections of

this size, the annual collective EDEs are estimated to be 4×104

 person-Sv and

9×104

 person-Sv (0.04 person-rem and 0.09 person-rem), respectively.

Numerous ore samples are on display in U.S. museums. In the current assessment it is

assumed that a uranium (1% by weight of U3O8) and monazite (0.03% by weight of thorium) ore

sample (5-cm diameter) are in a display case and the museum visitor stands 0.6 m from the

display for about 10 minutes. Based on these conditions, the estimated annual EDEs from

viewing these uranium and monazite samples are less than 1×105 mSv (<0.001 mrem) for

each, respectively. It is unknown how many visitors view mineral exhibits in museums. If

1 million people view uranium and thorium ore specimens in museums annually, the annual

collective dose equivalents would be less than 1×105

 person-Sv (<0.001 person-rem) for each

ore type.

3.3.4.2.2 Abandoned Uranium Mines

The National Park Service abandoned mine land inventory shows 42 abandoned uranium mine

sites within units of the National Park System and 2 immediately adjacent to park boundaries

(Phone call, J. E. Burghardt, Geologic Resources Division, National Park Service, Lakewood,

CO, July 1997). The majority of these sites are in the southern Utah parks of Canyonlands,

Capitol Reef, and Glen Canyon (Phone call, J. E. Burghardt, Geologic Resources Division,

National Park Service, Lakewood, CO, July 1997). There are numerous other abandoned

mines on the Colorado Plateau, a physiographic province centered on the four corners” area of

Utah, Colorado, New Mexico, and Arizona (Burghardt, 1996). There are many more

abandoned uranium mines on other Federal lands (Phone call, J. E. Burghardt, Geologic

Resources Division, National Park Service, Lakewood, CO, July 1997). In many cases,

abandoned uranium mines are in remote areas that are not easily accessible by the public. 

However, some abandoned uranium mines may be accessible to the public.

In the current assessment, a hypothetical abandoned uranium mine, located in a national park

in an area accessible by hikers, is evaluated. It is assumed park rangers lead groups of hikers

along a trail near an abandoned uranium mine. The trail is on land contaminated with residual

uranium ore (from an ore storage pad and truck loading area) and mine spoils. It is assumed

that due to weathering, the uranium contamination extends 5 cm into the soil. The uranium ore

and mine spoils contain 0.2% by weight of U3O8 and are spread over a number of acres. 

Dose coefficients for exposure to soil contaminated to a depth of 5 cm from Federal Regulatory

Guidance Report No. 12 (EPA–402–R–93–081) were used to estimate annual EDEs due to

direct irradiation. The exposure duration for the park ranger leading the hikes is 100 h/yr. For

an individual hiker, the exposure duration is 2 h/yr. Based on these assumptions, the annual

EDE for the park ranger and hiker due to direct irradiation is estimated to be 0.008 mSv and

2×104 mSv (0.8 mrem and 0.02 mrem), respectively. To estimate the annual EDE due to dust

inhalation from wind erosion, a mass loading factor of 100 g/m3 is used (Healy, 1980). The

annual individual EDE for park ranger and hiker due to dust inhalation would be about 0.04 mSv

(4 mrem) and 8×104 mSv (0.08 mrem), respectively. Doses due to radon emanation at the
3–68

abandoned uranium mine site are considered to be negligible. Therefore, the annual EDEs due

to direct irradiation and dust inhalation for a hypothetical park ranger and hiker are about

0.05 mSv (5 mrem) and 0.001 mSv (0.1 mrem), respectively.

It is unknown how many park rangers and hikers may walk by abandoned uranium mines on

public lands. However, if 100 park rangers and 100,000 hikers have access to abandoned

uranium mine property, the annual collective EDEs are 0.005 person-Sv (0.5 person-rem) and

0.1 person-Sv (10 person-rem), respectively.

3.3.4.3 Disposal

Uranium ore, once mined, is typically processed into a commercial product. Rather than be

disposed, it may be stockpiled to blend with other ores or be maintained until economic

conditions allow for cost-effective ore processing. Large inventories of ore cores and samples,

such as those maintained by the USGS, are being sent to a uranium mill for processing (Phone

call, D. DePlato, Radiation Safety Officer, RGC Minerals and Sands, Green Cove Springs, FL,

July 1997), and it is very unlikely that residual ore and mine spoils at an abandoned uranium

mine would be sent to at an off-site disposal site (Phone call, J. E. Burghardt, Geologic

Resources Division, National Park Service, Lakewood, CO, July 1997). At one facility that

commercially produced thorium, monazite is being reintroduced into the mineral sand mining

area (Phone call, W. Cofer, Health Physicist, State of Florida Department of Health and

Rehabilitative Services, Orlando, FL, July 1997).

The most likely scenario in which uranium and thorium ore may be disposed in municipal

landfills is the disposal of discarded small uranium and thorium ore samples. It is assumed that

3500 uranium (0.2% by weight of U3O8) and thorium (0.03% by weight of thorium) 1-kg ore

samples are disposed in 1 year. Summarized in Tables 3.3.2 and 3.3.3 are the individual and

collective EDEs due to disposal of ore samples, respectively. These EDEs are estimated using

the generic disposal methodology described in Appendix A.2.

3.3.4.4 Accidents

For this analysis, it is assumed that an accident occurs during transport, resulting in fire. It is

difficult to imagine a fire causing a significant fraction of the ore to become airborne; however,

firefighting activities could disperse the material and cleanup workers could encounter

contamination. Thus, this should be considered as an upper bounding scenario. It is assumed

that 21 MT of uranium (0.2% by weight of U3O8) and thorium (0.03% by weight of thorium) ore

is being transported. The generic accident methodology described in Appendix A.1 is used for

this analysis. In a transport accident involving uranium ore, the estimated EDE due to

inhalation and submersion to a firefighter is 0.03 mSv (3 mrem). The estimated EDE to a

cleanup worker due to particulate resuspension is 0.3 mSv (30 mrem); however, no respiratory

equipment is assumed for the cleanup worker. It is very likely that a cleanup worker for a truck

load of uranium ore would wear a respirator. If a dust respirator with an assigned protection

factor (APF) of 10 was used, the EDE would be reduced to 0.03 mSv (3 mrem).

In the transportation accident involving thorium ore, the estimated EDE due to inhalation and

submersion to a firefighter is 0.01 mSv (1 mrem). The estimated EDE to a cleanup worker due

to particulate resuspension is 0.1 mSv (10 mrem); however, no respiratory equipment is

assumed for the cleanup worker. As mentioned for the transportation accident involving
3–69

uranium ore, cleanup workers would also wear some type of respiratory equipment for

accidents involving thorium ore. Even if a dust respirator with an APF of 10 was used, the EDE

would be reduced to 0.001 mSv (0.1 mrem).

3.3.5 Summary 

It appears that the original application of this exemption was for possession of uranium ore by

miners and the transfer of the ore to an ore-buying station or directly to the milling facility. 

Based on the wording in 10 CFR 40.13 (b), a person would be exempt from the requirements of

a license if that person receives, possesses, uses, or transfers any type of unrefined and

unprocessed ore containing source material. Summarized in Table 3.3.4 are the results of the

current dose assessment, which evaluated the transfer, receipt, possession, and disposal of

uranium and thorium ore. In addition, the radiological impact of abandoned uranium mines

located on Federal lands that may be accessible to members of the public is evaluated. 

Based on the current analysis, the following general conclusions were obtained concerning

radiological impacts on truck drivers, and other members of the public who may come in contact

with ore samples or residual ore contamination at abandoned uranium mines:

Hypothetical annual EDEs to ore truck drivers appear to be about 0.3 mSv (300 mrem).

Doses to members of the public from ore samples in museum exhibits are very low. 

However, there is a potential of higher doses to members of the public from

contamination at abandoned uranium mines that may be accessible to the public.
3–70

Table 3.3.1 Individual Doses to Workers and the Public From Unrefined Uranium 

and Thorium Ore

 

Receptors

Uranium Ore (0.2% by weight)

Individual Annual Effective

Dose Equivalent

(mrem)a

Thorium Ore (0.03% by weight)

Individual Annual Effective

Dose Equivalent

(mrem)a

Truck driversb 300 34

Geologist with ore

 samplec

0.004 0.009

Museum visitorc <0.001 <0.001

Park rangerd 5 ----

Hikerd 0.1 ----

a

 1 mrem = 0.01 mSv.

b

 See Section 3.3.4.1.1 for further information on the parameters used to estimate the annual

individual and collective effective dose equivalents (EDEs) for ore-hauling truck drivers.

c

 See Section 3.3.4.2.1 for further information on the parameters used to estimate the annual

individual EDEs for the individual with ore samples or for museum visitors viewing a mineral

exhibit.

d

 See Section 3.3.4.2.2 for further information on the parameters used to estimate the annual

individual EDEs for park rangers and hikers walking on abandoned uranium mine property

located on public lands.
3–71

Table 3.3.2 Individual Doses Attributed to Disposal of Uranium and 

Thorium Ore Samples

 

Receptors

Uranium Ore (0.2% by weight)

Individual Annual Effective

Dose Equivalent

(mrem)a

Thorium Ore (0.03% by weight)

Individual Annual Effective

Dose Equivalent

(mrem)a

Waste collectors <0.001 <0.001

Landfill operators <0.001 <0.001

Off-site receptor

 Airborne releases

 Groundwater

 

<0.001

<0.001

<0.001

<0.001

On-site receptor <0.001 <0.001

a

 1 mrem = 0.01 mSv.
3–72

Table 3.3.3 Collective Doses Attributed to Disposal of Uranium and 

Thorium Ore Samples

 

Receptors

Uranium Ore (0.2% by weight)

Collective Effective Dose

Equivalent

(person-rem)a

Thorium Ore (0.03% by weight)

Collective Effective Dose

Equivalent

(person-rem)a

Waste collectors 0.001 0.001

Landfill operators <0.001 <0.001

Off-site receptor

 Airborne releases

 Groundwater 

<0.001

<0.001

<0.001

On-site receptor <0.001 0.006

a

 1 person-rem = 0.01 person-Sv.
3–73

Table 3.3.4 Summary of Doses Associated With Unrefined Uranium and 

Thorium Ore

Individual Annual

Effective Dose Equivalent

(mrem)b

Collective

Effective Dose Equivalenta

(person-rem)b

Activities and

Receptors

Uranium Ore Thorium Ore Uranium Ore Thorium Ore

TRANSPORTc

Truck driver 300 34

ROUTINE USEd

Worker exposures

 Geologist with sample

 Park ranger

0.004

5

 0.009 0.004

0.5

0.009

Public exposure

 Museum visitor

 Hiker 

 Amateur geologist

<0.001

0.1

 0.4

<0.001

 0.9

<0.001

10

 0.04

<0.001

 0.09

LANDFILL DISPOSALe

Occupational

 Collectors

 Landfill operators

 

Public exposure

 Off-site resident

 - Airborne releases

 - Groundwater releases

 

Future on-site resident

<0.001

<0.001

 

 

<0.001

<0.001

0.001

<0.001

<0.001

 

 

<0.001

<0.001

<0.001

0.001

<0.001

 

 

<0.001

0.004

<0.001

0.001

<0.001

 

 

<0.001

0.006

ACCIDENTf

Transportation fire

 Firefighter

 Cleanup worker

3

30

1

10

a

 Refer to text discussion for time period for collective dose assessment.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 Refer to Section 3.3.4.1 for further information on the estimation of annual individual and

collective dose equivalents attributed to transportation operations.

d

 Refer to Section 3.3.4.2 for further information on the estimation of annual individual and

collective dose equivalents attributed to routine operations.

e

 Refer to Section 3.3.4.3 for further information on the estimation of annual individual and

collective dose equivalents attributed to disposal. Also see Appendix A.2 for a description of

the generic disposal methodology.

f

 Refer to Section 3.3.4.4 for further information on the estimation of the individual dose

equivalent attributed to accidents. Also see Appendix A.1 for a description of the generic

accident methodology.

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3.4 Incandescent Gas Mantles

3.4.1 Introduction

In 10 CFR 40.13(c)(1)(i), any person is exempt from the requirements for a license to the extent

that the person receives, possesses, uses, or transfers any quantity of thorium contained in

incandescent gas mantles. This exemption was established on March 20, 1947 (12 FR 1855),

and has remained essentially unchanged since that time.

The Federal Register notice cited above provided no information on radiological impacts on the

public for use or disposal of gas mantles containing thorium. Information published in 1960 by

the Atomic Energy Commission (25 FR 8619) indicated that the exemption would not result in

an unreasonable hazard to life or property, but a supporting dose analysis was not published.

Estimates of potential radiation doses to members of the public from incandescent gas mantles

containing thorium have been published in other reports by Buckley et al. (NUREG/CR–1775),

O’Donnell and Etnier (NUREG/CR–1910), and the National Council on Radiation Protection and

Measurements (NCRP 95). These dose estimates are discussed in Section 3.4.3 of this report.

3.4.2 Description of Items

Thorium-containing mantles are available in a variety of designs and sizes, each intended to fit

into one of the many different lighting devices in use. To function, the mantle must be heated

to a temperature of 1870 to 2370C, which causes the thorium oxide in the mantle to

incandesce (NUREG/CR–1910). This is achieved by placing the mantle over, in, or near a gas

or kerosene flame that burns during the operation of the device.

Incandescent gas mantles containing thorium have been used for indoor and outdoor lighting

since the 1880s. In 1950, gas mantles were still being used to light the streets of some

communities (NUREG/CR–1775), and in 1952, 65% of the U.S. production of thorium was used

to make gas mantles (Cuthbert, 1958). In 1974, the U.S. production of gas mantles was

estimated to be 20 million units per year (NUREG/CP–0001, Barker and Tse), and in 1981, U.S.

domestic use of gas mantles was estimated to be 25 million units per year (NUREG/CR–1910). 

At present, U.S. sales of gas mantles are estimated to be 50 million units per year (Phone call,

R. May, The Coleman Company, Inc., Wichita, KS, April 1997). The Coleman Company, a

major U.S. supplier of gas mantles, recently replaced thorium in its gas mantles with

(nonradioactive) yttrium (Couch and Vaughn, 1995). However, thorium-containing mantles are

still available in the United States under a variety of brand names (Couch and Vaughn, 1995),

and they account for about half of the current U.S. sales, or about 25 million units per year

(Phone call, R. May, The Coleman Company, Inc., Wichita, KS, April 1997).

Thorium–containing mantles are manufactured by soaking a webbed fabric (or stocking) in a

solution containing nitrates of thorium and other metals (NUREG/CR–1910). The other metals

are added either to improve the luminosity of the mantles (e.g., cerium) or to help harden the

fragile structure of the mantle (e.g., beryllium) (Griggs, 1973). When saturated, the fabric is

removed from the solution and dried. In some cases, the fabric is also treated to convert the

soluble nitrates into insoluble compounds (most likely hydroxides), then it is rinsed and dried

(NUREG/CR–1910). The fabric is finally coated with lacquer, dried, cut, and fashioned into
3–76

mantles. Most mantles, when flattened, consist of two layers of fabric that are 4 to 11 cm long

by 2 to 7 cm wide and contain between 50 and 500 mg of thorium (NUREG/CR–1910). At this

stage, the mantles can be finished as either soft” or hard” mantles.

To form a soft mantle, the stocking is cut off, sewn shut to form a bag, and either equipped with

a drawstring for tying it directly to the burner assembly or affixed to a holder that attaches to the

burner assembly (NUREG/CR–1910). Hard mantles are bags that are preshaped by

attachment to a rigid support frame that either attaches to or sits over the burner assembly. 

Some hard mantles are also preburned at the factory to convert the thorium to thorium oxide

(NUREG/CP–0001, Cullen and Paschoa). This procedure volatilizes the fabric bag and the

lacquer coating, leaving a fragile shell of thorium and other metal compounds. If the hard

mantles are not preburned or soft mantles are being used, they are placed over the burner

assembly in a lighting device, which may accommodate one to four mantles, then preburned

with no fuel being supplied to convert the thorium to thorium oxide (NUREG/CR–1910).

Finished mantles are packaged and distributed as ordinary consumer products

(NUREG/CR–1910). Hard mantles are normally boxed singly, and one or two soft mantles are

usually sealed in a plastic pack. In both cases, the box or pack may be placed in a carton

containing a new lighting device, or they may be combined with other boxes or packs in a

carton containing replacement mantles. Except for mantles to be installed in outdoor lighting

devices, which are normally purchased and installed by utility companies, most mantles and

new lighting devices are purchased from retail stores and installed in the lighting devices by

users. 

Generally, the lifetime of a gas mantle depends on its use environment (NUREG/CR–1910). In

stable environments (e.g., fixed outdoor and indoor lamps), mantles may function for 6 to

8 months before breaking. In less stable environments (e.g., moveable residential lamps), they

may function for 1 to 3 months. In unstable environments (e.g., camping lanterns), they may

function over a time span that includes 50 hours of actual burning. In all use environments,

however, the lifetime of a gas mantle is estimated to be less than 1 year.

3.4.3 Summary of Previous Analyses and Assessments

A variety of distribution, routine use, disposal, accident, and misuse exposure scenarios were

considered in the previous assessments by Buckley et al. (NUREG/CR–1775) and O’Donnell

and Etnier (NUREG/CR–1910). For distribution, routine use, and disposal, these scenarios

include the following:

Exposures to workers and members of the public during distribution, transport, and

installation.

Exposures to campers during use of portable lanterns and changing of mantles.

Exposures from residential and commercial outdoor lighting and residential indoor

lighting.

Exposures from lighting in recreational vehicles.
3–77

Exposures from disposal of mantles in landfills and by incineration.

The exposure scenarios involving accidents and misuse of gas mantles that have been

considered include:

Contamination of a campground following breakage of a mantle and transport of thorium

and its decay products to a nearby reservoir used to supply drinking water.

Ingestion of a mantle by a small child.

Exposure to firefighters to thorium and its decay products released in a warehouse fire.

The dose estimates of O’Donnell and Etnier (NUREG/CR–1910) were based on an annual

distribution, use, and disposal of 25 million mantles. Both individual and collective doses were

presented as ranges of possible doses because of uncertainties regarding (1) the actual

radionuclide content of the mantles, (2) the rate of diffusion of 220Rn from the mantles and

stored lanterns, (3) the identities and quantities of radionuclides that may evaporate from the

mantles during use, and (4) the behavior of persons during and after mantle replacement. To

account for these uncertainties, doses were calculated with the following assumptions:

The radionuclides initially present in the mantles had activities corresponding to their

activities in natural thorium or in thorium that was purified and aged for 6 months (see

Table 3.4.1). Each mantle was assumed to contain 250 mg of thorium, although the

thorium content of some mantles was as high as 400 mg (NUREG/CR–1775). The

activity of 232Th in such a mantle is about 1 kilobecquerel (kBq) (27 nanocurie (nCi)). 

The total activity in the mantle is about 10 kBq (270 nCi) if the mantle contains natural

thorium or 7.8 kBq (210 nCi) if it contains 6-month-old thorium.

During times the mantles are not burning, 25% of the activities in Table 3.4.1 for 220Rn

and its progeny were assumed to diffuse into the air. The air concentrations used in the

dose estimates depended upon the volumes and ventilation rates of the contaminated

air spaces. For mantles stored in lanterns in homes, doses were also calculated for the

limiting cases when (a) all of the 220Rn diffuses from the mantles and disperses

immediately in the home and (b) only 5% of the diffusing 220Rn enters the home air.

To bracket possible effects of nuclide vaporization during mantle use, dose estimates

were made for complete vaporization of (a) only 220Rn and its progeny and (b) all except

the two thorium radionuclides, 232Th and 228Th (see Table 3.4.1).

To estimate doses from changing mantles, the following exposure scenario was used:

A. Exposure to two new gas mantles

(1) Direct exposure. The installer is exposed for 1 minute at an average

distance of 1 meter from two mantles containing a total of 500 mg of

6-month-old thorium that is depleted in 220Rn and its decay products

because of diffusion (25%) from the thorium.
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(2) Inhalation exposure. Twenty-five percent of the 220Rn escapes into a

hemispherical air space with a volume of 7 m3

 and a ventilation rate of

either 1000 volume changes per hour (an outdoor replacement in a

3-kilometer per hour breeze) or 1 volume change per hour (an indoor

replacement).

B. Exposure to two old mantles

(1) Direct exposure: Same as above.

(2) Inhalation exposure.

(a) Same as above.

(b) The installer blows broken mantles from the lighting device and

1% (5 mg) of the ash from the broken mantles is dispersed as

respirable particles into the air space.

(3) Ingestion exposure. While handling the old mantles 5% (25 mg) of the

thorium adhered to the installer’s hands, and subsequent washing left

only 1% (0.25 mg) of the adhered material, which was ultimately

ingested.

Depending on which parts of the mantle replacement scenario were used, O’Donnell and Etnier

(NUREG/CR–1910) estimated total doses (the sum of the dose equivalents from external

exposures plus the 50-year dose commitments from inhalation and ingestion) for outdoor

replacement of two mantles that ranged from 0.04 nanosievert (nSv) (4 nrem) to 0.002 mSv

(0.2 mrem) to the total body, from 0.04 nSv (4 nrem) to 0.01 mSv (1 mrem) to the bone, and

from 0.2 nSv (20 nrem) to 0.4 Sv (40 rem) to the lungs. Indoor replacement was estimated

to give total doses between 0.8 nSv (80 nrem) and 0.007 mSv (0.7 mrem) to the total body,

between 0.003 Sv (0.3 rem) and 0.06 mSv (6 mrem) to the bone, and between 0.4 Sv

(40 rem) and 0.03 mSv (3 mrem) to the lungs.

During distribution, routine use, and disposal, O’Donnell and Etnier (NUREG/CR–1910)

estimated the dose from both external and internal exposure, as appropriate for each exposure

scenario. The best estimates of individual and collective doses, rather than the minimum or

maximum values, obtained by O’Donnell and Etnier are summarized as follows:

For workers during transportation and distribution and installation, annual dose

equivalents to individuals were 0.002 mSv (0.2 mrem) to the whole body, 0.004 mSv

(0.4 mrem) to the bone, and 0.002 mSv (0.2 mrem) to the lungs. Annual collective dose

equivalents were 0.6 person-Sv (60 person-rem) to the whole body, 1 person-Sv

(100 person-rem) to the bone, and 0.5 person-Sv (50 person-rem) to the lungs.

For routine users of gas mantles (including campers and their families), persons

exposed during indoor and outdoor residential use, and persons using recreational

vehicles, the annual dose equivalents to individuals were 0.001 mSv (0.1 mrem) to the

whole body, 0.007 mSv (0.7 mrem) to the bone, and 0.02 mSv (2 mrem) to the lungs. 

Annual collective dose equivalents were 40 person-Sv (4000 person-rem) to the whole
3–79

body, 200 person-Sv (20,000 person-rem) to the bone, and 500 person-Sv

(50,000 person-rem) to the lungs.

For other members of the public who do not use gas mantles but receive exposure while

shopping in stores or living along truck routes or near incinerators, annual dose

equivalents to individuals were 0.01 Sv (1 rem) to the whole body, bone, and lungs. 

Annual collective dose equivalents were 3 person-Sv (300 person-rem) to the whole

body and bone and 2 person-Sv (200 person-rem) to the lungs.

The best estimates of the radiation doses for the accident and misuse scenarios are

summarized as follows:

For contamination of a campground following breakage of a mantle and transport of

thorium and its decay products to a nearby reservoir used to supply drinking water,

O’Donnell and Etnier (NUREG/CR–1910) estimated annual dose equivalents of less

than 0.01 Sv (<1 rem) to the whole body, bone, and lungs, and annual collective dose

equivalents of less than 0.001 person-Sv (<0.1 person-rem) to the whole body, bone,

and lungs.

For ingestion of a whole mantle, O’Donnell and Etnier estimated dose equivalents of

individuals of 1 mSv (100 mrem) to the whole body, 7 mSv (700 mrem) to the bone, and

0.4 mSv (40 mrem) to the lungs.

For exposure to firefighters at a warehouse fire involving 10,000 gas mantles, and using

conservative assumptions for the amount of thorium released in a fire plus no

respiratory protection for firefighters, Buckley et al. (NUREG/CR–1775) estimated dose

equivalents to individuals of 3 mSv (0.3 rem) to the whole body, 70 mSv (7 rem) to the

bone, and 40 mSv (4 rem) to the lungs.

Both O’Donnell and Etnier (NUREG/CR–1910) and Buckley et al. (NUREG/CR–1775)

considered doses from disposal of gas mantles in a landfill. However, neither analysis appears

to be adequate because only the mobilization and transport of thorium and its decay products to

a nearby source of water was considered. Since the assumed scenario is quite conservative

and seems unlikely for such an immobile radionuclide as thorium, a more likely exposure

scenario involves inadvertent intrusion into solid waste in a landfill by future on-site residents,

but doses from this scenario were not evaluated.

The dose analyses for thorium gas mantles were summarized in NCRP 95 by expressing the

dose estimates in terms of effective dose equivalents (EDEs). Based on the results of

O’Donnell and Etnier (NUREG/CR–1910), the NCRP came to the following conclusions:

The average annual EDE to individuals would be about 0.002 mSv (0.2 mrem), with the

maximum dose to individuals using worst-case assumptions possibly being a factor of

100 higher, or 0.2 mSv (20 mrem).

The annual collective EDE from distribution, routine use, and disposal of 25 million gas

mantles per year would be about 90 person-Sv (9000 person-rem).
3–80

Thus, it can be concluded that the existing analyses of potential doses to the public from use of

thorium gas mantles are adequate, for the most part. However, three refinements to the

existing analyses appear to be needed. First, the EDEs should be calculated for exposure

scenarios involving inhalation and ingestion of thorium and its decay products. Second,

accidents involving fire should be reevaluated using the generic accident methodology of

Appendix A.1 of this report. Third, realistic intrusion scenarios following disposal of gas mantles

in a landfill should be evaluated in addition to more realistic scenarios for release of radioactivity

to the general environment using the generic disposal methodology of Appendix A.2 of this

report.

3.4.4 Present Exemption Assessment

Table 3.4.2 presents results of the current assessment of potential radiation doses to members

of the public from the annual distribution, use, and disposal of 25 million gas mantles containing

thorium. The current assessment relies very heavily on the previous work of O’Donnell and

Etnier (NUREG/CR–1910) and on a study of radioactivity released from burning gas lantern

mantles by Luetzelschwab and Googins (1984).

Based on the work of O’Donnell and Etnier (NUREG/CR–1910), it is assumed that 20.25 million

of the gas mantles are used in portable lanterns (e.g., camping lanterns), 3 million in

commercial and residential outdoor gaslights, and 1.25 million in indoor residential gas or

kerosene lamps. It is also assumed based on their previous work that (1) the amount of

thorium contained in the gas mantles is 250 mg, and (2) the thorium has decayed for 6 months

after being purified (see Table 3.4.1). The value of 250 mg of thorium per mantle is in close

agreement with results from a survey of 15 mantle packages (2 mantles per package) by

Luetzelschwab and Googins (1984). They found that the average 232Th activity per package

was 2.2 kBq (65 nCi), which corresponds to a mass of 600 mg of 232Th per package (or 300 mg

per mantle). The activity of 6-month-old thorium was also found to provide a conservative

estimate of the total activity in the mantles for the first year or two after manufacture, when

most mantles are used (Luetzelschwab and Googins, 1984).

If radiation doses are primarily due to external irradiation of the total body by photons from the

gas mantles, the EDE can be equated with the whole body” dose obtained previously by

O’Donnell and Etnier (NUREG/CR–1910). This is possible because the external radiation

doses from the thorium decay series are due to high-energy gamma rays coming primarily from

228Ac, 212Pb, 212Bi, and 208Tl (Luetzelschwab and Googins, 1984). However, if internal exposures

from ingestion or inhalation are involved, the 50-year committed EDEs are reevaluated here

using dose conversion factors from Table 3.1.7 in Section 3.1 of this report. The dose

conversion factor in Table 3.1.7 for inhalation of 220Rn and its short-lived progeny, 216Po through

208Tl, assumes that all of these radionuclides are in radioactive equilibrium in air. It must be

multiplied by so-called equilibrium factor to account for the normal disequilibrium between the

220Rn and its short-lived products in air (International Commission on Radiological Protection

(ICRP) 50). The value of the equilibrium factor may vary typically between 0.1 and 0.01 for

indoor air of homes and by values even smaller than 0.01 for outdoor air.
3–81

3.4.4.1 Distribution and Transport

This section considers potential radiation doses for (1) transport of gas mantles to distributors,

and (2) distribution of gas mantles by retail stores and both distribution and installation of gas

mantles by utility companies. Utility company installers are a special case of distribution

workers because they could inhale or ingest radionuclides while replacing gas mantles in

commercial and residential outdoor gaslights.

During most distribution and transport activities, the potential doses to workers and members of

the public are due primarily to external radiation; thus the results of the previous assessment by

O’Donnell and Etnier (NUREG/CR–1910) are used. However, the potential doses to installers

are reassessed here to obtain up-to-date estimates of the 50-year committed EDEs for

inhalation and ingestion of radionuclides during the installation of gas mantles in a lighting

device.

3.4.4.1.1 Transport

For transport of gas mantles to distributors, O’Donnell and Etnier (NUREG/CR–1910) estimated

the annual collective dose equivalent to be 0.01 person-Sv (1 person-rem). The annual

individual EDEs ranged from a low of less than 1×105

 mSv (<0.001 mrem) to members of the

public along truck routes to a high of about 0.002 mSv (0.2 mrem) for truck drivers

(NUREG/CR–1910).

3.4.4.1.2 Installation

During mantle installation, the installer may be exposed via (1) direct external irradiation,

(2) inhalation of 220Rn and its short-lived progeny because of diffusion of 220Rn from the

mantles, (3) inhalation of ash from broken mantles that is dispersed in air as respirable

particles, and (4) ingestion of ash from the used mantles. Exposure scenarios used in this

assessment for mantle replacement are the same as those in the previous work by O’Donnell

and Etnier (NUREG/CR–1910) (see Section 3.4.3).

Table 3.4.3 lists total EDEs from both external and internal exposure for mantle changes

occurring either outdoors or indoors. For outdoor replacement of two mantles, the EDE to the

installer could range from a low of less than 1×105

 mSv (<0.001 mrem) to a high of 0.02 mSv

(2 mrem). For indoor replacement of two mantles, the EDEs could range from a low of

2×104 mSv (0.02 mrem) to a high of 0.04 mSv (4 mrem). The lower doses are due to direct

external irradiation and submersion in or inhalation of airborne radionuclides due to diffusion of

220Rn from the mantles. The higher doses are mainly due to inhalation of airborne ash from the

used mantles and, to a lesser extent, to ingestion of ash from the used mantles. For utility

company installers, it is assumed that all mantle changes typically occur outdoors rather than

indoors.

Each utility company installer is assumed to replace 500 mantles per year in 250 outdoor

commercial and residential gaslights (NUREG/CR–1910). Thus, the annual individual EDE to

an installer could range between 5×104

 mSv (0.05 mrem) and 5 mSv (500 mrem). It is

doubtful, however, that these professional installers would in all cases blow mantle parts from

the gaslights and ingest ash from the used mantles. In the latter case, subsequent work

activities and hand washing would likely remove all of the adhered mantle ash from the hands. 
3–82

Hence, doses to installers were arbitrarily chosen to be 10% of the higher values given above

for outdoor mantle changes. This yielded annual individual effective doses to installers of

0.5 mSv (50 mrem), and a collective EDE of about 3 person-Sv (300 person-rem) to the 6,000

installers needed to replace the 3 million mantles used each year in commercial and residential

outdoor gaslights.

3.4.4.1.3 Distribution

For distribution of gas mantles to routine users, the collective dose equivalent is estimated to be

7 person-Sv (700 person-rem) based on the results from Section 3.4.4.1.2 of this report and the

previous assessment by O’Donnell and Etnier (NUREG/CR–1910). This total collective dose is

based on estimates of 0.04 person-Sv (4 person-rem) to warehouse workers, 0.006 person-Sv

(0.6 person-rem) to truck drivers, 0.6 person-Sv (60 person-rem) to store workers, 3 person-Sv

(300 person-rem) to store customers, 0.01 person-Sv (1 person-rem) to utility stock clerks,

3 person-Sv (300 person-rem) to utility company installers, and 6×10-5 person-Sv

(0.006 person-rem) to the public along truck routes. The annual individual EDEs ranged from a

low of less than 1×105

 mSv (<0.001 mrem) for members of the public along truck routes to a

high of 0.1 mSv (10 mrem) for warehouse workers (NUREG/CR–1910). For store customers

and store clerks, the annual individual EDEs averaged 2×105

 mSv (0.002 mrem) and

0.002 mSv (0.2 mrem), respectively.

3.4.4.2 Routine Use

This section considers potential radiation doses from routine use of gas mantles in (1) camping

lanterns, (2) outdoor commercial and residential gaslights, and (3) indoor residential lights.

During routine use, radionuclides may escape from mantles and become airborne via two

different processes: vaporization and diffusion (NUREG/CR–1910). Diffusion produces a

steady-state flux of 220Rn from a gas mantle, and the effect of the 220Rn diffusion on the thorium

decay chain is simply to reduce the values in Table 3.4.1 by the fraction of the 220Rn that

escapes. For example, the assumed diffusion rate of 25% for a mantle in storage results in

relative activities for 220Rn and its progeny that are only 75% of those given in Table 3.4.1. 

Unless a barrier intervenes such as the plastic pack around new mantles, the 220Rn that diffuses

from the mantles will provide a steady flux of 220Rn into the surrounding air. This 220Rn will

decay and introduce its progeny into the air, and the resulting airborne concentrations are a

potential source for exposure, particularly indoors.

During operation of a lighting device, gas mantles may also reach high enough temperatures to

volatilize radionuclides contained within the mantles (NUREG/CR–1910). This process was

noted in a study by Griggs (1973) and is supported by measurements from a study of

Luetzelschwab and Googins (1984). To discuss this process, it is convenient to divide the

thorium decay chain of Table 3.4.1 into three groups of nuclides: (1) 232Th and 228Th; (2) 228Ra,

228Ac and 224Ra; and (3) 220Rn and its progeny (216Po through 208Tl). Vaporization is

characterized by an initial, pulse-type release of radionuclides in Groups 2 and 3, but not those

in Group 1 (Luetzelschwab and Googins, 1984). Unlike diffusion, which is a continuing effect,

the effects of vaporization are short-term because the 220Rn and its progeny (Group 3) will

retain radioactive equilibrium within the mantle over a period of a few days. Thus, vaporization

of radionuclides from mantles is a complex function of the prior use of the mantles

(NUREG/CR–1910).
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The effects of vaporization and diffusion of radionuclides from the mantles have been taken into

account in the internal dose calculations of this study. Based on the measurements of

Luetzelschwab and Googins (1984), it was assumed that 30% of the radium nuclides (Group 2)

and 100% of the 220Rn and its progeny (Group 3) were released from the mantles and into the

air by vaporization in an instantaneous, pulse-like release during ignition of the lighting devices. 

It was assumed further that (1) 100% of the 220Rn diffused from the mantles and into the air

throughout continued operation of the lighting devices, and (2) 25% of the 220Rn diffused from

the mantles and into air while the mantles were simply being stored inside the nonignited

lighting devices. The effects of depletion of radionuclides in the mantles by both vaporization

and diffusion also were considered in the previous external dose calculations of O’Donnell and

Etnier (NUREG/CR–1910).

3.4.4.2.1 Camping Lanterns

Persons who use incandescent gas mantles in camping lanterns may be exposed to them

under a multitude of conditions. Thus, O’Donnell and Etnier (NUREG/CR–1910) constructed a

list of exposure events for a 2-night camping trip by an avid camper. These exposure events

are summarized in Table 3.4.4 of this report. The table identifies the types of absorbers

considered in the external dose calculations and the sources of airborne radionuclides

considered in the internal dose calculations. The times for the various exposure events during

the camping trip are total times, except for those involving the loading (or unloading) of the car. 

For loading of the car, the times are only those involved in handling the lanterns and mantles. 

The total time spent loading the car would be longer and would depend to some extent on

whether the camper is an avid camper or a one-time camper, as defined below.

Categories of campers considered here include the principal camper and three other members

in a family of avid or one-time campers. A principal camper is defined as one who participates

in all the exposure events in Table 3.4.4, with the exception of traveling in the rear seat. Other

members are defined as those who travel in the vehicle and are not involved in either purchase

or ignition of mantles. Avid campers are those who make 26 two-night camping trips per year

with two double-mantled lanterns and four replacement mantles; one-time campers are those

who make 1 two-night camping trip per year with a single-mantled lantern and one replacement

mantle. Also, a principal avid camper is assumed to replace 8 mantles per year (in 2 doublemantle lanterns two times per year); a principal one-time camper is assumed to replace

1 mantle per year. Campers also are assumed to be exposed between camping trips by

inhalation of radionuclides in air due to diffusion of 220Rn from mantles stored in lanterns in the

home.

Table 3.4.5 presents estimates of potential individual doses from the various activities

associated with camping lantern use. For the camping activities defined in Table 3.4.4, it was

estimated that avid campers and their families could receive annual EDEs of between

5×104

 mSv and 0.06 mSv (0.05 and 6 mrem) and one-time camping families could receive

between 2×105

 and 6×104 mSv (0.002 and 0.06 mrem). For the outdoor inhalation exposures,

hemispherical air spaces were assumed with the same radii and ventilation rates as those used

previously by O’Donnell and Etnier (NUREG/CR–1910). For indoor inhalation exposures, the

volume of the car was assumed to be 6 m3

 with a ventilation rate of 5 volume changes per hour

(see Appendix A.1), and the volume of the tent was assumed to be 20 m3

 and its ventilation rate

was assumed without basis to be the about the same as that of the car (i.e., 5 volume changes

per hour). For these inhalation exposures, committed EDEs were calculated using the
3–84

dosimetric data from ICRP 50 (see references) and Federal Guidance Report No. 11

(EPA–520/1–88–020). For the direct external exposures, EDEs were estimated using

MicroShield (Computer Codes, Grove Engineering, 1996). 

Table 3.4.5 also presents estimates of potential individual doses from mantle replacement (see

Section 3.4.4.1.2) and from storage of camping lanterns in the home. Annual EDEs were

calculated to be 6×104

 mSv (0.06 mrem) from home storage of a camping lantern with one

mantle (by one-time campers) and 0.002 mSv (0.2 mrem) from home storage of eight mantles

(by avid campers). For both avid and one-time campers, the EDE for the mantle replacement

dominates the annual EDEs given in Table 3.4.5. These calculations assume that a house has

a volume of 450 m3

 and a ventilation rate of 1 volume change per hour (see Appendix A.1) and

that campers are in their homes for 200 effective days or a total of 4,800 h/yr

(NUREG/CR–5512).

To estimate the collective dose to members of the public from routine use of mantles in

camping lanterns, the annual individual doses in Table 3.4.5 for each of the camper types was

multiplied by the number of campers of that type (one principal camper and three other

campers) and divided by the number of mantles each type of principal camper purchased and

burned during the year (one by one-time principal campers and eight by avid principal

campers). The resulting value of 1×105

 person-Sv (0.001 person-rem) per mantle was then

multiplied by the number of mantles used for camping, approximately 20.25 million, to give an

annual collective EDE of about 400 person-Sv (40,000 person-rem).

3.4.4.2.2 Indoor Residential Lamps

Indoor lamps containing mantles may be used functionally (e.g., in vacation cabins and

cottages) and decoratively (e.g., in residential homes). No data are available concerning the

relative magnitude of the two uses. O’Donnell and Etnier (1981) assumed that 75% were used

functionally in vacation cottages and 25% were used decoratively in permanent residences.

The mantles in decorative lamps will be used only occasionally, and they are assumed to

behave primarily like mantles in camping lanterns stored in a home (O’Donnell and

Etnier,1981). The average annual EDE from a decorative lamp in a permanent residence

containing two mantles is estimated to be about 0.2 mSv (20 mrem). This dose estimate is

based on the following assumptions: (1) 25% of the 220Rn contained in two mantles emanates

from the lamp and disperses into the home air; (2) the exposed persons are in the residence for

200 effective days per year (4800 h/yr) (NUREG/CR–5512); (3) the volume of the residence is

450 m3

, and the ventilation rate is 1 volume change per hour.

The mantles in functional lamps will be used regularly, and the annual EDE from a functional

lamp in a vacation cottage is calculated using the following assumptions: (1) two mantles are

purchased and replaced each year; (2) the vacation cottage is used for 52 days (26 weekends)

and is occupied for 12 h/day; (3) the lamp is actually lighted for 4 of the 12 hours; and (4) the

vacation cottage home has a volume of 225 m3

 (one-half of the volume of a residence) and a

ventilation rate of 1 volume change per hour. The person who replaces the two mantles during

the year receives an additional annual EDE of 0.002 mSv (0.2 mrem). The average annual

effective dose from use of mantles in a vacation cottage is 0.1 mSv (10 mrem).
3–85

If there are two mantles per residence and the average residence is occupied by four persons,

use of 1.25 million mantles could expose 640,000 persons in approximately 160,000 permanent

residences and 1.86 million persons in approximately 465,000 vacation cottages or cabins. 

Thus the annual collective EDE to the public from indoor residential lamps is estimated to be

about 300 person-Sv (30,000 person-rem). 

There is a special scenario regarding the continuing use of mantles in a permanent residence

as the only source of lighting in remote areas. The mantles in four functional lamps are used

regularly, and the annual EDE from use is calculated to be 2 mSv (200 mrem). This dose

estimate is based on the following assumptions: (1) 25% of the 220Rn contained in four mantles

emanates from the lamp and disperses into the home air; (2) the exposed persons are in the

residence for 200 effective days per year (4800 h/yr) (NUREG/CR–5512); (3) the volume of the

residence is 500 m3

, and the ventilation rate is 1 volume change per hour; (4) during six colder

months, the lanterns burn for 10 hours per day; and during six warmer months, the lanterns

burn for 2 hours per day. It is unknown how many people actually use mantles as their primary

source of lighting; however, the hypothetical annual EDE for four people exposed in each of a

nominal 1000 residences would be about 8 person-Sv (800 person-rem).

3.4.4.2.3 Outdoor Gaslights

Outdoor residential gaslights are usually located several meters from residences

(NUREG/CR–1910). For a gaslight containing two mantles that is located 18 meters from three

homes and swept by a 3-km/h breeze, essentially all of the radiation dose can be attributed to

direct external exposure to the mantles contained in the gaslights. Thus, based on the previous

results of O’Donnell and Etnier (NUREG/CR–1910), it was estimated that the annual EDE from

external irradiation of the whole body of an exposed individual would be less than 1×105

 mSv

(<0.001 mrem).

If 1.5 million outdoor gaslights are used to house the 3 million mantles and that 12 persons from

the two homes are exposed to the mantles in each gaslight, the collective EDE could be

0.05 person-Sv (5 person-rem). It is assumed here that the collective doses from residential

outdoor gaslights and commercial outdoor gaslights are similar. More people would be

exposed to commercial outdoor gaslights, but the individual doses would be significantly less

since the people would be exposed for much shorter periods of time.

3.4.4.3 Disposal

This section provides estimates of individual and collective doses to members of the public from

landfill disposal and by incineration of thorium-containing mantles. These dose estimates are

based on the generic disposal methodology in Appendix A.2 of this report.

Used mantles may be discarded as domestic solid waste or as litter at mantle replacement sites

(NUREG/CR–1910). The potential doses to campers from used thorium-containing mantles

discarded as litter is one of the exposure events considered during a typical 2-night camping

trip by an avid camper (see Table 3.4.4). It is assumed here that 10 million mantles per year

are discarded as litter at camping sites or other mantle replacement sites and 15 million

mantles per year are discarded as domestic solid waste. The amount of thorium contained in

the 15 million mantles disposed of as solid waste at landfills and by incineration would be

3.75×106

 g, and it would be in a form that is readily dispersible in air and readily accessible to
3–86

infiltrating ground water at landfills. Thus, the dose-to-source ratios for individual and collective

doses in Appendix A.2 are used without modification.

For disposal at landfills, the annual individual EDE would be about 0.01 mSv (1 mrem) to either

waste collectors or future on-site residents. Annual individual doses to workers at landfills and

off-site members of the public would be substantially less. The total collective dose to members

of the public was found to be about 10 person-Sv (1000 person-rem), due almost entirely to

exposure to future on-site residents at landfills for 1000 years after loss of institutional controls

over the sites. If exposure to future on-site residents is not considered, the collective EDE

would be 0.07 person-Sv (7 person-rem), due mainly to exposure to waste collectors and

workers at landfills.

For disposal by incineration, the annual individual EDE would be about 0.07 mSv (7 mrem) to

waste collectors. Annual individual doses are substantially less to workers at incinerators and

off-site members of the public. The total collective dose to members of the public would be

about 0.01 person-Sv (1 person-rem), due mainly to exposure to waste collectors at

incinerators.

3.4.4.4 Accidents and Misuse

This section considers: (1) the internal dose to a small child who ingests part of a used gas

mantle, (2) the external dose to a person who uses a pack of 2 gas mantles as a calibration

source, and (3) the total internal and external dose to individuals from a residential fire involving

8 gas mantles (4 mantles installed in lanterns and 4 replacement mantles) and a warehouse fire

involving 10,000 gas mantles.

For ingestion by a small child, it was assumed the child would find a discarded mantle at a

camp site and digest part of the mantle while playing with it (NUREG/CR–1775). If 20% of the

ash from the used mantle (50 mg of thorium) adheres to the child’s hands, if subsequent wiping

and washing removes all but 5% of the adhered ash (2.5 mg of thorium), and if the 2.5 mg of

thorium is ultimately ingested by the child, the child could receive an EDE of about 0.01 mSv

(1 mrem), based on dose conversion factors for an adult (see Section 3.1), or 0.02 mSv

(2 mrem), based on age-dependent dose conversion factors for a 1- to 5-year-old child

(ICRP 69).

For misuse as a calibration source, it was assumed an individual holds the mantles in his or her

hands for about 10 minutes per day or 40 h/yr while calibrating radiation survey instruments for

photons and beta particles. The dose equivalent rate at 1 meter from a package with two

mantles containing 500 to 600 mg (250 to 300 mg per mantle) of 6-month-old thorium is well

established as 2×104 Sv/h (0.02 rem/h) (Luetzelschwab and Googins, 1984). Actual

measurements of the dose equivalent rate at the surface of a pack with two mantles containing

300 mg (150 mg per mantle) of thorium have been reported by O’Donnell and Etnier

(NUREG/CR–1910). The dose rate near the surface of the package from beta particles and

photons was approximately 0.01 mSv/h (1 mrem/h), and the photon dose rate was estimated to

be 0.001 mSv/h (0.1 mrem/h). If these dose rates are normalized to a pack with two mantles

containing 500 to 600 mg (250 to 300 mg per mantle), then the dose equivalent rates at the

surfaces of the pack from beta particles and photons would be 0.02 mSv/h (2 mrem/h), and the

photon dose rate would be approximately 0.002 mSv/h (0.2 mrem/h). Thus, the annual

individual EDE from external irradiation of the total body by photons would be less than
3–87

1×105

 mSv/h (<0.001 mrem), and the annual dose equivalent to the hands from both beta

particles and photons could be 0.08 mSv (8 mrem).

For accidents involving fire, a residential fire involving 8 gas mantles and a warehouse fire

involving 10,000 gas mantles were considered. A release fraction of 0.1% is assumed for the

radionuclides in each mantle which is assumed further to contain 250 mg of 6-month-old

thorium (see Table 3.4.1). Based on these assumptions and the generic accident methodology

in Appendix A.1, the estimates of individual dose from fires involving thorium-containing

mantles are summarized as follows:

For a person trying to escape from a residential fire or a neighbor trying to rescue a

person from a residential fire, the individual EDE from eight mantles containing a total of

2 g of thorium could be 0.002 mSv (0.2 mrem).

For a firefighter wearing a respirator at a residential fire, the individual EDE from eight

mantles containing a total of 2 g of thorium could be less than 1×105

 mSv

(<0.001 mrem). The individual EDE could be 2×105

 mSv (0.002 mrem) to a worker who

is involved in the cleanup following the fire and does not wear a respirator.

For a firefighter wearing a respirator at a warehouse fire, the individual effective dose

from 10,000 mantles containing a total of 2.5 kilogram (kg) of thorium (in equilibrium

with its daughters) could be 0.001 mSv (0.1 mrem). The individual EDE could be

0.005 mSv (0.5 mrem) to a worker who is involved in the cleanup following the fire and

does not wear a respirator.

3.4.5 Summary

Table 3.4.2 presents the results of the current assessment of potential radiation doses to the

public for an annual distribution, use, and disposal of 25 million gas mantles containing thorium. 

It is assumed each gas mantle contains 250 mg of 6-month-old thorium (see Table 3.4.1).

However, Buckley et al. (NUREG/CR–1775) reported that the thorium content could be as high

as 400 mg. In this case, the estimated individual doses for routine use, accidents, and misuse

would be about two times greater than the doses for mantles containing 250 mg of thorium.

For routine use, the most highly exposed individual (a resident whose only source of lighting

was the use of mantles) was estimated to receive an annual EDE of 2 mSv (200 mrem). The

total collective dose to the public was estimated to be 700 person-Sv (70,000 person-rem).

For misuse, this analysis estimated an individual EDE of 0.02 mSv (2 mrem) to a small child

who plays with a used mantle and subsequently ingests part of the ash from the mantle. The

individual effective doses were found to be less for other potential misuse and accident

scenarios.

Previous studies (NCRP 95) estimate a total population dose of about 90 person-Sv

(9000 person-rem) to the public from gas mantles containing thorium, and this assessment

finds a total population dose that is three times larger. These differences can be attributed to

the following factors:
3–88

Dose equivalents to the lungs and to bone (the red marrow in bone and the endosteal

cells on the bone surfaces) were not treated properly by the older dosimetry used in the

previous studies (EPA–520/1–88–020).

Previous estimates of collective dose were based on averages over rather large dose

ranges. Point values used in the current estimates of collective dose tend to be closer

to the maximum dose values calculated in the previous studies (NUREG/CR–1910).

The standard assumption used in estimating collective doses in this study is four

individuals per family. Previous studies assumed only three individuals per family

(NUREG/CR–1910).
3–89

Table 3.4.1 Radioactivity in a Gas Mantle Containing 250 mg of Natural Thorium or

Thorium That Has Been Chemically Extracted From Ore and Aged for 6 Months

Radionuclide

Activity in Natural Thorium

(nCi)a

Activity in 6-Month-Old Thorium

(nCi)a

232Th

228Ra

228Ac

228Th

224Ra

220Rn

216Po

212Pb

212Bi

212Po

208Tl

27

27

27

27

27

27

27

27

27

17

10

27

 2

 2

23

23

23

23

23

23

15

 8

Total 270 192

a

 1 nCi = 37 Bq.
3–90

Table 3.4.2 Potential Radiation Doses From Incandescent Gas Mantles

Containing Thorium

Exposure Pathway

Individual Annual Effective

Dose Equivalent

(mrem)b

Collective Effective

Dose Equivalenta

(person-rem)b

Distribution and transport

 Transport

 Distribution

 0.2c

 50d

1

 700

Routine use

 Portable camping lanterns

 Indoor residential lamps

 Permanent cabin

 Outdoor gas lights

 10e

 20 

 200

 <0.001f

 

 40,000

 30,000

 800

 5

Disposal

 Landfills

 Incineration

 1g

 7h

 1,000

 1

Accidents and misuse

 Use as a calibration source

 Fire

 Ingestion of mantle ash by small child

 0.008i

 0.5j

 2k

a

 Collective doses are based on the annual distribution, use, and disposal of 25 million gas

mantles containing 250 mg each of thorium.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 Dose estimate applies to truck drivers; dose estimates are less for workers in truck terminals

and for members of the public along truck routes (see Section 3.4.4.1.1).

d

 Dose estimate applies to utility company installers; dose estimates are less for distribution

warehouse workers, truck drivers, store workers and customers, stock clerks, and members of

the public along truck routes (see Sections 3.4.4.1.2 and 3.4.4.1.3).

e

 Dose estimate applies to principal avid-camping users of thorium-containing mantles; dose

estimates for all campers using thorium-containing mantles are less (see Section 3.4.4.2.1).

f

 Dose estimate applies to persons exposed to thorium-containing mantles in outdoor residential

gaslights; dose estimates would be less for persons exposed to thorium-containing mantles in

outdoor commercial gaslights (see Section 3.4.4.2.4).

g

 Dose estimate applies to waste collectors or future on-site residents at landfills; dose

estimates are less for workers at a landfill and off-site members of public (see Section 3.4.4.3).

h

 Dose estimate applies to waste collectors at incinerators; dose estimates would be less for

workers at incinerators and off-site members of public (see Section 3.4.4.3).

i

 Dose estimate applies to photon irradiation of total body of a person using two mantles in a

pack as a calibration source; dose estimate for beta-particle and photon dose to hands is

0.06 mSv/yr (6 mrem/yr).

j

 Dose estimate applies to a worker who is involved in the cleanup following a warehouse fire

and does not wear a respirator; dose estimates would be less for a firefighter involved in a

warehouse fire and for a firefighter or others involved in a residential fire (see Section 3.4.4.4).

k

 Dose estimate applies to 1- to 5-year-old child; dose estimates would be less for older children

or adults (see Section 3.4.4.4).
3–91

Table 3.4.3 Summary of Potential Radiation Doses From Changing Two Mantles

Effective Dose Equivalent

Exposure Condition Outdoor Changes

(mrem)a

Indoor Changes

(mrem)a

1b

2c

3d

<0.001

 2

 2

0.02

 2

 4

a

 1 mrem = 0.01 mSv.

b

 Condition 1: Includes both external irradiation of the whole body by photons from mantles and

inhalation of radionuclides from diffusion of 220Rn from mantles (see Section 3.4.3). c

 Condition 2: Includes Condition 1 plus ingestion of ash from used mantles (see Section 3.4.3).

d

 Condition 3: Includes Condition 2 plus inhalation of ash from blowing used mantle parts from a

lighting device (see Section 3.4.3).
3–92

Table 3.4.4 Exposure Conditions for a 2-Night Camping Trip 

for an Avid Camper and Family

Exposure

Eventa

Duration of

Exposure

Number of

Mantles

Average

Distance

From

Mantles

(meters)

Identity of

Absorbers

Sources of

Airborne

Radionuclides

FIRST DAY

 Purchase 1 h 4 1 None None

 Load car 0.5 min 4

4

1

1

Lantern

Box

None

None

Travel to camp

 Front seat

 Back seat

1.5 h

 1.5 h

4

4

4

4

2

2

1

1

Lantern, seats,

cargo

Box, seats, cargo

Lantern, seats,

cargo

Box, seats, cargo

Diffusion

None

Diffusion

None

 Unload car 0.5 min 4

4

1

1

Lantern

Box

None

None

 Set up camp 2 h 2

2

4

3

6

3

Lantern

Lantern

Box

Diffusion

Diffusion

None

 Ignite lanterns 1 min 4

2

4

1

6

6

Lantern

Lantern

Box

Vaporization &

diffusion

Vaporization &

diffusion

None

 Under light

 outside

3 h 4

4

3

3

Lantern

Box

Diffusion

None
3–93

Table 3.4.4 Exposure Conditions for a 2-Night Camping Trip

for an Avid Camper and Family (continued)

Exposure

Eventa

Duration of

Exposure

Number of

Mantles

Average

Distance

From

Mantles

(meters)

Identity of

Absorbers

Sources of

Airborne

Radionuclides

FIRST DAY

Inside tent

 Lantern on 0.5 h 2

2

4

1

6

6

Lantern

Lantern

Box

Diffusion

None

None

 Lantern off 7.5 h 2

2

4

1

6

6

Lantern

Lantern

Box

Diffusion

None

None

SECOND DAY

At site 6 h 4

4

6

6

Lantern

Box

Diffusion

None

Ignite lanterns (same as above)

Under lights outside (same as above)

Inside tents (same as above)

THIRD DAY

At site (same as above)

Load car (same as above)

Travel home (same as travel to camp)

Unload car (same as above)

a

 It is also assumed each person is exposed to ground contamination of 0.6 g/cm2 from

discarded mantles for a total of 38 hours during stay at camping site.
3–94

Table 3.4.5 Estimates of Annual Effective Dose Equivalent 

for One-Time and Avid Campers

One-Time Campers Avid Campers

Exposure Events

Principal

(mrem)b

Othersa

(mrem)b

Principal

(mrem)b

Othersa

(mrem)b

 Camping activities 0.06 0.002 6 0.05

 Mantle replacement 1 8

 Home storage 0.06 0.06 0.2 0.2

a

 Includes other family members.

b

 1 mrem = 0.01 mSv.
3–95

3.5 Thorium in Vacuum Tubes

3.5.1 Introduction

In 10 CFR 40.13(c)(1)(ii), any person is exempt from the requirements for a license to the

extent that the person receives, possesses, uses, or transfers any quantity of thorium contained

in vacuum tubes. This exemption was established on March 20, 1947 (12 FR 1855), and has

remained essentially unchanged since then.

The Federal Register notice establishing the exemption for vacuum tubes containing thorium

provided no information on radiological impacts on the public from the use and disposal of this

product. Information published by the Atomic Energy Commission in 1960 (25 FR 8619)

indicated that the exemption would not result in an unreasonable hazard to life or property, but

no indication was given that a dose analysis had been performed to support the exemption.

3.5.2 Description of Items

Many vacuum tubes used in applications requiring high output powers contain directly heated

cathodes (Ewell, 1981). These cathodes are frequently made from thoriated-tungsten wires

containing 1 to 2% by weight of thorium. The thorium is used because of its ability to emit

electrons at relatively low temperatures when heated in a vacuum (Davis, 1977).

The vacuum tubes containing thoriated-tungsten cathodes are of many varied designs (Davis,

1977). High-output power tubes may range from tens to hundreds of centimeters in dimension

and from a few to 150 kg in mass. The cathodes may have masses between 100 g and 100 g

and contain between 1 g and 2 g of thorium. Tube envelopes may be metal, glass, ceramic,

or any combination thereof. The envelopes will normally contain any 220Rn that diffuses from

the thoriated-tungsten wire.

The use of vacuum tubes in electronics has been in decline since the development of the

transistor and other solid-state devices after World War II (Liao, 1985). In spite of the success

of solid-state devices in most electronics applications, there are certain specialized functions

that only vacuum tubes can perform (Liao, 1985; Parker and Abrams, 1992). These functions

usually involve operation at extremes of power or frequency. However, sales of such vacuum

tubes during the 1980s and beyond are estimated to be 100,000 units per year or less (Ewell,

1981; Garoff, 1979).

The current market for vacuum tubes is dominated by the magnetrons in microwave ovens sold

for home and restaurant use. It is currently estimated that more than 90% of all U.S. homes

own at least one microwave oven (Reda, 1995; Kulman, 1997). A single magnetron is typically

used in the low-wattage ovens for home use (i.e., 600 to 1000 watts) (Microwave Ovens, 1996),

while as many as three or four magnetrons may be used in the heavy-duty, high-wattage ovens

for restaurant use (i.e., 2000 to 2600 watts). Sales of such magnetrons are estimated to be

approximately 10 million units per year.

The typical cathode in magnetrons for home use has 8 to 10 turns of thoriated-tungsten wire on

a diameter of 200 mils, the diameter of the wire is 15 to 20 mils, and the amount of thorium in

the wire is 1 to 2% by weight (Phone call, J. Osepchuk, Full Spectrum Consulting, Concord,
3–96

MA, June 1997). If the density is taken as 19.3 g/cm3

 for tungsten and 11.2 g/cm3 for thorium,

the mass of the thoriated-tungsten wire ranges from 300 to 600 mg. The typical wire mass in a

microwave oven magnetron is about 400 mg per tube, and the amount of thorium is typically

4 to 8 mg per tube (Phone call, Amana Consumer Affairs, Amana Refrigeration, Inc.,

Amana, IA, 55204, June 1997).

3.5.3 Summary of Previous Analyses and Assessments

There are no known previously published analyses of radiological impacts on the public

associated with this exemption. Furthermore, use of thorium in vacuum tubes is not mentioned

in either of the reports on consumer products by the National Council on Radiation Protection

and Measurements (NCRP 56; NCRP 95).

3.5.4 Present Exemption Analysis

Table 3.5.1 presents results of the current assessment of potential radiation doses from

vacuum tubes containing thorium. These results are based on an annual distribution of

10 million magnetrons in microwave ovens containing 8 mg of natural thorium each, and an

annual distribution of 100,000 other types of vacuum tubes containing an average of 0.25 g of

natural thorium each. The effective lifetime of both the magnetrons used in microwave ovens

and the vacuum tubes used in other electronic applications is assumed to be 10 years.

3.5.4.1 Distribution and Transport

This section estimates the potential radiation doses to the public from the distribution and

transport of vacuum tubes containing natural thorium, based on the generic distribution

methodology of Appendix A.3.

3.5.4.1.1 Magnetrons in Microwave Ovens

In applying the generic methodology in Appendix A.3 to magnetrons in microwave ovens, the

following assumptions have been made. The microwave ovens, each containing one

magnetron, are assumed to be shipped from the manufacturers or importers to distribution

centers. The number of microwave ovens in a typical shipment is assumed to be 1000

magnetrons containing 8 mg of natural thorium each. It is assumed further that commercial

semi-trucks are used to ship the microwave ovens between distribution centers, and microwave

ovens pass through an average of three distribution centers before being delivered to large

retail stores. In addition, it is assumed that (1) a semi-truck driver may pick up a shipment as

often as once per week (50 shipments per year) from the same manufacturer during the year,

(2) a stockhandler in a distribution center may be exposed to 3 shipments of microwave ovens

per week (150 shipments per year), and (3) a store clerk may be exposed continuously to a

display of 20 microwave ovens in a retail store.

Based on the above assumptions and the generic methodology in Appendix A.3, the annual

individual effective dose equivalent (EDE) would be less than 1×105

 millisievert (mSv)

(<0.001 mrem) to a stockhandler in a large distribution center (i.e., large warehouse). Individual

doses would be less to semi-truck drivers, stockhandlers in retail stores, store clerks and

customers in retail stores, and members of the public along truck routes. The collective EDE
3–97

from all distribution and transportation activities could be 0.2 person-Sv (20 person-rem), due

almost entirely to exposure to retail store clerks and customers.

3.5.4.1.2 Other Types of Vacuum Tubes

In applying the generic methodology of Appendix A.3 to vacuum tubes used in other

applications, the following assumptions have been made. The vacuum tubes are assumed to

be shipped from the manufacturer or importer to the customer by commercial semi-truck, and

the number of vacuum tubes in a typical shipment is assumed to be 10 tubes containing 0.25 g

of natural thorium each. It is assumed further that (1) the vacuum tubes pass through an

average of three truck terminals before reaching their final destination, (2) the exposure to

workers in a truck terminal is the same as those estimated for workers in a large warehouse,

and (3) the same semi-truck driver could pick up a shipment as often as once per week (50

shipments per year) from the same manufacturer.

Based on the above assumptions and the generic methodology in Appendix A.3, the annual

individual EDE would be less than 1×105

 mSv (<0.001 mrem) to a semi-truck driver who picks

up a number of shipments from the same manufacturer during the year. Individual doses would

be less to other semi-truck drivers, workers in truck terminals, and members of the public along

truck routes. The collective EDEs from all distribution and transportation activities could be

1×104

 person-Sv (0.01 person-rem).

3.5.4.2 Routine Use

This section estimates potential radiation doses to routine users from magnetrons and other

vacuum tubes containing thorium. The exposure scenarios used here may be somewhat out of

date, particularly for vacuum tubes other than magnetrons in microwave ovens, but the results

are still considered to be indicative of the potential radiation doses from typical routine use of

vacuum tubes in radio and TV broadcasting and in various radar applications. Routine users

will not receive any internal doses because 220Rn should not diffuse through the walls of the

vacuum tubes. Exposures will be from photons emitted by thorium and its radioactive decay

products (see Section 3.1).

3.5.4.2.1 Magnetrons in Microwave Ovens

Because of concern over microwave leakage from early microwave ovens, surveys were made

of the spatial distance of a microwave oven user as a function of operating time (Osepchuk,

1979). Results of these surveys can be summarized as follows: (1) typical microwave oven

usage in the home involved about seven starts per day, with 2- to 3-minute cooking times per

operation, (2) distance from the microwave oven during the preparation of a meal ranged from

0.7 meter for a 1-minute operation to 2 meters for a 30-minute or longer operation, and (3) the

most probable average distance from the microwave oven during the preparation of a meal was

1.2 to 1.5 meters.

A typical magnetron in a microwave oven for home use has a wall thickness of approximately

0.2 cm of copper and contains 8 mg of thorium, which is assumed to be 20-year-old natural

thorium (see Section 3.1). To estimate potential radiation doses to users of home microwave

ovens, it is assumed that the principal user is exposed at a distance of 0.7 meter for 5 min/day

during two short cooking operations and at an average distance of 1.2 meters for 1 h/day during
3–98

meal preparation, and other family members are exposed for 5 min/day at an average distance

of 0.7 meter during two short cooking operations (i.e., one of these could be during lunchtime at

work and the other during a snacktime at home). All family members also are assumed to be

exposed at an average distance of 3 meters for 30 min/day during breakfast and dinner. Thus,

the individual effective dose to the principal microwave user and to other family members could

be less than 1×105

 mSv (<0.001 mrem), based on calculations with MicroShield (Computer

Codes, Grove Engineering, 1996). The annual collective dose from 10 million microwave ovens

could be 0.04 person-Sv (4 person-rem). The total collective dose over the 10-year effective

lifetime of the ovens could be 0.4 person-Sv (40 person-rem).

The above estimates of individual dose apply to home users of low-wattage microwave ovens

with a single magnetron. The maximum dose scenario for a high-wattage, heavy-duty

microwave oven containing four magnetrons (see Section 3.5.2). It is assumed that a

restaurant cook is exposed at an average distance of 1.2 meters for 1000 h/yr (i.e., 4 h/day for

250 day/yr) during the preparation of meals. The annual individual EDE to the cook could be

2×105

 mSv (0.002 mrem) based on calculations with MicroShield (Computer Codes, Grove

Engineering, 1996).

3.5.4.2.2 Other Types of Vacuum Tubes

Because vacuum tubes other than the microwave oven magnetron vary widely in design and in

application, it is not feasible to attempt a detailed assessment of all potential applications. 

Thus, for this analysis it was decided to indicate the potential radiation doses using the

following representative scenarios: Scenario I, an individual exposed at an AM broadcasting

station to 10 vacuum tubes containing 0.1 g of thorium each; and Scenario II, an individual

exposed at a large radar installation to 22 vacuum tubes containing 0.4 g of thorium each. It is

assumed here that 50,000 of the 100,000 such vacuum tubes distributed per year are used in

radio and TV broadcasting and the other 50,000 are used in radar surveillance systems.

Scenario I. An AM radio broadcasting station with a maximum power of 50 kW contains five

operating tubes and five spare tubes. Each tube contains 0.1 g of thorium, which is assumed to

be 20-year-old natural thorium (see Section 3.1). The tube envelopes are cylinders of

aluminum with a wall thickness of 0.2 cm, and all tubes are in an equipment room that is

separated from the operator by a cinder block wall containing a 2.5-cm-thick glass window. The

operating tubes also are enclosed in a steel cabinet with a wall thickness of 0.1 cm.

One operator is assumed to work in the control room for 2000 h/yr at an average distance of

1 meter from each tube, and four operators are required for continuous broadcasting. 

Accounting for photon absorption in the aluminum cylinders, glass window, and steel cabinet

(for the five operating tubes), each operator is estimated to receive an annual individual EDE of

0.002 mSv (0.2 mrem) based on calculations with MicroShield (Computer Codes, Grove

Engineering, 1996). The annual collective dose from 50,000 tubes could be 0.04 person-Sv

(4 person-rem), and the total collective dose over the assumed 10-year effective lifetime of the

vacuum tubes could be 0.4 person-Sv (40 person-rem).

Scenario II. A large radar surveillance station contains 11 operating vacuum tubes and

11 spare tubes. Each tube contains 0.4 g of thorium, assumed to be natural thorium. The tube

envelope is inside a 0.75-cm-thick ceramic material that is covered by a 0.2-cm-thick aluminum

cylinder. The operating tubes are also enclosed in a 0.1-cm-thick iron cabinet.
3–99

A five-person crew operates the station and rotates duties so that each operator is in the control

room about 5 hours per shift. Continuous operation of the station requires four operating

crews. Hence, each operator is assumed to be in the control room for about 1300 h/yr. 

Approximately 25% of this duty time (325 h/yr) is spent working at an average distance of

0.7 meter from 2 operating tubes, 2 meters from the other 9 operating tubes, and 3 meters from

the 11 spare tubes. During the remainder of the duty time (975 h/yr), each operator is assumed

to be exposed at an average distance of 3 meters from all 22 tubes. Each operator is

estimated to receive an annual individual EDE of 0.002 mSv (0.2 mrem) based on calculations

with MicroShield (Computer Codes, Grove Engineering, 1996). The annual collective dose from

50,000 tubes could be 0.09 person-Sv (9 person-rem). The total collective dose over the

assumed 10-year effective lifetime of the vacuum tubes could be 0.9 person-Sv

(90 person-rem).

3.5.4.2.3 All Types of Vacuum Tubes

As mentioned previously, the above exposure scenarios for vacuum tubes other than the

magnetrons in home microwave ovens may be outdated because they were developed from

data relevant to the early 1980s. Nevertheless, the results are still considered to be indicative

of potential radiation doses from the routine use of vacuum tubes in radio and TV broadcasting

and in various radar applications. The total collective dose from routine use of 1 year’s

distribution of all types of vacuum tubes containing thorium is estimated to be 2 person-Sv

(200 person-rem).

3.5.4.3 Disposal

This section estimates potential radiation doses from disposal of magnetrons in microwave

ovens and other types of vacuum tubes based on the generic disposal methodology in

Appendix A.2. The amount of natural thorium in the magnetrons of microwave ovens is

assumed to be 80 kg, and the amount of thorium in other types of vacuum tubes is assumed to

be 25 kg. Further, it is assumed here that all of the microwave ovens and equipment containing

the other vacuum tubes are sent to landfills for disposal and that the thorium remains within the

tubes during landfill disposal. Thus, the following adjustments are made to the source-to-dose

ratios in Appendix A.2: (1) there is no inhalation or ingestion by waste collectors or workers at a

landfill, and (2) there is no exposure to off-site members of the public due to airborne releases

during landfill operations.

Based on the above assumptions and the generic disposal methodology in Appendix A.2, the

individual and collective doses can be summarized as follows:

The annual individual EDEs to waste collectors at landfills could be 2×104

 mSv

(0.02 mrem) from microwave oven magnetrons, 5×105

 mSv (0.005 mrem) from other

types of vacuum tubes. Individual doses to workers at landfills, off-site members of the

public, and future on-site residents would be less.

The collective EDEs could be 0.3 person-Sv (30 person-rem) from microwave oven

magnetrons, 0.1 person-Sv (10 person-rem) from other types of vacuum tubes, and

0.4 person-Sv (40 person-rem) from both sources, due almost entirely to exposure to

future on-site residents at landfills for 1000 years after the loss of institutional controls

over the sites.
3–100

If the exposure to future on-site residents is not considered, the collective dose would be

4×104 person-Sv (0.04 person-rem) from microwave oven magnetrons,

1×104

 person-Sv (0.01 person-rem) from other types of vacuum tubes, and 5×104

person-Sv (0.05 person-rem) from both sources, due mainly to exposure to workers at

landfills.

3.5.4.4 Accidents and Misuse

This section estimates the potential radiation doses from accidents and misuse of vacuum

tubes using the generic accident methodology in Appendix A.1.

In the case of misuse, the exposure to a person who finds and vandalizes a large vacuum tube

containing 2 g of natural thorium (see Section 3.5.2) is considered. The person is assumed to

destroy the vacuum tube out of curiosity, inhaling radioactive material released from the tube

(i.e., 220Rn and its short-lived progeny). It is assumed further that the person is exposed for

30 minutes in a small room with an enclosed volume of 18 m3

 and a ventilation rate of 1 volume

change per hour (see data for a small watch repair shop in Tables A.1.2 and A.1.9 of

Appendix A.1), and the thorium is 20-year-old thorium and the amount of 220Rn in the tube is

8 kilobecquerels (kBq) (0.22 microcurie (Ci)) (i.e., all of the 220Rn and daughters escape from

the cathode of the tube). Based on these assumptions and the generic accident methodology

of Appendix A.1, the individual EDE to this person could be 0.01 mSv (1 mrem). For a

microwave oven size tube, the individual EDE could be less than 1×105

 mSv (<0.001 mrem)

In the case of accidents, the following assumptions have been made: (1) a residential fire

involving a single microwave oven magnetron containing 8 mg of natural thorium, (2) a

transportation fire involving a typical shipment of 1000 microwave ovens containing 8 mg of

natural thorium each, and (3) a typical shipment of 10 vacuum tubes of other types containing

0.25 g of natural thorium each. It is also assumed that the release factor for the thorium

contained in the microwave oven magnetrons and other types of vacuum tubes is 0.01%. 

Based on these assumptions and the generic accident methodology in Appendix A.1, the

individual doses can be summarized as follows:

For a person trying to escape from a residential fire or a neighbor trying to rescue a

person from a residential fire, the individual EDE from a microwave oven magnetron

containing 8 mg of 20-year-old natural thorium could be less than 1×105

 mSv

(<0.001 mrem).

For a firefighter wearing a respirator at a residential fire, the individual EDE from a

microwave oven magnetron containing 8 mg of natural thorium could be less than

1×105

 mSv (<0.001 mrem). The individual EDE could be less than 1×105 mSv

(<0.001 mrem) to a worker who is involved in the cleanup following the fire and who

does not wear a respirator.

For a firefighter wearing a respirator at a transportation fire, the individual EDE from a

shipment of 1000 microwave oven magnetrons containing a total of 0.8 g of natural

thorium could be 2×105

 mSv (0.002 mrem). The individual EDE from a shipment of

10 vacuum tubes containing a total of 2.5 g of natural thorium could be less than

1×105

 mSv (<0.001 mrem).
3–101

For a worker who is involved in the cleanup following a transportation fire and who does

not wear a respirator, the individual EDE from the shipment of 1000 microwave oven

magnetrons could be 2×104

 mSv (0.02 mrem). The individual EDE from the shipment

of 10 vacuum tubes of other types could be 5×105

 mSv (0.005 mrem).

3.5.5 Summary

Table 3.5.1 presents results of the current assessment of the potential radiation doses to the

public from vacuum tubes containing thorium. These results are based on an assumed annual

distribution of 10 million magnetrons in microwave ovens containing 8 mg of natural thorium

each and 100,000 other types of vacuum tubes containing an average of 0.25 g of natural

thorium each. The effective lifetime of both the magnetrons used in microwave ovens and the

vacuum tubes used in other electronic applications is assumed to be 10 years.

For routine use of these vacuum tubes, including distribution and disposal in landfills, the

annual EDE could be 0.002 mSv (0.2 mrem) to a routine user, and the total collective dose from

1 year’s distribution of vacuum tubes containing thorium could be 3 person-Sv

(300 person-rem), due mainly to exposure to individuals during routine use. For routine use of

microwave ovens in the home, the annual individual effective dose was estimated to be less

than 1×105

 mSv (<0.001 mrem).

For an accident involving fire, the individual EDE could be as much as 5×105

 mSv

(0.005 mrem), and for misuse involving vandalism of a vacuum tube, the individual effective

dose could be 1×104

 mSv (0.01 mrem). Thus, the potential radiation doses from accidents and

misuse appear to be very low.
3–102

Table 3.5.1 Potential Radiation Doses From Vacuum Tubes Containing Thorium

Exposure Pathway

Individual Annual

Effective

Dose Equivalent

 (mrem)b

Collective Effective

Dose Equivalenta

(person-rem)b

Distribution and transport <0.001c 20

Routine use

 Radar systems

 Microwave oven

Total

0.2d

<0.001e

200f

Landfill disposal 0.02g 40

Accidents and misuse

 Fire

 Vandalism of vacuum tube

0.02h

1i

a

 Collective doses are based on an assumed annual distribution of 10 million magnetrons in

microwave ovens containing 8 mg of natural thorium each and 100,000 other high-powered

vacuum tubes containing 0.25 g of natural thorium each. A 10-year effective lifetime also is

assumed for both the magnetrons and other vacuum tubes. Refer to text discussion for time

period for collective dose assessment.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 Dose estimate applies to a stockhandler in a distribution center or large warehouse; dose

estimates are less for semi-truck drivers, workers and customers in retail stores, and members

of the public along truck routes (see Section 3.5.4.1.1).

d

 Dose estimate applies to operators at radio and TV broadcasting stations or at radar

surveillance systems (see Section 3.5.4.2.2)

e Collective dose from routine use of one year’s distribution of all types of vacuum tubes

containing thorium (see Section 3.5.4.2.3).

f

 Dose estimate applies to use as of magnetrons in microwave ovens for homes (see Section

3.5.4.2.1). Dose estimate for a cook in a restaurant is higher (see Section 3.5.4.2.1).

g

 Dose estimate applies to waste collectors at landfills; dose estimates are less for workers at

landfills, off-site members of the public, and future on-site residents at landfills (see

Section 3.5.4.3).

h

 Dose estimate applies to a worker who is involved in cleanup following a transportation fire;

dose estimates are less for firefighters and others involved in a residential or transportation fire

(see Section 3.5.4.4).

i

 Dose estimate applies to inhalation of 220Rn and its short-lived progeny by a person who finds

and vandalizes a vacuum tube containing 2 g of natural thorium (see Section 3.5.4.4).
3–103

3.6 Welding Rods Containing Thorium

3.6.1 Introduction

In 10 CFR 40.13(c)(1)(iii), persons who receive, possess, use, or transfer welding rods

containing any amount of thorium are exempted from licensing requirements for source

material. An exemption for thoriated tungsten containing not more than 3% by weight of

thorium, but with no restrictions on the products or devices in which the exempted material

could be used, was published on March 15, 1949 (14 FR 1156). The exemption in its present

form, which applies only to welding rods but with no limit on the amount of thorium that may be

contained in this product, was proposed on September 7, 1960 (25 FR 8619), and was issued

as a final rule on January 14, 1961 (26 FR 284).

The Federal Register notices establishing the present exemption for thorium in welding rods

cited above states that the exemption would not result in an unreasonable hazard to life or

property. However, quantitative information on radiation doses to the public associated with the

exemption was not provided. Assessments of radiological impacts on members of the public

from distribution, use, and disposal of welding rods containing thorium have been performed by

McDowell-Boyer (NUREG/CR–1039) and Buckley et al. (NUREG/CR–1775). The assessment

by Buckley et al. relied extensively on the previous results of McDowell-Boyer, which also were

adopted in a review by the National Council on Radiation Protection and Measurements

(NCRP 95). Results of the previous dose assessments are described in Section 3.6.3.

3.6.2 Description of Items

Tungsten inert-gas (TIG) arc welding is a process in which an electrical arc is struck between

an inert, gas-cooled, nonconsumable electrode (i.e., an electrode that does not provide filler

material), also called a welding rod, and the metal work pieces. The electrical arc heats the

work pieces and causes them to coalesce (i.e., melt together). Most TIG welding machines are

used in automated processes that do not require the continuous presence of an operator. 

However, a significant amount of manual TIG welding, which requires the presence of an

operator, also is conducted.

Many electrodes used in TIG welding consist of tungsten wire that contains thorium dioxide

(ThO2) or another metal oxide (e.g., magnesium, lanthanum, cerium, zirconium, or yttrium). 

The metal oxide additions provide several benefits, including increased electron emissivity,

current-carrying capacity, resistance to contamination of the electrode, and an increased useful

life by decreasing the rate of electrode erosion and promoting retention of the desired tip

geometry. These characteristics result in easier initiation of the arc, promotion of arc stability

during welding operations, and a reduction in the frequency of electrode-tip regrinding and

electrode replacement.

The TIG welding process is one of the more expensive welding processes and, thus, is used

primarily in industries that require high-quality, low-contamination welds. Examples include the

aircraft industry, the petrochemical industry, manufacturing of food processing equipment, and

nuclear power plant construction and maintenance. Other industries that use TIG welding to a

lesser extent include gas and oil piping and ship building (NUREG/CR–1775).
3–104

Although the exemption allows any amount of thorium in welding electrodes, the industry

standard is nominally 1 to 2% by weight of ThO2, and most electrodes in current use appear to

contain about 2% by weight of ThO2. Thoriated-tungsten electrodes are produced in nine

standard diameters between 0.25 and 6.35 mm, and in six standard lengths between 7.6 and

61 cm. In the assessment by McDowell-Boyer (NUREG/CR–1039), a typical electrode was

assumed to be 2.4 mm in diameter and 15.2 cm in length, and to contain 0.23 g of thorium.

The number of thoriated-tungsten welding electrodes in current use in the United States is not

known. Based on information available at the time, Buckley et al. (NUREG/CR–1775)

estimated an annual production of 5.2 million such electrodes. However, the annual production

may have declined since that time (Hedrick, 1985; Hedrick, 1991) due to the increased use of

other metal oxides in welding rods. The assessment by McDowell-Boyer (NUREG/CR–1039)

assumed an annual production of 1 million electrodes containing thorium. The assessment by

Jankovic et al. (1999) reported the U.S. annual production at 4 to 5 million.

3.6.3 Summary of Previous Analyses and Assessments

A variety of scenarios for routine distribution, use, and disposal of thoriated-tungsten welding

rods, as well as certain scenarios for accidents and misuse, were considered in previous

analyses and assessments by McDowell-Boyer (NUREG/CR–1039) and Buckley et al.

(NUREG/CR–1775). The thorium content of the welding rods was assumed to be 2% by

weight, and the thorium was assumed to have been aged for 20 years after chemical

separation. This section summarizes the results of the previous assessments. Unless

otherwise noted, the assessment by Buckley et al. was based on the results obtained previously

by McDowell-Boyer. Therefore, the results of Buckley et al. are discussed only when the

exposure scenarios differ from those assumed by McDowell-Boyer.

Additionally, there are four studies performed in the past decade of the airborne thorium activity

caused by thoriated-tungsten welding operations that are considered pertinent to this study and

are summarized below.

3.6.3.1 Recent Airborne Activity Measurements

3.6.3.1.1 Study by Ludwig et al.

For routine use of thoriated-tungsten welding rods, Ludwig et al. (1999) conducted field tests in

26 different shops in Germany. About half of the study dealt with electrodes containing twice

the amount of thorium currently believed to be in use in the United States (4% versus 2%). 

Samples included some for both alternating current (AC) and direct current (DC) welding. The

airborne activity generated through welding and grinding of electrodes was measured using

personal air samplers and stationary air samplers that collected the inhalable (respirable) dust

fraction. For welding, a modified air sampler was integrated into the protective shield, and

samples were taken inside the shield during work including preparation time, but not grinding. 

Samples were collected separately during grinding. Volume flow rates were selected to best

match inhalable fraction conventions. Ludwig et al. provided data about the presence or

absence of suction during welding and grinding, annual working times, the number of instances

of grinding per working hour, AC or DC power supply, and the assayed quantities of 232Th, but

did not include information on general ventilation rates. The airborne levels of 232Th during
3–105

welding as measured by Ludwig et al. ranged from a low of less than 0.1 to 91 millibecquerel

(mBq)/m3 (<7×109 to 5 ×106 microcurie (Ci)/m3) with a geometric mean of about 0.9 mBq/m3

(3×108 Ci/m3). For measurements where suction was not used (i.e., no local exhaust), the

geometric mean 232Th airborne level was 1.2 mBq/m3

 (3.2 ×108 Ci/m3). The data did not

show significant differences in airborne levels for 2% versus 4% thorium rods; however, AC

welding yielded results that were about a factor of 30 higher than DC welding. For AC welding,

the geometric mean airborne level without local exhaust was 10 mBq/m3

 (2.8×107 Ci/m3); and

for DC welding the geometric mean was 0.34 mBq/m3

 (9.2×109 Ci/m3).

Ludwig et al. (1999) reported that in a room with volume of about 100 m3

, and without any

ventilation or suction system, 35 electrodes (45 by weight thorium) were ground in 15 minutes. 

Their graph indicates an activity concentration for 232Th during the grinding of about

180 mBq/m3

 (5×106 Ci/m3) with the airborne activity concentration decreasing soon after the

end of grinding. The only case from the Ludwig et al. (1999) study that involved local exhaust

ventilation indicated a reduction factor of 100.

The study estimated levels of 232Th intakes for welders from less than 0.1 to 144 Bq/yr

(<2.7×103

 to 3.9 nCi/yr) during welding and from 0.02 to 30.2 Bq/yr (5.4×104 to 0.82 nCi/yr)

during grinding. In six cases the estimated total intake was estimated to exceed the annual limit

for intake to the public of thorium in oxide form, as derived from ICRP Publication 71 (ICRP 71),

based on the projections from limited sample data. Four of the six cases involved welders,

working with alternating current, where the main exposure was caused by the welding process. 

The other two welders used direct current welding with the grinding causing the estimated

intake to exceed the intake limit.

3.6.3.1.2 Study by Vinzents et al.

The study by Vinzents et al. (1994) was performed to estimate airborne levels while grinding

thoriated-tungsten welding rods. To provide sufficient sample size, the measurement was

made during the grinding of 4% thorium electrodes for 139 minutes by one person. No exhaust

hood was used and the jet of the grinding was directed towards the floor. Vinzents determined

a 232Th respirable airborne mean concentration of 440 mBq/m3

 (2×105 Ci/m3) (with

approximately 0.3 respirable fraction). However, he acknowledged the limited data and

believed it to represent a worst case estimate. Vinzents reported that electrodes are typically

ground once each day for approximately one minute. 

3.6.3.1.3 Study by Crim and Bradley 

Crim and Bradley (1995) performed measurements of airborne thorium concentrations during

the grinding of 2% thorium oxide welding rods. In order to collect sufficient sample size, the

total grinding time was 6 minutes which is substantially longer than the reported typical grinding

time of 30-90 seconds. Air samples were collected to the sides during one grinding and above

and to the side during another. The range for the 4 air samples was 1.3 - 300 mBq/m3

(3.6×108 - 8.2×106 Ci/m3) with a geometric mean of 23 mBq/m3 (6.3×107 Ci/m3). Breathing

zone air samples collected showed no detectable activity at a minimum detectable level of

15 mBq/m3

 (4×107 Ci/m3). The data presented by Crim and Bradley (1995) did not specify if

the represented total thorium or 232Th.
3–106

3.6.3.1.4 Study by Jankovic et al.

The study by Jankovic et al.(1999) provided results of air sampling conducted during 15

different welding operations with 2% thorium; four (4) with dilution ventilation and air samplers

outside the welding helmet, two (2) with dilution ventilation and air samplers inside the welding

helmet, four (4) with local exhaust and air samplers outside the helmet, and five (5) with local

exhaust and air samplers inside the helmet. These operations were all carried out with direct

current.

The data for welding with only dilution air yielded the largest values for airborne 232Th with the

geometric mean of 0.7 mBq/m3

 (2×108 Ci/m3) for the sample outside the helmet and

0.15 mBq/m3

 (4×109 Ci/m3) for the sample inside the helmet. With local exhaust, the values

were 0.04 mBq/m3

 (1×109 Ci/m3) for the sample outside the helmet and 0.1 mBq/m3

(3×109 Ci/m3) for the sample inside the helmet. The geometric mean for all sample data was

0.15 mBq/m3

 (4×109 Ci/m3) unadjusted for respirable fraction (10 m or less). Jankovic

reported an activity median aerodynamic diameter (AMAD) of welding aerosols of 3.5 m at

30 cm from the point of operation. Measurements during grinding with no local exhaust ranged

from 14 to 190 mBq/m3

 (3.8×107 to 5.1×106 Ci/m3) with a geometric mean of 52 mBq/m3

(1.4×106 Ci/m3), unadjusted for a measured 60% respirable fraction. The respirable fraction

for electrode sharpening was 60% (10 m or less) and the AMAD was 5 m in the breathing

zone. This fraction is somewhat higher than that reported by Vinzents but could easily be a

function of sampling method (i.e., effects of gravitational settling relative to sampler location).

3.6.3.2 Assessment by McDowell-Boyer

3.6.3.2.1 Distribution and Transport

Doses from distribution and transport of thoriated-tungsten welding rods were estimated by

McDowell-Boyer (NUREG/CR–1039), based on a methodology essentially the same as that

presented in Appendix A.3 of this report. Distribution and transport were assumed to involve

handling by parcel-delivery workers, over-the-road trucking workers, and retail store workers;

also considered were exposures to customers in retail stores. Estimates of individual and

collective doses resulting from an annual distribution of 1 million electrodes obtained by

McDowell-Boyer are summarized in Table 3.6.1. 

During distribution and transport, external exposure to thorium and its decay products was

assumed to be the only exposure pathway of concern. For external exposure, the dose to the

whole body in Table 3.6.1 was essentially the same as the effective dose equivalent (EDE). For

packaged electrodes, there is no credible scenario for ingestion or inhalation of radionuclides in

particulate form. The decay product 220Rn could be released from the welding rods, and

inhalation exposure to radon and its short-lived decay products could occur. However, for the

worst-case assumption that all radon produced in the electrodes would be released into the air,

the estimated annual individual dose equivalents to the lungs for all scenarios were less than

1×1010 millisievert (mSv) (<1×108 mrem) and, thus, were many orders of magnitude less than

the estimated doses from external exposure.
3–107

3.6.3.2.2 Routine Use

For routine use of thoriated-tungsten welding rods, McDowell-Boyer (NUREG/CR–1039)

considered scenarios describing exposure to heavy-use welders, including welding at home as

well as in the workplace, occasional-use welders, and nonwelders in the vicinity of welding rods

during use. Exposures during grinding of electrodes by welders also were considered.

Assumptions about release of radioactive material during welding and grinding used by

McDowell-Boyer include the following: (1) an average loss rate of thorium during welding with

normal rods of 0.8 to 6.0 mg/h, (2) an average loss rate of thorium during welding with

contaminated rods (either oxidized or contaminated with the weld metal) 10 times higher than

for normal rods, with welding using contaminated rods occurring 10% of the time, (3) a percent

of material released during welding that is present in the fumes of 0.2 to 1%, and (4) an

average loss of thorium during a single grinding of 8 to 16 mg. In addition, the ventilation rate

in welding shops was assumed to be 3420 m3

/h, in accordance with an American National

Standards Institute standard for a welder’s workplace (ANSI Z49.1–1973). The ventilation rate

in the work area in a home was assumed to be 230 m3

/h for a normal-sized room with an air

turnover rate of 1 change per hour. Finally, 25,000 welders were assumed to be exposed to the

1 million electrodes used annually, with 10% of the population assumed to use thoriatedtungsten electrodes for all welds and 90% of the population assumed to use thoriated-tungsten

electrodes for 50% of the welds.

Based on the assumptions described above, the estimates of individual and collective doses to

bone from inhalation of thorium and its decay products, obtained by McDowell-Boyer

(NUREG/CR–1039), are summarized as follows (bone is the critical organ for inhalation of

thorium):

For 12,500 heavy-use welders who are assumed to be exposed to air concentrations of

thorium of 0.048 to 1.8 g/m3

 for 900 hours while welding in a work shop, 12 to

460 g/m3

 for 4.2 hours while grinding tips in the work shop, 0.48 to18 g/m3 for

100 hours while welding with a contaminated electrode in the work shop (50 of which

are also assumed to be exposed to concentrations of 0.7 to 26 g/m3

 for 180 hours

while welding in a home shop), and 7 to 260 g/m3

 for 20 hours while welding with a

contaminated electrode in the home, the maximum annual individual dose equivalent to

bone would be 0.2 to 10 mSv (20 to 1,000 mrem), the average annual individual dose

equivalent to bone would be 0.04 to 1.6 mSv (4 to 160 mrem), and the annual collective

dose equivalent to bone would be 0.5 to 20 person-Sv (50 to 2,000 person-rem).

For 12,500 occasional-use welders who are assumed to be exposed to air

concentrations of thorium of 0.048 to 1.8 g/m3

 for 180 hours while welding in a work

shop, 12 to 460 g/m3

 for 0.83 hour while grinding tips in the shop, and 0.48 to 18 g/m3

for 20 hours while welding with a contaminated electrode in the shop, the maximum

annual individual dose equivalent to bone would be 0.02 to 0.6 mSv (2 to 60 mrem), the

average annual individual dose equivalent to bone would be 0.009 to 0.3 mSv (0.9 to

30 mrem), and the annual collective dose equivalent to bone would be 0.1 to

4 person-Sv (10 to 400 person-rem).

For 75,000 nonwelders who are assumed to be exposed to air concentrations of thorium

of 0.048 to 1.8 g/m3

 for 540 hours during normal welding and 0.48 to 18 g/m3 for
3–108

60 hours during welding with contaminated electrodes, the maximum annual individual

dose equivalent to bone would be 0.20 to 1.5 mSv (20 to 150 mrem), the average

annual individual dose equivalent to bone would be 0.07 to 0.5 mSv (7 to 50 mrem), and

the annual collective dose equivalent to bone would be 5 to 40 person-Sv (500 to

4,000 person-rem).

Ranges of estimated doses reflect the assumed ranges in the release of thorium from

electrodes during welding or grinding and the fraction of the releases from welding that would

be present in the inhaled fumes.

Doses to the whole body from inhalation of thorium and its decay products also were calculated

by McDowell-Boyer (NUREG/CR–1039). These doses were about an order of magnitude less

than the doses to bone given above. However, this calculation of dose to the whole body from

internal exposure represented the total energy absorbed in all body tissues divided by the

nominal body mass of 70 kg. The calculation is not a meaningful quantity when highly

nonuniform irradiations of the whole body occur, as is the case for inhaled thorium. This

calculated dose to the whole body for internally deposited thorium is not related to the EDE and

does not provide a meaningful basis for comparison.

McDowell-Boyer (NUREG/CR–1039) also estimated external doses to the whole body for

heavy-use welders, occasional-use welders, and nonwelders in the vicinity of welding rods

during use. As noted previously, dose to the whole body from external exposure to aged

thorium essentially was the same as the EDE. These dose estimates are summarized as

follows:

For 12,500 heavy-use welders who are assumed to be exposed at a distance of 30 cm

during welding or grinding, 50 of whom are also assumed to be exposed at a distance of

90 cm during welding at home, the maximum annual individual dose equivalent to the

whole body would be 1.5×103

 mSv (0.15 mrem), the average annual individual dose

equivalent would be 8×104

 mSv (0.08 mrem), and the annual collective dose equivalent

would be 0.01 person-Sv (1 person-rem).

For 12,500 occasional-use welders who are assumed to be exposed at a distance of

30 cm during welding or grinding, the maximum annual individual dose equivalent to the

whole body would be 8×105

 mSv (0.008 mrem), the average annual individual dose

equivalent would be 5×105

 mSv (0.005 mrem), and the annual collective dose

equivalent would be 6×104

 person-Sv (0.06 person-rem).

For 75,000 nonwelders who are assumed to be exposed at a distance of 3.6 meters

while working in a shop, the maximum annual individual dose equivalent to the whole

body would be 2×105

 mSv (0.002 mrem), the average annual individual dose equivalent

would be less than 1×105

 mSv (<0.001 mrem), and the annual collective dose

equivalent would be 6×104

 person-Sv (0.06 person-rem).

The importance of external exposure compared with inhalation exposure during routine use of

welding electrodes can be evaluated as follows. For inhalation of ThO2, which is assumed to be

Class Y material, the ratio of the EDE to the dose equivalent to bone is about 0.06

(EPA–520/1–88–020). If the doses to bone calculated by McDowell-Boyer (NUREG/CR–1039)

and summarized above are multiplied by this factor, the resulting EDEs are approximately an
3–109

order of magnitude or more greater than the calculated doses to the whole body (i.e., EDEs)

from external exposure. Therefore, based on the assessment by McDowell-Boyer, inhalation

exposure appears to be considerably more important than external exposure during routine use

and the grinding of welding electrodes.

3.6.3.3 Disposal

Individual and collective doses from disposal of thoriated-tungsten welding rods in landfills or by

incineration were estimated by McDowell-Boyer (NUREG/CR–1039) and Buckley et al.

(NUREG/CR–1775). Different approaches or assumptions were used in the dose calculations

for disposal in the two assessments. Therefore, results from both assessments are

summarized in the following sections.

3.6.3.3.1 Assessment by McDowell-Boyer

McDowell-Boyer (NUREG/CR–1039) assumed disposal of 1 million welding electrodes per year,

with 50% distributed equally among 100 landfills and 50% distributed equally among

100 incinerators. Doses were estimated for individuals residing 1 km from a landfill or

incinerator, and for an assumed population within a distance of 72 km. The amount of thorium

and its decay products present in the welding rods was assumed to be equal to the initial

amount (2% by weight of thorium, with the material aged for 20 years), less the total amount

assumed to be lost during welding and grinding. Dose estimates for disposal are summarized

as follows.

For disposal in landfills, doses were estimated only for releases of 220Rn, based on an

assumption that all of the radon in the welding rods would be released to the atmosphere

following disposal. The estimated annual individual dose equivalent to the lungs from inhalation

of radon and its short-lived decay products was 2×104

 mSv (0.02 mrem). The estimated

annual collective dose equivalent was 0.03 person-Sv (3 person-rem). Estimated doses to

bone from ingestion pathways were more than four orders of magnitude lower.

For disposal by incineration, doses were estimated by assuming that all of the thorium and its

decay products in the welding rods would be released to the atmosphere. The estimated

annual individual dose equivalent to bone from inhalation and ingestion was 0.002 mSv

(0.2 mrem). The estimated annual collective dose equivalent was 2.2 person-Sv

(220 person-rem). About 85% of the dose from incineration was due to ingestion of

contaminated food following deposition of the atmospheric release, with the remainder due to

inhalation.

3.6.3.3.2 Assessment by Buckley et al.

Buckley et al. (NUREG/CR–1775) considered doses from waste collection, disposal in landfills,

and incineration. An annual disposal of 5.2 million welding rods was assumed, but the

estimates of collective dose summarized below have been adjusted to represent an annual

disposal of 1 million. In all disposal scenarios, the amount of thorium remaining in the welding

rods was assumed to be one-half of the initial content.

For waste collectors exposed to welding rods in municipal refuse, the annual individual dose

equivalent to the whole body from external exposure was estimated to be less than 
3–110

1×106

 mSv (<1×104 mrem), and the annual collective dose equivalent was less than

1×104

 person-Sv (<0.01 person-rem). Doses from inhalation of radionuclides in particulate

form or inhalation of radon and its short-lived decay products were not estimated, but were

presumed to be insignificant compared with the doses from external exposure.

For disposal in a landfill, collective doses were estimated for a nearby population resulting from

releases of thorium to groundwater. Based on the worst-case” assumption that the release

rate of thorium from the landfill would be 1% per year for previously incinerated sources or

0.1% per year for intact sources, the collective dose equivalent to bone from disposal of

1 million welding electrodes was estimated to be 400 person-Sv (40,000 person-rem) over a

1000 year period. Individual doses were not estimated for this scenario.

For disposal by incineration, individual and collective doses in a nearby population were

estimated by assuming that one-half of the material incinerated would be released to the

atmosphere. The estimated annual individual dose equivalent to bone from incineration of

1 million welding rods was 2×104

 mSv (0.02 mrem), and the annual collective dose equivalent

was 2 person-Sv (200 person-rem). Doses from inhalation of radon and its short-lived decay

products were several orders of magnitude lower. The individual dose estimate is about an

order of magnitude greater than the previous estimate by McDowell-Boyer (NUREG/CR–1039)

for essentially the same scenario, but the collective doses estimated in the two assessments

are nearly the same.

3.6.3.4 Accidents or Misuse

Individual doses resulting from accidents or misuse involving thoriated-tungsten welding rods

were considered by McDowell-Boyer and Buckley et al. Assumed exposure scenarios and

resulting dose estimates are summarized in the following sections.

3.6.3.4.1 Assessment by McDowell-Boyer

McDowell-Boyer (NUREG/CR–1039) considered an accident scenario involving a warehouse

fire and a scenario for misuse involving carrying extra welding electrodes in a shirt pocket. The

scenario for a fire assumed that a warehouse would contain 5000 electrodes, which is the same

number that was assumed to be incinerated at any site (see Section 3.6.3.3.1). Assuming

complete volatilization of the thorium and its decay products in a fire, the inhalation dose to a

member of the public near the burning warehouse would be about the same as the dose to an

individual near an incinerator. Thus, the estimated dose equivalent to bone for an individual

near a warehouse fire was 0.002 mSv (0.2 mrem). Doses to persons in the warehouse also

were discussed but were believed to be insignificant because of the difficulty in volatilizing

thorium at temperatures obtained in a small fire and the likelihood of rapid evacuation of the

warehouse after a fire starts. Doses to firefighters during the fire were not considered.

For the scenario for misuse, an individual was assumed to carry three welding rods in a shirt

pocket for 8 h/day and for 250 days (1 working year). The estimated dose equivalent to the

maximally exposed portion of the skin for this scenario was 0.04 mSv (4 mrem).
3–111

3.6.3.4.2 Assessment by Buckley et al.

Buckley et al. (NUREG/CR–1775) considered a fire in a warehouse containing 50,000 welding

electrodes. Doses to firefighters were estimated based on the conservative assumptions of

exposure for 8 hours and no use of respiratory protection. The estimated dose equivalent to

bone during the warehouse fire was 1.3 Sv (130 rem).

3.6.4 Present Exemption Analysis

This section presents estimates of doses to members of the public from distribution and

transport, routine use, and disposal of welding electrodes containing thorium. Estimates of

dose also are presented for accident or misuse scenarios involving a fire in a storage

warehouse or the inadvertent carrying of welding electrodes in a shirt pocket.

Based on the current practices discussed in Section 3.6.2, each thoriated-tungsten welding rod

is assumed to contain 2% by weight of thorium, unless otherwise stated. Furthermore, based

on a survey of welders at Oak Ridge National Laboratory (ORNL) performed for this

assessment (Jankovic et al.), a typical welding rod is assumed to be 18 cm in length and

0.24 cm in diameter. Thus, for a density of tungsten of 19.35 g/cm3

, each welding rod is

assumed to contain 0.3 g of thorium. The assumed thorium content per electrode is 30%

greater than the value assumed previously by McDowell-Boyer (NUREG/CR–1039), due to the

greater assumed length of an average electrode. Except as noted in the dose assessment for

disposal, contributions to dose from thorium decay products are taken into account in all

exposure scenarios by assuming that the thorium has been aged for 20 years after chemical

separation.

The number of thoriated-tungsten welding rods distributed annually is not known, but is

assumed in this assessment to be 5 million. The assumed value could be too low, based on

information that suggests the annual distribution may be about 10 million (Phone call,

G. Goodwin, ORNL, Oak Ridge, TN, June 1997). However, the higher value represents foreign

as well as domestic sales, and the fraction sold domestically was not provided. The individual

doses for distribution and transport and disposal and the collective doses for all scenarios

estimated in this assessment can be adjusted for the actual distribution if it is known.

3.6.4.1 Distribution and Transport

As described in Section 3.6.3.1, the doses from distribution and transport of thoriated-tungsten

welding rods were estimated previously by McDowell-Boyer (NUREG/CR–1039) using a

methodology that is essentially the same as that presented in Appendix A.3 of this report. 

Thus, the previous results are considered suitable for the present assessment. The scenario

for distribution and transport of welding rods assumed by McDowell-Boyer is described as

follows. Of the total number of welding rods distributed, 80% are assumed to be shipped by

truck to United Parcel Service (UPS) terminals over distances greater than 400 km, 10% to

UPS terminals over distances between 32 and 400 km, and 10% directly to warehouses and

retail stores over distances less than 32 km. Of the welding rods shipped to UPS terminals,

80% then are assumed to be shipped to warehouses and 20% to retail stores. Thus, for the

assumed total distribution of 5 million welding rods per year, each of 80 warehouses is
3–112

assumed to receive twenty-five shipments of 2,000, or a total of 50,000 per year, and each of

200 retail stores is assumed to receive ten shipments of 500, or a total of 5,000 per year.

During distribution and transport, only external exposures are assumed to occur. There are no

credible scenarios by which radionuclides in particulate form in packaged electrodes could be

inhaled or ingested. Releases of 220Rn could occur, resulting in inhalation exposure to the

radon and its short-lived decay products. However, an assessment by McDowell-Boyer

(NUREG/CR–1039), based on the conservative assumption that all of the radon would be

released from the packaged electrodes, indicated that the dose from exposure to radon would

be many orders of magnitude less than the dose from external exposure (see Section 3.6.3.1). 

Therefore, doses from inhalation or ingestion during distribution and transport can be

neglected.

The dose estimates for external exposure during distribution and transport that are adopted in

the present assessment are obtained from the results of McDowell-Boyer, given in Table 3.6.1,

by increasing the doses by 30% to reflect the higher thorium content per welding rod assumed

in this assessment. Collective doses are also increased by a factor of 5 to reflect current

distribution. In estimating individual dose, the maximum values given in Table 3.6.1 are used. 

As noted previously, the dose to the whole body from exposure to thorium and its decay

products is essentially identical to the EDE. Results are summarized as follows:

The annual EDE to individual truck drivers, UPS terminal employees, retail employees,

and warehouse employees would be 8×104 to 0.003 mSv (0.08 to 0.3 mrem), with the

dose to truck drivers being the highest. The dose to individual customers in retail stores

would be considerably less.

The collective EDE for 5 million welding rods distributed per year, including the dose to

truck drivers, UPS terminal employees, retail store employees, warehouse employees,

and customers in retail stores, would be 0.05 person-Sv (5 person-rem).

3.6.4.2 Routine Use

Doses from routine use of thoriated-tungsten welding rods would be experienced primarily

during welding and grinding (sharpening) of the electrodes. This assessment considers

individual and collective doses during these activities.

3.6.4.2.1 Welding in Shops

Professional welders working in a shop should receive the highest individual doses during

welding. In this assessment, inhalation doses for welders are estimated based primarily on the

results of a study of air concentrations by Ludwig et al. (1999). In addition, data is used from a

study by Jankovic et al. (1999) which included measurements of airborne concentrations of

232Th around and inside the welders’ masks, measurements of the particle size distribution of

the total aerosol, and modeling of airborne concentrations of 232Th. Doses from external

exposure during welding also are estimated, based on the results of a previous assessment by

McDowell-Boyer (NUREG/CR–1039) discussed in Section 3.6.3.2.

These studies indicate that the dose to a welder from inhalation exposure depends on several

factors, including (1) the release rate of thorium and its decay products from the electrode
3–113

during welding, (2) the fraction of the working time spent welding, (3) the effect of the welder’s

face mask on concentrations of airborne particulates in the breathing zone, (4) diffusion and

sedimentation of airborne particulates, (5) the ventilation rate in the welding area, (6) the

presence or absence of local exhaust ventilation at the welding site, (7) the distribution of

particle sizes in the inhaled aerosols, and (8) the weight % of thorium in the welding electrode.

Based on the concentration measurements from Ludwig et al. (1999) for DC welding using 2%

and 4% by weight ThO2, the geometric mean air concentration of 232Th in the breathing zone

area of a welder, without the benefit of local exhaust ventilation, would be about 0.34 mBq/m3

(9.2×109 Ci/m3). This value is essentially the same as that measured by Jankovic, where the

geometric mean for similar measurements was 2×108 Ci/m3

 uncorrected for a 45% respirable

fraction. As measured by Ludwig et al. (1999), AC welding yielded results that were about a

factor of 30 higher than DC welding. For AC welding, the geometric mean airborne level

without local exhaust was 10 mBq/m3

 (2.8×107 Ci/m3).

Because of the differences in the reported data, an AMAD particle size of 1 m was selected

for this dose assessment. This value is considered reasonable because there is only about a

35% change in the inhalation coefficient from an AMAD of 0.2 m to 1 m or from 1 m to

3.5 m.

Based on the measurements and assumptions described above, the inhalation dose to a welder

who is assumed to be exposed while welding for 4 hours/day and 250 days/yr, for a total

exposure time of 1000 h/yr, is estimated as follows. For thorium that has been aged for 20

years following chemical separation, the activity of 228Th is 87% of the activity of 232Th and no

other decay products of 232Th contribute significantly to the inhalation dose (see tables 3.1.5

and 3.1.7). For inhalation of thorium in oxide form, which is Class Y material, and for an AMAD

of 1 m, the dose coefficients are 3.1×104

 Sv/Bq (1.15×103 rem/Ci) for 232Th and

9.23×105

 Sv/Bq (3.42×102 rem/Ci) for 228Th (EPA–520/1–88–020). For a breathing rate

appropriate for light activity of 1.2 m3

/h (see Appendix A.1), the resulting EDE from inhalation

exposure would be 0.2 mSv (20 mrem). For AC welding, the annual dose is calculated to be

5 mSv (500 mrem), using a geometric mean airborne level of 10 mBq/m3 (2.8 ×107 Ci/m3

) as

measured by Ludwig et al. (1999).

The use of local exhaust ventilation at the welding site, in addition to normal room ventilation, is

potentially important. The study by Ludwig et al. (1999) and Jankovic et al. (1999) suggests

that local exhaust ventilation, which is used in some welding shops, would reduce average

inhalation exposure to welders. Based on the summary of measurements presented in Section

3.6.3.1 , the average welder breathing zone concentration is estimated to be a factor of 10 less

if local exhaust is utilized, which would yield a calculated dose of 10 lower than that estimated

above.

Estimating inhalation dose obtained above is subject to considerable uncertainty attributable in

part to limited and often inconsistent data. Also, there is a great deal of variation in the

concentrations of thorium that occur during welding under a variety of actual conditions. In the

study by Ludwig et al. (1999), for data evaluating 26 welding shops, the concentrations varied

by a factor of about 900 from low to high and no data were presented regarding ventilation flow

rates. 
3–114

The type and amperage of the current during welding are also sources of variability in airborne

exposures during welding. Increasing the welding arc current from 100 to 250 amperes (A)

causes the electrode to be consumed faster and significantly increases the thorium airborne

concentrations (according to Jankovic et al., a factor of seven higher at 250A). The study by

Ludwig et al. (1999) did not address arc current but the data indicates that alternating current

(AC) welders can be subjected to airborne concentrations that are about a factor of 30 above

those made with direct current (DC) welding.

Another potential uncertainty in the actual dose is whether significant activity of 230Th is in the

welding rods due to the presence of uranium in the ores from which the thorium was obtained. 

The presence of 230Th would increase the dose estimate obtained in the present assessment,

because the estimate is based on measurements of airborne concentrations of 232Th. However,

in a set of measurements on electrodes from three different manufacturers, Jankovic et al.

(1999) concluded that the mass abundance of 230Th in the welding rods used in their studies

was less than 1 ppm. Therefore, based on the half-lives given in Tables 3.1.2 and 3.1.3, the

activity of 230Th would be less than 20% of the activity of 232Th. Based on the relative activities

of the two isotopes and the dose coefficients for inhalation given in Tables 3.1.6 and 3.1.7, the

dose from inhalation of 230Th then would be less than 4% of the dose from inhalation of 232Th

and its decay products. This result is not particularly sensitive to the clearance class for the

inhaled aerosols (EPA–520/1–88–020). Therefore, if the welding rods used in the studies by

Jankovic et al. (1999) are representative of electrodes in use at the present time, the dose

contribution from 230Th would be insignificant.

Welders also would receive an external exposure during welding. Based on the assumed

welding time of 4 h/day and an assumed average distance from the electrode of 30 cm, the

dose to the whole body estimated by McDowell-Boyer (NUREG/CR–1039), when increased by

30% to take into account the higher amount of thorium per welding rod assumed in this

assessment, would be 0.002 mSv/yr (0.2 mrem/yr). Thus, as expected, external doses to

welders should be insignificant compared with inhalation doses.

3.6.4.2.2 Welding in Homes

In this assessment, doses to an occasional home welder are estimated based on (1) the dose

estimate for a welder in a shop given above and, (2) assumptions about the exposure time for a

home welder. The home welder is assumed to be exposed during a single 4-hour work session

per week, for a total of 200 h/yr (NUREG/CR–1039). The assumed exposure time in a home is

a factor of 5 less than the assumed exposure time in a shop. Therefore, if the AMAD is

assumed to be the same in the two work environments, and without the benefit of a local

exhaust system, the EDE to a home welder would be about a factor of 5 lower than the dose to

a welder in a shop obtained previously, or about 1 mSv (100 mrem) for AC welding and

0.03 mSv (3 mrem) for DC welding. Use of local exhaust is estimated to reduce the doses by a

factor of 10. 

3.6.4.2.3 Grinding

Inhalation exposures also can occur during grinding (sharpening) of welding rods. Two types of

individuals may grind electrodes: (1) dedicated grinders who spend most of their working time

at this activity in support of many welders, and (2) welders who typically grind electrodes only a

few times per day for their own use. The latter situation presumably is far more common. 
3–115

However, Sinclair and Thind (1992) observed two workers at a nuclear construction site who

spent most of their time grinding as many as 250 electrodes per day. Doses for both groups of

workers are considered in this assessment.

Doses to welders who grind their own electrodes are estimated as follows. Ludwig et al. (1999)

reported an activity concentration for 232Th during the grinding of about 180 mBq/m3

(5×106 Ci/m3) with the airborne activity concentration decreasing soon after the end of

grinding. Crim and Bradley reported an average airborne concentration during grinding of

23 mBq/m3

 (6.3×107 Ci/m3) and Jankovic et al. reported a value of 52 mBq/m3

(1.4×106 Ci/m3). Vinzents reported value of 440 mBq/m3 (1.2×105 Ci/m3) is considered

atypical due to the method utilized for the measurement. Based on these data, an average

airborne activity for grinding of 85 mBq/m3

 (2.3×106 Ci/m3) is determined. 

Ludwig et al. (1999) reported that studies performed by others indicated that the AMAD of

grinding particles is about 1 m. Jankovic reported a value of 5 m. The concentration and

measured particle size would depend upon many factors such as: the presence of any local

exhaust ventilation, the grinding wheel or belt and its direction and speed, the presence of any

particle guards, the application pressure, and the required grinding time. An AMAD size of

1 m was selected for this dose assessment. This value is considered reasonable as there is

an increase in the inhalation factor by only about 1.6 from an AMAD range of 10 to 0.2 m. 

The annual EDE for these assumptions for a welder performing his own sharpening would be

0.2 mSv (20 mrem). Use of local ventilation is estimated to reduce this dose by about a factor

of 10. As in the case of welding, uncertainty in the assessments of estimated dose during

grinding is due to variables in the airborne concentration of thorium. 

Occasional home welders also would grind welding electrodes for their own use. In the

assessment of exposures in a home, the inhalation dose for welders was estimated to be a

factor of 5 times lower due to the reduced work time. This ratio also should apply to grinding in

a home shop. Therefore, the estimated EDE to a home welder from grinding electrodes is

0.03 mSv (3 mrem) to the home welder.

Doses to dedicated grinders are estimated based on the average airborne activity described

above during grinding by normal welders without benefit of local exhaust ventilation. The

sharpening efficiency of a dedicated grinder is estimated to be 20 seconds per sharpening. 

This is about one third of the time typically reported for a welder sharpening his own welding

rod. For a hypothetical large welding operation consisting of 50 welders, about 150 welding

rods would have to be sharpened per work day. For one person performing the sharpening this

amounts to 200 hours per year. Assuming the same AMAD, inhalation dose coefficient, and

breathing rate as in the analysis for grinding by welders given above, the estimated EDE to

dedicated grinders would be about 8 mSv (800 mrem). All of the uncertainties discussed above

apply to this scenario. Parameters were developed to be on the conservative side, with certain

of them set at the maximum expected, e.g., no local exhaust ventilation. This scenario was

developed to review the significance of this exposure pathway.

The average air concentration of thorium over an 8-hour work day for the dedicated grinders,

with local exhaust ventilation, estimated by Sinclair and Thind (1992), is 1.4 g/m3

, which

corresponds to an activity concentration of 232Th of 5.6 mBq/m3

 (1.5×107 Ci/m3). Assuming

the same AMAD, inhalation dose coefficient, and breathing rate as in the analysis for grinding
3–116

by welders given above, and assuming exposure for 200 working hours per year, the estimated

EDE to dedicated grinders would be 0.5 mSv (50 mrem). This estimate is more than an order

of magnitude less than the dose for a dedicated grinder in a welding shop given above. This

reduction presumably is due primarily to the use of local exhaust ventilation in this assessment,

but not by those in the hypothetical operation described above.

3.6.4.2.4 Combined Doses from Welding and Grinding and Collective Doses 

The total dose estimate for a professional welder (without benefit of local exhaust) would be the

sum of the contributions from welding and from grinding and is 5 mSv (500 mrem) for AC

welding and 0.3 mSv (30 mrem) for DC welding. The collective dose to professional welders for

an assumed equal use of AC and DC welding and use of 4 million electrodes per year

(reserving 1 million of the 5 million annual production for home welders and automatic welders)

can be estimated as follows. Using the ratios of welders to the number of electrodes

established by McDowell-Boyer (NUREG/CR–1039), there would be 100,000 welders for

4 million electrodes (allowing a portion of the annual production to be used for home or

automated welding). The collective EDE would be 100,000 times the average of the AC and

DC welding doses or 300 person-Sv (3×104

 person-rem).

Following the model developed earlier, the total dose estimate for the home welder (without

benefit of local exhaust) would also be the sum of the contributions from welding and from

grinding and is 1 mSv (100 mrem)for AC welding and 0.06 mSv (6 mrem) for DC welding. The

collective EDE for the home welder would be 10,000 times the average of the AC and DC

welding doses or 5 person-Sv (500 person-rem).

3.6.4.3 Disposal

Following their useful lifetime, unused portions of welding rods and grinding fines from floors

may be sent to landfills or incinerators for disposal. Jankovic et al. (1999) concluded that only

3% of a rod is actually consumed in welding. Therefore, this dose assessment assumes that

disposal is made of 5 million pieces of left-over welding rods and floor fines. The quantity

disposed could be 97% of the annual distribution or about 1.5×103

 kg of thorium. Recycling of

used electrodes is considered unlikely.

In this assessment, the generic methodology in Appendix A.2 is used to estimate individual and

collective doses from disposal of welding rods containing thorium. Dose estimates obtained

previously by McDowell-Boyer (NUREG/CR–1039) and Buckley et al. (NUREG/CR–1775), and

described in Section 3.6.3.3, are not used. For disposal in landfills, McDowell-Boyer estimated

doses to off-site individuals and populations due only to releases of 220Rn, based on the

conservative assumption that all radon produced in the electrodes would be released to the

atmosphere, and Buckley et al. estimated collective doses from groundwater transport using

very conservative assumptions about thorium release rates. For incineration, McDowell-Boyer

and Buckley et al. based their estimates of individual and collective doses on the very

conservative assumption that 50 to 100% of the thorium in the electrodes would be released to

the atmosphere.

The survey of welders at ORNL performed for this assessment indicated that a typical electrode

with an initial length of about 18 cm is about 4 cm in length at the end of its useful life. All

decay products of 232Th are assumed to be present in activity equilibrium.
3–117

3.6.4.3.1 Disposal in Landfills

Of the 1.5×103

 kg of thorium assumed to be disposed per year, 80% are assumed to be sent to

landfills (see Appendix A.2). In the generic methodology described in Appendix A.2, doses are

estimated for waste collectors, workers at landfills, off-site individuals who reside near landfills

during operations or after closure, and individuals who might reside on the landfill sites after

closure. In estimating doses, no allowance is made for the non-dispersibility of the solid pieces

of welding rods which amounted to 20% of the total quantity disposed. 

Based on the assumptions described above and the generic disposal methodology in

Appendix A.2, the following estimates of individual and collective doses from disposal in landfills

are obtained:

The EDE to individual waste collectors would be 0.004 mSv (0.4 mrem). The annual

EDE to a future on-site resident would be about 0.004 mSv (0.4 mrem), and the annual

EDE to individual workers at landfills would be about 5×104

 mSv (0.05 mrem). The

dose to off-site individuals after closure of the landfills would be many orders of

magnitude less. 

The collective EDE from 1 year’s disposals of welding rods would be 4 person-Sv

(400 person-rem), due almost entirely to exposure to future on-site residents for

1000 years after facility closure. If exposure to future on-site residents are not taken

into account, the collective EDE from exposure to waste collectors and workers at

landfills would be 0.02 person-Sv (2 person-rem) per year.

3.6.4.3.2 Disposal in Incinerators

Of the 1.5×103

 kg of thorium assumed to be disposed per year, 20% are assumed to be sent to

incinerators (see Appendix A.2). In the generic methodology for disposal in incinerators

described in Appendix A.2, doses are estimated for waste collectors, workers at incinerators,

and off-site individuals who reside near the incinerators during operations. In estimating doses,

no allowance is made for the non-dispersibility of the solid pieces of welding rods which

amounted to 20% of the total quantity disposed.

Based on the assumptions described above and the generic disposal methodology in

Appendix A.2, the following estimates of individual and collective doses from disposal in

incinerators are obtained:

The EDE to individual waste collectors would be 0.02 mSv (2 mrem). The annual EDE

to individual workers at incinerators would be about three orders of magnitude less. The

dose to individual off-site residents near incinerators would be about six orders of

magnitude less.

The collective EDE from 1 year’s disposals would be 0.3 person-Sv (0.03 person-rem),

due almost entirely to exposure to waste collectors.

The estimated dose for an off-site member of the public was estimated to be less than

1×105

 mSv (<0.001 mrem) with an off-site collective EDE of 7×105 person-Sv

(0.007 person-rem).
3–118

3.6.4.4 Accidents or Misuse

As described in Section 3.6.3.4, two scenarios involving accidents or misuse of welding rods

containing thorium were considered in previous assessments by McDowell-Boyer and

Buckley et al.: (1) a fire in a storage warehouse, and (2) the carrying of extra electrodes in a

shirt pocket. Both of these scenarios are considered in the present assessment.

3.6.4.4.1 Warehouse Fire

Individual doses resulting from a fire in a warehouse containing welding rods are estimated

using the generic methodology in Appendix A.1. For this scenario, doses are estimated for

firefighters during the fire and for individuals involved in cleanup activities after the fire.

In this assessment, the number of welding rods stored in a warehouse is assumed to be

100,000, or 2% of the assumed distribution of 5 million electrodes per year. For an assumed

mass of thorium per electrode of 0.3 g, the 100,000 electrodes in a warehouse would contain

30 kg of thorium. Application of the generic methodology for a warehouse fire to this amount of

thorium gives the following results:

For a fire in a warehouse containing 100,000 welding rods, the EDE to a firefighter using

respiratory protection during the fire would be 0.001 mSv (0.1 mrem). The EDE to an

individual not using respiratory protection during cleanup after the fire would be

0.06 mSv (6 mrem).

The estimated dose to a firefighter given above is much less than the value obtained previously

by Buckley et al. (NUREG/CR–1775), as given in Section 3.6.3.4.2. The large reduction in

estimated dose, compared with the previous result, is due primarily to the assumptions that

(1) 0.01% of the thorium in welding rods would be released in a fire, instead of 100%; (2) the

firefighter would use a respirator with a protection factor of 1000, instead of no respiratory

protection; and (3) the firefighter would be exposed for 0.5 hour, instead of 8 hours.

3.6.4.4.2 Carrying of Electrodes in a Shirt Pocket

McDowell-Boyer (NUREG/CR–1039) previously considered a scenario for misuse of welding

rods involving carrying of extra electrodes in a shirt pocket (see Section 3.6.3.4.1). An

individual was assumed to be exposed to three electrodes for 2000 h/yr. Such an exposure

could occur, for example, if the electrodes were carried inadvertently in the pocket of a lab coat.

In this assessment, the EDE to an individual is estimated for the same exposure conditions

assumed by McDowell-Boyer, as described above. For external exposure to thorium in welding

rods carried next to the body, the EDE is assumed to be given by the dose equivalent at a

depth of 10 cm in tissue (see Appendix 4). For exposure to three electrodes containing a total

of 0.9 g of thorium that has been aged for 20 years, the EDE, as calculated using CONDOS II

(Computer Codes, O’Donnell et al., 1981), is 0.08 mSv (8 mrem). This estimate agrees

reasonably well with the dose equivalent to the skin of 0.04 mSv (4 mrem) estimated previously

by McDowell-Boyer for welding rods containing 30% less thorium.
3–119

3.6.5 Summary

This assessment has considered doses to members of the public from distribution, use, and

disposal of welding rods containing thorium. The amount of thorium contained in welding rods

was assumed to be 2% by weight, in accordance with current practices, and the amount of

thorium in a typical electrode was assumed to be 0.3 g. The number of thoriated-tungsten

welding rods distributed, used, and disposed per year was assumed to be 5 million. Results of

the assessment are presented in Table 3.6.2 and may be summarized as follows:

The highest individual and collective doses should occur during routine use of welding

rods by welders and grinders. Doses during distribution and transport and from disposal

in landfills and incinerators should be very small by comparison.

The estimated collective dose to professional welders is based on an assumption that

5 million welding rods containing thorium are used per year. This estimate would be

conservative if fewer than 5 million electrodes were used per year or if a substantial

fraction of the electrodes were used on automatic welding machines.

The estimates of individual and collective doses to welders and grinders during routine

use are subject to considerable uncertainty, due primarily to the variability and

uncertainty in the concentrations of thorium and its decay products in air during welding

and grinding. This uncertainty could be as large as an order of magnitude. In addition,

the doses during welding could be substantially different depending upon the arc-current

and the type of current used. On the other hand, the doses during welding and grinding

could be substantially lower when local exhaust ventilation is used in addition to normal

room ventilation.

Doses for particular scenarios involving accidents and misuse also were estimated in this

assessment, including doses resulting from a fire in a warehouse and from inadvertent carrying

of extra welding rods in a shirt pocket. However, the estimated doses for the scenarios

considered are quite low, and the results suggest there are no credible scenarios for accidents

or misuse that could result in doses comparable to the doses that would be experienced by

welders and grinders during routine use of welding rods. This is a reasonable conclusion when

one considers that the scenarios for routine exposure to welders and grinders involve release

into the air of most of the thorium contained in welding rods.
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Table 3.6.1 Individual and Collective Doses From Distribution and Transport 

of Thoriated-Tungsten Welding Rods Obtained in Previous Assessments a

Annual Individual Dose Equivalent

(mrem)b Annual Collective

Dose Equivalentc

(person-rem) Population Group Maximum Average b

Truck drivers 0.2 0.002 0.0052

UPS terminal

 employees

0.15 <0.001 0.01

Retail employees 0.06 0.06 0.3

Warehouse

 employees

0.15 0.15 0.062

Customers in retail

 stores

<0.001 <0.001 0.35

Total 0.72

a

 Dose estimates for welding rods containing 2% by weight of thorium that has been aged for

20 years after chemical separation and 0.23 g of thorium per welding rod, as obtained by

McDowell-Boyer (NUREG/CR–1039), based essentially on methodology presented in

Appendix A.3 of this report. Estimates are doses to whole body from external exposure, which

are essentially the same as effective dose equivalents. Doses from inhalation or ingestion are

negligible by comparison (see Section 3.6.3.2.1).

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 Doses assuming annual distribution of 1 million welding rods.
3–121

Table 3.6.2 Summary of Potential Radiation Doses From Use

of Welding Rods Containing Thorium a

Exposure Scenario

Individual Annual Effective

Dose Equivalent

(mrem)c

Collective Effective

Dose Equivalentb

(person-rem)c

Distribution and transport 0.3b,d 5

Routine use (without local

exhaust)e

 Professional welders

 AC Current

 DC Current

 Grinding

 Home welders

 AC Current

 DC Current

 Grinding

 Dedicated grinder

500

 20

 20

 100

 3

 3

800f

30,000

 500

Disposal

 Landfills

 Incinerators

 0.4g

 2i

400h

0.3

Accidents or misusej

 Warehouse fire

 Carrying electrodes

 6k

 8l

NAm

a

 Dose estimates are based on the assumption that each welding rod contains 0.3 g of thorium

prior to use. Dose estimates for distribution and transport, routine use, and accidents or misuse

assume that thorium has been aged for 20 years after chemical separation. Dose estimates for

disposal assume that all decay products of 232Th are in activity equilibrium. b

 Dose estimates are based on assumption that 5 million welding rods are distributed, used, and

disposed per year.

c

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

d

 Dose estimate applies to truck drivers; dose estimates are somewhat less for UPS terminal

employees, retail employees, or warehouse employees, and considerably less for customers in

retail stores (see Table 3.6.1).

e

 Dose estimate for inhalation dose during welding and during grinding of welding rods. Use of

local exhaust is estimated to reduce doses by a factor of 10.

f

 Dose estimate applies to individuals who are assumed to spend most of their working time

grinding electrodes, but this activity apparently is not a common occurrence.

g

 Dose estimate applies to waste collectors and future on-site residents at landfills; doses are

considerably less for workers at landfills and future on-site residents (see Section 3.6.4.3.1).

h

 Dose estimate applies to future on-site residents at landfills for 1000 years after disposal due

to 1 year’s disposals; if dose to future on-site residents is not taken into account, estimated

annual collective dose to waste collectors and workers at landfills is 0.02 person-Sv

(2 person-rem).

i

 Dose estimate applies to waste collectors at incinerators; dose estimates are considerably less

for workers at incinerators and off-site members of the public (see Section 3.6.4.3.2).

j

 Dose estimates apply to single occurrence of accident or misuse scenario.

k

 Dose estimate applies during cleanup after fire in warehouse containing 100,000 welding rods;

dose estimate for firefighter using respiratory protection is considerably less.

l

 Dose estimate assumes that individual carries three electrodes in shirt pocket for 2,000 hours.

m Not applicable.

3–123

3.7 Thorium in Electric Lamps for Illuminating Purposes

3.7.1 Introduction

In 10 CFR 40.13(c)(1)(iv), persons who receive, possess, use, or transfer electric lamps

containing thorium for illuminating purposes are exempted from licensing requirements for

source material, provided each lamp does not contain more than 50 mg of thorium. The

maximum amount of thorium (50 mg) allowed under this exemption corresponds to an activity of

0.2 kilobecquerel (kBq) (5.5 nanocurie (nCi)). This exemption was proposed on

August 17, 1965 (30 FR 10203), and issued as a final rule on December 22, 1965

(30 FR 15802).

Quantitative and qualitative information on radiation doses to the public resulting from the use

and disposal of electric lamps containing thorium in amounts allowed under this exemption,

including assessments for accident scenarios, are available from two sources. First, the

proposed rule cited above contains information on doses from routine use and disposal of

electric lamps, and potential doses from accidental breakage of lamps are also discussed. 

Second, Buckley et al. (NUREG/CR–1775) performed a dose assessment for fluorescent lamp

starters containing thorium that considered routine exposure to workers during distribution,

routine exposures during use of lamps, exposures from disposal in landfills or by incineration,

and accidental exposure to firefighters during a warehouse fire. Results of these assessments

are discussed in Section 3.7.3.

3.7.2 Description of Items

A variety of electric lamps used for illuminating purposes may contain thorium in amounts

allowed under this exemption. In the past, thoriated-tungsten filaments were used extensively

in incandescent lamps intended for general lighting purposes. However, most such lamps now

use rhenium-tungsten filaments. Thorium has also been widely used in fluorescent lamp

starters to produce ionization within the starter (NUREG/CR–1775) and in certain specialty

illuminating lamps, such as those requiring high electrode emissivities or hot strength, and

lamps that emit intense light or light with specific spectra. For example, photoflash lamps,

which are attached to cameras by the user, and lamps used in vehicles (e.g., directional-signal

lamps in automobiles and train lights) have been identified as important types of specialty

lamps containing thorium. However, detached photoflash lamps for cameras are no longer are

used.

Thorium used in electric lamps is in an insoluble oxide form. The thorium normally is enclosed

by an airtight glass envelope that is penetrated by the necessary electrical connections, and

additional containment materials (e.g., an aluminum housing) may be included. These

envelopes ensure that, during normal use, radioactive material (including 220Rn) would not be

released from the lamps. Therefore, unless the lamps are broken, only external exposure to

thorium and its decay products would be of concern.

3.7.3 Summary of Previous Analyses and Assessments

This section summarizes the results of previous analyses and assessments of doses to

members of the public from use of electric lamps—for illuminating purposes—that contain 
3–124

amounts of thorium allowed under this exemption. As indicated in Section 3.7.1, quantitative

and qualitative estimates of dose are provided in the proposed rule for this exemption and in the

report by Buckley et al. (NUREG/CR–1775).

3.7.3.1 Assessments by Atomic Energy Commission

The proposed rule for this exemption cited in Section 3.7.1 discusses dose assessments

performed by the Atomic Energy Commission (AEC) (30 FR 10203). Exposure scenarios

considered include external dose during routine use of lamps, exposures following routine

disposal of broken or burned-out lamps, reclamation and recycling of the thorium for use in

other products, and ingestion or inhalation of thorium following accidental breakage of a lamp. 

Exposures during distribution and transport of lamps were not considered. The assessments

performed by the AEC are summarized as follows:

Routine Use. The dose equivalent from external exposure for an individual using a lamp

should not exceed a few tenths of a microsievert (Sv)/yr (few hundredths of a

mrem/yr). This result was based on an estimated external exposure rate of

1.5 nanocoulomb (nC)/kg-h (6 microroentgen (R)/h) at a distance of 1 meter from a

lamp containing 50 mg of thorium that had been aged for 20 years after chemical

separation. Inhalation and ingestion exposures would not be of concern.

Disposal and Recycling. Routine disposal of broken or burned-out lamps in normal

refuse disposal facilities is highly unlikely to result in any significant radiological impact,

and reclamation and recycling of thorium in other products was considered very unlikely.

Accidental Breakage. Ingestion or inhalation exposures could result from accidental

breakage of lamps containing thorium. However, ingestion of significant quantities of

thorium should be highly unlikely, due to the care that normally would be taken in

cleaning up and discarding broken glass. In addition, experiments in which lamps were

broken in small, unventilated rooms indicated it is very unlikely that the resulting

concentration of thorium in air would exceed the maximum permissible concentration

(MPC) for members of the public that had been established in Appendix B, Table II, of

10 CFR 20 (prior to the 1991 revision of 10 CFR20). The MPC in air was based on an

annual committed dose equivalent to the lungs of 15 mSv (1.5 rem) from continuous

exposure.

3.7.3.2 Assessments by Buckley et al.

Buckley et al. (NUREG/CR–1775) performed an assessment of doses to the public from use of

fluorescent lamp starters containing thorium. Based on available product information, each

starter was assumed to contain 45 g of thorium, but the results described below, have been

adjusted to represent the maximum thorium content of 50 mg per lamp allowed under this

exemption. Buckley et al. also assumed that 5 million lamp starters were distributed and

disposed of annually and that 50 million starters were in use at any time.

The scenarios for routine exposure to fluorescent lamp starters considered by Buckley et al.

include exposure to workers during distribution, exposures during use of the lamps, and

exposures from disposal of lamps in landfills or by incineration. In addition, an accident
3–125

scenario involving exposure to firefighters during a warehouse fire was considered. Results of

the assessments are summarized as follows:

Distribution. External doses to individual workers during distribution of lamp starters

were assumed to be very low compared with doses to individual users. Internal doses

to workers during distribution were presumed to be essentially zero (0), because no

radioactive material should be released from unbroken lamps.

Routine Use. Based on assumptions that a user of fluorescent lamps would spend an

average of 2 h/day at a distance of 1 meter from two lamps, and that the thorium had

been chemically separated from its decay products 20 years previously, the dose

equivalent to whole body from external exposure would be 7×105

 mSv/yr

(0.007 mrem/yr). For an assumed population of users of 70 million, the collective dose

equivalent then would be 5 person-Sv/yr (500 person-rem/yr).

Incineration. Based on an assumption that 5 million lamp starters would be incinerated

per year, the committed dose equivalents to maximally exposed individuals would be

7×105

 mSv/yr (0.007 mrem/yr) to bone and 4×105 mSv/yr (0.004 mrem/yr) to the lungs. 

The collective dose equivalents would be 0.7 person-Sv/yr (70 person-rem/yr) to bone

and 0.4 person-Sv/yr (40 person-rem/yr) to the lungs.

Disposal in Landfills. Based on assumptions that 5 million fluorescent lamp starters per

year would be sent to landfills, and that half of the thorium in the starters would be

transported in groundwater to assumed receptor locations, the collective dose

equivalents to bone and the lungs from disposal in landfills would each be

30 person-Sv/yr (3,000 person-rem/yr). Doses to off-site individuals, workers at the

landfill, or future residents at the disposal site were not estimated.

Accidental Fires. Based on assumptions that 500,000 lamp starters would be stored in

a warehouse, all of the thorium would be released during the fire, and the exposure time

for firefighters would be 8 hours without respiratory protection, the committed dose

equivalent to bone would be 5.2 Sv (520 rem).

Several of the assumptions used by Buckley et al. should be very conservative, including the

assumption for landfill disposal that half of the thorium in the starters would be transported in

groundwater to off-site locations, and the assumptions for the accident scenario that all of the

thorium would be released during a fire and that a firefighter would be exposed for 8 hours

without respiratory protection. In addition, as noted previously, the average amount of thorium

in lamp starters is only about 0.1% of the maximum amount allowed under this exemption, and

the dose estimates based on the amount of thorium actually used would be reduced

accordingly.

3.7.4 Present Exemption Analysis

This section presents estimates of doses to the public from routine distribution and transport,

use, and disposal of electric lamps containing thorium in amounts allowed by this exemption. 

Estimates of dose for accident scenarios involving breakage of lamps and a fire in a storage

warehouse also are presented.
3–126

The types of electric lamps containing thorium considered in this analysis include incandescent

lamps, fluorescent lamp starters, and automobile signal lamps. Some of these lamps

apparently are not widely used at the present time, including, for example, incandescent lamps

with thoriated-tungsten filaments. However, the estimates of individual and collective doses do

not depend significantly on the type of electric lamp being considered, particularly in regard to

the details of lamp construction. Rather, the estimates of individual and collective doses

depend primarily on assumptions about the conditions of exposure, and the dose estimates

presented in this section should be reasonably representative of any type of thorium-containing

electric lamp that might be widely used at the present time.

The amount of thorium per lamp used in most dose assessment is the same as that assumed

by Buckley et al., 45 g, while the maximum allowed by the exemption of 50 mg is reviewed for

routine use, accident and misuse scenarios. If information were available regarding the thorium

content in a product, it is stated. As noted in discussing the previous assessment for

fluorescent lamp starters (see Section 3.7.3.2), the amount of thorium actually used in a lamp

may, in some cases, be only a small fraction of the maximum allowable amount.

During routine use, as well as distribution and transport, of any type of electric lamp containing

thorium, external exposure to photons emitted by the thorium and its decay products is the only

exposure pathway of concern. This is because the radioactive material normally is sealed in an

airtight glass envelope, and ingestion or inhalation exposures, including exposures to 220Rn,

would not normally occur. Furthermore, the external dose rate from any type of lamp can be

estimated by assuming a point source, because the thorium is confined to a small region of the

lamp. Finally, for many types of lamps, a reasonable estimate of the dose rate can be obtained

by assuming that the source is unshielded, except for the presence of air between the source

and receptor locations, because the high-energy photons emitted by some of the decay

products of thorium require considerable shielding to reduce the external dose significantly.

3.7.4.1 Distribution and Transport

During the distribution of electric lamps that are assumed to be intact, external doses could be

received by distribution workers and by members of the public who purchase the products. 

During the transport of intact electric lamps, external doses could be received by truck drivers

and terminal workers. Members of the public along the truck routes also could receive external

exposure, but their doses would be much less than the doses to workers.

Individual and collective doses during distribution and transport of electric lamps containing

45g of thorium were estimated using the methodology described in Appendix A.3. The actual

distribution of electric lamps containing thorium is not known and probably varies considerably

depending on the type of lamp. For example, the annual distribution of fluorescent lamp

starters presumably is considerably greater than the annual distribution of automobile signal

lamps. In this assessment, collective doses are estimated for 1 million lamps distributed per

year, but dose estimates could be adjusted to represent actual distributions if they are known.

The distribution and transportation network for electric lamps containing thorium also is not

known, and it probably varies depending on the type of lamp. In this assessment, the

distribution and transportation network assumed for welding rods containing thorium, which was

developed by McDowell-Boyer (NUREG/CR–1039) and is described in Section 3.6.4.1, is

assumed to apply to electric lamps containing thorium. The essential features of the assumed
3–127

distribution and transportation network are the following. First, 80% of the lamps are assumed

to be shipped by truck to United Parcel Service (UPS) terminals over distances greater than

400 km, 10% to UPS terminals over distances between 32 and 400 km, and 10% directly to

warehouses and retail stores over distances less than 32 km. Second, 80% of the lamps

shipped to UPS terminals are assumed to be shipped to warehouses and 20% to retail stores. 

Finally, each warehouse is assumed to receive 1% of the total annual distribution of lamps, and

each retail store is assumed to receive 0.1%. 

The assumed distribution and transportation network is appropriate for products which, first, are

mostly transported by truck over appreciable distances and, second, are not sold in retail stores

to a large extent. These assumptions should be reasonable for such specialty products as

fluorescent lamp starters and lamps used in vehicles. The assumptions may be less

reasonable for incandescent lamps and photoflash lamps containing thorium, which normally

would be sold mainly in retail stores, but these types of lamps are not widely distributed at the

present time. Therefore, the assumed distribution and transportation network should be

reasonable for the types of lamps subject to this exemption that are most commonly distributed

at the present time.

Based on the assumptions described above, the individual and collective doses from

distribution and transport of 1 million lamps per year can be obtained based on an assumed

distribution of 1 million lamps each containing 45 g. The following results are as follows:

The effective dose equivalent (EDE) to individual truck drivers, UPS terminal employees,

retail employees, and warehouse employees would be less than 1×105

 mSv

(<0.001 mrem). The dose to individual customers in retail stores would be substantially

less.

The collective EDE for 1 million lamps distributed per year, including the dose to truck

drivers, UPS terminal employees, retail store employees, warehouse employees, and

customers in retail stores, would be less than 1×105

 person-Sv (<0.001 person-rem).

3.7.4.2 Routine Use

For routine use of electric lamps containing thorium, estimates of individual and collective doses

are obtained based on one of two approaches, depending on the type of lamp. In some cases,

estimates were based on the external dose rate at a distance of 1 meter from an unshielded

point source representing a lamp, using the external dose coefficients given in Table 3.1.7. In

other cases, results were obtained using CONDOS II (Computer Codes, O’Donnell et al., 1981),

which takes into account the appropriate amount of shielding between the source and receptor

locations when the shielding is significant. In all assessments for routine use, the thorium is

assumed to have been aged for 20 years after chemical separation. The types of lamps

considered in this assessment again include incandescent and fluorescent lamps used for

lighting, photoflash lamps used by amateur photographers, and automobile signal lamps. As

noted previously, the doses from routine use of these types of lamps should be reasonably

representative of the doses from use of any other types of lamps.
3–128

3.7.4.2.1 Incandescent and Fluorescent Lamps

For incandescent and fluorescent lamps used in the home or workplace, an individual is

assumed to spend 8 h/day at a distance of 1 meter from a single lamp, and 20 h/day at a

distance of 2 meters from two lamps. Based on the external dose rate at a distance of 1 meter

from a point source containing 45 g of thorium of 7.4×1011 mSv/h (7.4×109

 mrem/h), as

obtained from Table 3.1.7, and taking into account that the dose rate varies inversely as the

square of the distance from the source, the EDE to an individual would be less than 1×105

 mSv

(<0.001 mrem).

In estimating collective dose during routine use, it is unreasonable to assume that the exposure

times and distances for an individual given above would apply to an average member of the

exposed population. Rather, an average individual is assumed to spend 2 h/day at a distance

of 1 meter from a single lamp, and 20 h/day at an average distance of 2 meters from a single

lamp. Based on the external dose rate given above at a distance of 1 meter from a point

source, the collective EDE for 1 million lamps in use would be 2×104

 person-Sv

(0.02 person-rem).

As noted in Section 3.7.2, most incandescent lamps no longer contain thorium. In addition, as

noted in Section 3.7.3.2, the average amount of thorium in fluorescent lamps is only about 0.1%

of the maximum amount allowed under this exemption. Dose estimates given above should be

increased proportionally to represent doses from maximum amounts. At the maximum amount

allowed, the EDE to an individual would be about 2×104

 mSv (0.02 mrem). 

3.7.4.2.2 Automobile Signal Lamps

Estimates of external dose were obtained for individuals who drive automobiles with signal

lamps containing 170 g of thorium as estimated by one manufacturer. (Phone call,

A. Zielinski, GE Lighting, Cleveland, OH, October 1999). An automobile was assumed to have

four such lamps, two mounted near the front and two near the rear. The front lamps were

assumed to be located 0.9 meter from the driver and shielded by an equivalent of 1 cm of iron. 

The rear lamps were assumed to be 2.4 meters from the driver and shielded by an equivalent

of 0.75 cm of iron.

For the conditions described above, the EDE rate to the driver of an automobile calculated

using CONDOS II (Computer Codes, O’Donnell et al., 1981) is 0.009 nSv/h (0.9 nrem/h). If a

driver is assumed to spend an average of 4 h/day in an automobile, a reasonable maximum

value for most individuals, the EDE would be 1×105

 mSv (0.001 mrem) in a year. Increasing

the amount of thorium to the maximum allowed in all signal lamps and two headlamps, would

represent a scenario for the maximum individual EDE as 3×105

 mSv (0.03 mrem).

The number of automobile signal lamps in use containing thorium is not known. The collective

dose for 1 million lamps was estimated by assuming that 69% of all automobiles are occupied

by four persons for 1 h/day, 30% are occupied by two persons for 2 h/day, and 1% are

occupied by one person for 4 h/day. The resulting collective EDE for 1 million lamps in use

would be 9×104

 person-Sv (0.09 person-rem).
3–129

3.7.4.3 Disposal

Following their useful lifetime, electric lamps may be sent to landfills or incinerators for disposal. 

As noted in Section 3.7.3.1, the AEC previously concluded that recycling and reuse of thorium

in electric lamps is relatively unimportant, and there is no evidence that this conclusion is

inappropriate. Thus, recycling and reuse is not considered further in this assessment.

In this section, the generic methodology described in Appendix A.2 is used to obtain estimates

of individual and collective doses from disposal of electric lamps containing thorium. The

previous dose estimates of Buckley et al. (NUREG/CR–1775) described in Section 3.7.3.2 are

not used in this assessment. Particularly for disposal in landfills, Buckley et al. used

unreasonably conservative assumptions about release and transport of thorium to assumed

receptor locations. Doses to waste collectors, workers at a landfill, and future on-site residents

were not considered.

The number of lamps that might be disposed during a year is not known. As in the analysis of

collective doses from routine use in Section 3.7.4.2, doses are estimated assuming that

1 million lamps are disposed per year. It is further assumed that each lamp contains 45 g of

thorium so then, 45 g of thorium is assumed to be disposed per year. Finally, the dose

estimates assume that 80% of the lamps are sent to landfills and 20% to incinerators.

3.7.4.3.1 Disposal in Landfills

In the generic methodology described in Appendix A.2, doses are estimated for waste

collectors, workers at landfills, off-site individuals who reside near landfills during operations or

after closure, and individuals who might reside on the landfill sites after closure. In estimating

inhalation and ingestion doses to waste collectors and workers at landfills and in estimating

doses to off-site individuals during landfill operations, the thorium in the lamps is assumed to be

dispersed in amounts 10 times less than loose materials in trash, due to the physical form of

thorium in the lamps (see Section A.2.3.1.5). Similarly, for exposure to future on-site residents

at landfills, the amount of material inhaled or ingested is assumed to be a factor of 10 less than

for loose materials in the waste. However, for exposure to off-site individuals after closure of

landfills, the leaching of thorium from the lamps into groundwater is assumed to be the same as

the leaching of finely dispersed thorium in waste.

Based on the assumptions described above and the generic disposal methodology in

Appendix A.2, the following estimates of individual and collective doses from disposal in landfills

are obtained:

The EDE to individual waste collectors, a future on-site resident, individual workers at

landfills, and to off-site individuals during landfill operations and after closure of the

landfills would each be less than 1×105

 mSv (<0.001 mrem).

The collective EDE from 1 year’s disposals of lamps would be about 1×104

 person-Sv

(0.01 person-rem), due almost entirely to exposure to future on-site residents for

1000 years after facility closure. If exposure to future on-site residents are not taken

into account, the collective EDE from exposure to waste collectors and workers at

landfills would be less than 1×105

 person-Sv (<0.001 person-rem).
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3.7.4.3.2 Disposal in Incinerators

In the generic methodology for disposal in incinerators described in Appendix A.2, doses are

estimated for waste collectors, workers at incinerators, and off-site individuals who reside near

the incinerators during operations. In estimating doses to waste collectors and workers at

incinerators, the thorium in the lamps is assumed to be 10 times less dispersible than thorium in

loose materials, as in the dose assessment for disposal in landfills, due to the physical form of

the lamps.

Based on the assumptions described above and the generic disposal methodology in

Appendix A.2, the following estimates of individual and collective doses from disposal in

incinerators are obtained:

The EDE to individual waste collectors, individual workers at incinerators, and individual

off-site residents near incinerators would each be less than 1×105

 mSv (<0.001 mrem).

The collective EDE from 1 year’s disposals of lamps would be less than

1×105

 person-Sv (<0.001person-rem).

3.7.4.4 Accidents

As described in Section 3.7.3, two accident scenarios involving electric lamps containing

thorium were considered in previous assessments by the AEC and Buckley et al.

(NUREG/CR–1775): accidental breakage of lamps and a fire in a large warehouse. Both

situations could result in exposures to thorium and its decay products, and both are considered

in this assessment. Doses resulting from a fire in a home and in a small storage room are also

considered. In each case the lamps are assumed to contain an average quantity of 45 g or

50 mg of thorium, which is the maximum amount allowed under this exemption, and both doses

are estimated using the generic methodology described in Appendix A.1. All dose estimates

include the contributions from 220Rn, although in most cases this contribution is negligible

compared with the dose from other radionuclides in the 232Th decay chain.

3.7.4.4.1 Breakage of Lamps

In this assessment, a single lamp is assumed to be broken in a room in a home. An

upper-bound estimate of the dose to an individual who is in the room when the lamp breaks,

and who cleans up the broken lamp, is obtained by assuming that the model developed in

Appendix A.1 for the dose from a spill of material in powder form applies to this scenario. This

assumption should be conservative, because the amount of thorium transferred from a broken

lamp into the air or to the hands of an individual should be considerably less than the amounts

for a spill of material in powder form. In estimating the dose following breakage, the room

volume is assumed to be 30 m3

, the air ventilation rate is 1 volume change per hour, the

breathing rate is 1.2 m3

/h, and the exposure time is 0.5 hour. Based on these assumptions, the

EDE from inhalation and ingestion would be 0.002 mSv (0.2 mrem) for the maximum quantity

and less than 1×105

 mSv (<0.001 mrem) for the minimum quantity.
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3.7.4.4.2 Fire in Home, Storage Room, or Warehouse

In the generic methodology described in Appendix A.1, doses are estimated for a fire in a

home, a small storage room, or a warehouse. For each scenario, doses are estimated for

individuals during the fire, either the occupant of a home or storage area or a firefighter during a

warehouse fire, as well as individuals involved in cleanup activities after the fire. The results of

the dose assessment based on the generic methodology are as follows.

For a fire in a home, a room in which the fire occurs is assumed to contain 10 lamps with

maximum allowed quantities the dose to an individual in the room during the fire would be

4×104

 mSv (0.04 mrem), and the dose during cleanup after the fire would be less than

1×105

 mSv (<0.001 mrem). Doses for average use quantities are obtained by multiplying by

0.0009.

For a fire in a small storage room, the room is assumed to contain 1000 lamps with maximum

allowed quantities the dose to an individual during the fire would be 0.09 mSv (9 mrem), and

the dose during cleanup after the fire would be 0.001 mSv (0.1 mrem). The dose estimate

during the fire would be quite conservative if the individual uses effective respiratory protection. 

Doses for average use quantities are obtained by multiplying by 0.0009.

For a fire in a warehouse, the storage area is assumed to contain 10,000 lamps. The dose to a

firefighter using respiratory protection during the fire would be 2×104

 mSv (0.02 mrem), and the

dose to an individual who is not using any respiratory protection during cleanup after the fire

would be 0.001 mSv (0.1 mrem). Doses for average use quantities are obtained by multiplying

by 0.0009.

Buckley et al. used one million lamps for their assessment and our doses must be adjusted

higher by a factor of 100 for comparison. With this adjustment, the estimated dose to a

firefighter during a warehouse fire is much less than the value obtained previously by Buckley et

al. (NUREG/CR–1775), as given in Section 3.7.3.2. The large reduction in estimated dose,

compared with the previous result, is due primarily to the assumptions in this assessment that

0.1% of the thorium in the lamps would be released in a fire, instead of 100%; the firefighter

would use a respirator with a protection factor of 1000, instead of no respiratory protection; and

the firefighter would be exposed for 0.5 hour, instead of 8 hours.

3.7.5 Summary 

This assessment has considered doses to the public from use —for illuminating purposes—

and disposal of electric lamps containing thorium. Results of the assessment are presented in

Table 3.7.1 and may be summarized as follows:

Doses to individuals during routine use from lamps with typical amounts of thorium

(45 g) appear to be quite low, i.e., less than 0.001 mSv/yr (<0.1 mrem/yr), due

primarily to the small amount of thorium that each lamp is allowed to contain. Doses to

individuals during distribution and transport and disposal also appear to be low, for a

maximum thorium context of 50 mg. The dose to an individual from routine use could

be 3×104

 mSv/yr (0.03 mrem).
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Estimates of collective dose for routine exposure scenarios are based on an assumption

that 1 million lamps per year would be distributed, transported, used, and disposed. For

some products covered by this exemption, the actual number could be considerably

greater than 1 million per year. If the actual number of lamps containing thorium were

known, dose estimates for these scenarios could be adjusted by the appropriate

amount.

This assessment also considered doses to individuals resulting from accidental breakage of a

lamp, from a fire in a home, a small storage room, and a large warehouse. Potential doses

from breakage of a single lamp appear to be very low, due to the likelihood that only a small

fraction of the thorium in a lamp would be inhaled or ingested. Potential doses from a fire in a

home are low, unless a large number of lamps would be involved. Potential doses from a fire in

a large warehouse are also low, if the building is assumed to contain 1 million lamps and

firefighters are assumed to use respiratory protection. Potential doses from a fire in a small

storage room are substantially higher if the room is assumed to contain 1,000 lamps and an

individual fighting the fire does not use respiratory protection.
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Table 3.7.1 Summary of Potential Radiation Doses From Use 

of Electric Lamps Containing Thorium a

Exposure Scenario

Individual Annual Effective

 Dose Equivalent

(mrem)c

 Collective Effective

Dose Equivalentb

(person-rem)c

Distribution and transportb <0.001d <0.001

Routine use

 Lightinge

 Auto signal lamps

<0.001f

<0.001g

0.02

0.09

Disposalb

 Landfills

 Incinerators

<0.001h

<0.001j

0.01i

<0.001

Accidents or misusek

 Lamp breakage

 Home fire

 Storage room fire

 Warehouse fire

<0.001

<0.001

0.008l

<0.001

NAm

See following page for footnotes.
3–134

Footnotes to Table 3.7.1

a

 Dose estimates are based on assumption that all lamps contain average amount of thorium of

45 g. Dose estimates for distribution and transport, routine use, and accidents or misuse

assume that thorium has been aged for 20 years after chemical separation. Dose estimates for

disposal assume that all decay products of 232Th are in activity equilibrium.

b

 Dose estimates are based on assumption that 1 million electric lamps containing thorium are

distributed, used, or disposed per year. Actual use of some types of lamps could be

substantially higher, but other types of lamps apparently are not used to a significant extent at

the present time.

c

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

d

 Dose estimate applies to truck drivers; UPS terminal employees, retail employees, or

warehouse employees, and customers (see Section 3.7.4.1 and Table 3.6.1).

e

 Dose estimates apply to use of incandescent or fluorescent lamps in home and workplace.

f

 Dose estimates for individual of the maximum thorium content of 50 mg would be 4×104 mSv

(0.04 mrem).

g

 Dose estimate based a 4 signal lamps each containing 45 g. At the maximum content of

50 mg the dose would be 3×104

 mSv (0.03 mrem).

h

 Dose estimate applies to waste collectors at landfills; workers at landfills, off-site members of

the public, and future on-site residents (see Section 3.7.4.3.1).

i

 Dose estimate applies to future on-site residents at landfills for 1,000 years after disposal due

to one year's disposals.

j

 Dose estimate applies to waste collectors at incinerators; workers at incinerators and off-site

members of the public (see Section 3.7.4.3.2).

k

 Dose estimates apply to single occurrence of accident or misuse scenario; number of lamps

assumed to be involved in each scenario is described in Section 3.7.4.4.

l

 Dose estimate applies during firefighting without respiratory protection; dose estimate during

cleanup after fire is considerably less (see Section 3.7.4.4.2).

m Not applicable.
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3.8 Thorium in Germicidal Lamps, Sunlamps, and Lamps for 

Outdoor or Industrial Lighting

3.8.1 Introduction

In 10 CFR 40.13(c)(1)(v), persons who receive, possess, use, or transfer thorium-containing

germicidal lamps, sunlamps, and lamps for outdoor or industrial lighting are exempted from

licensing requirements for source material, provided each lamp does not contain more than 2 g

of thorium. The maximum amount of thorium (2 g) allowed by the exemption corresponds to an

activity of 8 kilobecquerel (kBq) (0.22 microcurie (Ci)). This exemption was proposed on

August 17, 1965 (30 FR 10203), and issued as a final rule on December 22, 1965

(30 FR 15802).

Quantitative and qualitative information on radiation doses to the public resulting from use and

disposal of germicidal lamps, sunlamps, and lamps for outdoor or industrial lighting that contain

thorium in amounts allowed under the exemption, including assessments for accident

situations, has been published only in the proposed rule cited above. This information is

discussed in Section 3.8.3.

3.8.2 Description of Items

Thorium used in germicidal lamps, sunlamps, and lamps for outdoor or industrial lighting is in

an insoluble oxide form. The thorium normally is incorporated in the cathodes of the lamps or,

alternatively, as a coating on the cathodes. In addition, the thorium normally is enclosed by an

airtight glass envelope, which ensures that radioactive material (including 220Rn) would not be

released from the lamp during normal use. Therefore, unless the lamps are broken, only

external exposure to thorium and its decay products would be of concern.

The most common type of outdoor and industrial lamps containing thorium appears to be

high-intensity discharge (HID) lamps, including mercury-vapor, metal-halide, and

mercury-xenon arc lamps. These lamps are constructed with thick-walled glass envelopes

designed to withstand considerable temperature variations and rough use. In addition to

general outdoor or industrial lighting, HID lamps are used for roadway lighting and for lighting in

large indoor structures. The lamps normally are mounted at least 3 to 4 meters above the

ground or floor level, but they sometimes are mounted much higher, e.g., in sports arenas.

HID light is produced by the passage of an electric current through a gas or vapor under

pressure. According to a manufacturer, two electrodes are used per light and typically for

industrial lighting each electrode will contain 200 mg of tungsten of which 2% is thorium (Phone

call, A. Zielinski, GE Lighting, Cleveland, OH, October 1999).

The annual distribution of thorium-containing germicidal lamps, sunlamps, and lamps for

outdoor or industrial lighting is not known. According to one manufacturer, the annual

distribution of HID lamps may constitute less than 3% of the total distribution of all lamps used

for lighting purposes, but the total distribution of all lamps is unknown.
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3.8.3 Summary of Previous Analyses and Assessments

As indicated in Section 3.8.1, the proposed rule for this exemption contains the only previous

assessments of radiation doses from use of germicidal lamps, sunlamps, and lamps for outdoor

or industrial lighting that contain allowable amounts of thorium. These assessments were

performed by the Atomic Energy Commission (AEC) (30 FR 10203).

Exposure scenarios considered by the AEC included external dose during routine use of lamps,

exposures following routine disposal of broken or burned-out lamps, reclamation and recycling

of the thorium for use in other products, and ingestion or inhalation of thorium following

accidental breakage of a lamp. Results are summarized as follows:

Routine Use. External exposure would be the only pathway of concern, and the dose

equivalent to an individual using a lamp should not exceed a few tenths of a microsievert

(Sv)/yr (few hundredths of a mrem/yr). This result was based on an estimated external

exposure rate of 6.5×102

 nanocoulomb (nC)/kg-h (0.25 microroentgen (R)/h) at a

distance of 1 meter from a lamp containing 2 g of thorium with its decay products in

activity equilibrium.

Disposal and Recycling. Routine disposal of broken or burned-out lamps in normal

refuse disposal facilities is highly unlikely to result in any significant radiological impact,

and reclamation and recycling of thorium in other products was considered very unlikely.

Accidental Breakage. Ingestion or inhalation exposures could result from accidental

breakage of lamps containing thorium. However, ingestion of significant quantities of

thorium should be highly unlikely, due to the care that normally would be taken in

cleaning up and discarding broken glass. In addition, experiments in which lamps were

broken in small, unventilated rooms indicated that it is very unlikely that the resulting

concentration of thorium in air would exceed the maximum permissible concentration

(MPC) for members of the public that had been established in Appendix B, Table II, of

10 CFR 20 (prior to the 1991 revision of 10 CFR20). The MPC in air was based on an

annual committed dose equivalent to the lungs of 15 mSv (1.5 rem) from continuous

exposure.

The AEC did not consider individual doses from distribution of lamps containing thorium, and

estimates of collective dose from routine use were not provided.

3.8.4 Present Exemption Analysis

This section presents estimates of dose to members of the public from routine distribution and

transport, use, and disposal of germicidal lamps, sunlamps, and lamps for outdoor or industrial

lighting that contain thorium in amounts allowed by this exemption. Estimates of dose for

accident scenarios involving breakage of lamps and a fire in a storage warehouse also are

presented.

Dose assessments for all scenarios assume that each outdoor or industrial lamp contains 8 mg

of thorium, or 32 Bq (0.88 nCi) of 232Th. The actual amount of thorium contained in the different
3–137

types of lamps covered by this exemption is not known but, in most cases, presumably would

be much less than the maximum allowed. 

During routine use, as well as distribution and transport, of any type of lamp containing thorium,

external exposure to photons emitted by thorium and its decay products is the only exposure

pathway of concern, because the radioactive material is normally sealed in a glass envelope. 

Therefore, ingestion or inhalation exposures, including exposures to 220Rn, would not normally

occur. These exposure pathways would be of concern only for disposal and accidents.

3.8.4.1 Distribution and Transport

During distribution of lamps that are assumed to remain intact, external doses could be

received by distribution workers and by members of the public who purchase the products. 

During the transport of intact lamps, external doses could be received by truck drivers and

terminal workers. Members of the public along the truck routes also would receive external

exposure, but their doses would be much less than the doses to workers.

Individual and collective doses during distribution and transport of germicidal lamps, sunlamps,

and lamps for outdoor or industrial lighting that contain 8 mg of thorium were estimated using

the methodology described in Appendix A.3. The actual distribution of these lamps is not

known, and it probably varies considerably depending on the type of lamp. In this assessment,

collective doses are estimated for 1 million lamps distributed per year. Dose estimates could be

adjusted to represent actual distributions if they were known.

In this assessment, the distribution and transportation network assumed for welding rods

containing thorium, which was developed by McDowell-Boyer (NUREG/CR–1039) and is

described in Section 3.6.4.1, is assumed to apply to a mixture of germicidal lamps, sunlamps,

and lamps for outdoor or industrial lighting. 

The essential features of the assumed distribution and transport network are the following. 

First, 80% of the lamps are assumed to be shipped by truck to United Parcel Service (UPS)

terminals over distances greater than 400 km, 10% to UPS terminals over distances between

32 and 400 km, and 10% directly to warehouses and retail stores over distances less than

32 km. Second, 80% of the lamps shipped to UPS terminals are assumed to be shipped to

warehouses and 20% to retail stores. Finally, each warehouse is assumed to receive 1% of the

total annual distribution of lamps, and each retail store is assumed to receive 0.1%. Thus, the

assumed distribution and transportation network is appropriate for products which, first, are

mostly transported by truck over appreciable distances and, second, are distributed mainly to

the user but are sold in retail stores only to a lesser extent. These assumptions should be

reasonable for the mixture of lamps subject to this exemption (see also Section 3.8.4.2.4).

Based on the assumptions described above, the individual and collective doses from

distribution and transport of 1 million lamps per year can be obtained by scaling of the results

for welding rods containing thorium given in Table 3.6.1 by the total amount of thorium

distributed per year. Dose estimates for welding rods in Table 3.6.1 were based on an

assumed distribution of 2.3×105

 g of thorium per year, and the amount of thorium distributed

under this exemption is assumed to be 8×103

 g/yr. Therefore, the estimates of individual and

collective doses for this exemption are a factor of 0.035 lower than the corresponding dose

estimates for welding rods in Table 3.6.1. The following results are obtained:
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The effective dose equivalent (EDE) to individual truck drivers, UPS terminal employees,

retail employees, and warehouse employees would be less than 1×105 to 7×105 mSv

(<0.001 to 0.007 mrem), with the dose to truck drivers being the highest. The dose to

individual customers in retail stores would be considerably less.

The collective EDE for 1 million lamps distributed per year, including the dose to truck

drivers, UPS terminal employees, retail store employees, warehouse employees, and

customers in retail stores, would be 2×104

 person-Sv/yr (0.02 person-rem/yr).

The results given above may be somewhat conservative, because they do not consider any

self-shielding provided by the types of lamps subject to this exemption. However, the thicker

glass envelopes and generally more rugged construction, when compared, for example, with

electric lamps for illuminating purposes, should not reduce the dose estimates by more than a

few tenths of a percent for the high-energy photons of concern in the decay of some of the

thorium decay products. Such adjustments are negligible compared with uncertainties in the

exposure times and distances from sources used in defining the exposure scenarios.

3.8.4.2 Routine Use

As indicated in Section 3.8.3, the AEC concluded that annual dose equivalents to individuals

from exposure to a single lamp containing 2 g of thorium should not exceed a few tens of a Sv

(few hundreds of a mrem). However, more detailed assessments of doses for specific uses of

lamps and estimates of collective dose were not provided.

In this assessment, individual and collective doses to the public were estimated using Micro

Shield (Computer Codes, Grove Engineering, 1996) for three routine use situations: HID lamps

in a sports arena, HID lamps in a factory, and sunlamps in a home. In each case, the lamps

are assumed to contain 8 mg of thorium for the average lamp or 2 g of thorium, the maximum

amount allowed under this exemption, and the thorium is assumed to have been aged for

20 years after chemical separation. During routine use of lamps, only external exposure is

assumed to be important. Although other exposure scenarios can be postulated (e.g.,

exposure to motorists traveling along streets lit by HID lamps and exposure to individuals in

homes located near street lights), the scenarios evaluated should be reasonably representative

and, furthermore, may provide conservative estimates of dose for other routine exposure

scenarios. Dose assessments for the three exposure scenarios considered are described in

the following sections.

3.8.4.2.1 HID Lamps in Sports Arena

The following assumptions were made in estimating external doses from routine use of HID

lamps in a sports arena. First, the lamps were assumed to be mounted above the center of the

arena in a single square array 24 meters on a side containing 81 lamps, each lamp located

3 meters from its nearest neighbor; the array was assumed to be located 23 meters above the

occupants of the arena. Second, an individual was assumed to occupy the arena for 3 hours

per event and to attend one event per week for an entire year; the arena was assumed to be

occupied by 5000 persons per night for each night of the year. Third, all exposures were

assumed to occur at a location beneath the center of the array of lamps.
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For the assumptions described above, the EDE for an individual attending one event per week

for an entire year would be less than 1×105 mSv (<0.001 mrem). For an attendance of

5000 persons every night for a year at the arena, the collective EDE per arena would be less

than 1×105

 person-Sv/yr (<0.001 person-rem/yr).

For the maximum allowed quantity of 2 g, the EDE for an individual attending one event per

week for an entire year would be 4×105

 mSv (0.004 mrem).

The dose estimates given above are believed to be conservative, because it is unlikely that an

individual would attend an average of one event per week during a year and that an arena

would be used each day of the year. In addition, most individuals in an arena would not be

located beneath the center of an array of lamps, where the external dose would be the highest.

3.8.4.2.2 HID Lamps in Factory

In estimating external doses from routine use of HID lamps in a factory, the lamps were

assumed to be arranged as described above for HID lamps in a sports arena, except the array

of lamps was assumed to be located 6 meters above the factory workers. All exposures were

assumed to occur at a location beneath the center of the array of lamps.

For the assumptions described above, the EDE for an individual worker who spends 2000 h/yr

in the factory would be less than 1×105

 mSv (<0.001 mrem). For a factory that is assumed to

employ 50 workers, the collective EDE per factory would be less than 1×105

 person-Sv

(<0.001 person-rem).

For the maximum allowed quantity of 2 g per lamp, the EDE for an individual worker would be

0.003 mSv (0.3 mrem).

The dose estimates given above are believed to be conservative, because it is unlikely that

individual workers would spend as much as 2000 h/yr beneath the lamps. Furthermore, when

exposed to the lamps, the workers would not always be located beneath the center of an array.

3.8.4.2.3 Sunlamp in Home

In estimating external doses from routine use of a sunlamp in a home, an individual user was

assumed to be located at a distance of 1.8 meters from the lamp, while actually using it, for

18 h/yr, and at an average distance of 6 meters from the lamp, while not using it, for 7300 h/yr. 

Other occupants of the home were assumed to be located at an average distance of 6 meters

from the lamp for 4400 h/yr.

For the assumptions described above, the estimated individual and collective doses from use of

a sunlamp in a single home are summarized as follows. First, for sunlamp containing 8 mg Th,

the EDE for a sunlamp user and other occupants of the home, would be less than 1×105

 mSv

(<0.001 mrem). For the maximum allowed quantity of 2 g per lamp, the EDE to the user would

be 4×104

 mSv (0.04 mrem).

The dose estimates given above are believed to be conservative, primarily because it is unlikely

that average individuals, either users of a sunlamp or other occupants of the home, would

spend as much time in the home as assumed in this analysis.
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3.8.4.2.4 Collective Dose From Mixture of Lamps

Based on the dose estimates given above for use of HID lamps in a sports arena or factory and

use of a sunlamp in a home, the collective dose from a mixture of 1 million lamps can be

estimated by assuming a relative mixture of lamps among the different uses considered in this

assessment. The actual mixture of lamps in sports arenas, factories, and homes is unknown. 

In this assessment, it is arbitrarily assumed, first, that 90% of the lamps are HID lamps and

10% are sunlamps and, second, that 20% of the HID lamps are used in sports arenas and 80%

in factories. If lamp replacement is ignored, these levels of lamp use would require

approximately 2,200 arenas and 9,000 factories, assuming that each contains a single array of

81 lamps, and 100,000 homes, each containing a single sunlamp.

For the assumptions described above, and using values not rounded, the estimated collective

EDE for 1 million lamps in use would be 0.004 person-Sv/yr (0.4 person-rem/yr). About 65% of

the collective dose would result from exposures in sports arenas, somewhat less than 32% from

exposures in factories, and only about 3% from exposures to sunlamps.

The estimate of collective dose for 1 million lamps in use given above should be conservative,

primarily because, for the scenarios considered, the duration of exposure to an average

individual to lamps and the average distance from lamps are likely to be overestimates.

3.8.4.3 Disposal

Following their useful lifetime, germicidal lamps, sunlamps, and lamps for outdoor or industrial

lighting may be sent to landfills or incinerators for disposal. In addition, these types of lamps

could be recycled for salvage and reuse of valuable materials, although it is not known if this

actually occurs. As noted in Section 3.8.3, the AEC previously concluded that recycling and

reuse of thorium are relatively unimportant in these types of lamps.

In this section, the generic methodology described in Appendix A.2 is used to obtain estimates

of individual and collective doses from disposal of the types of lamps covered by this exemption

in landfills, by incineration, or by recycling.

The number of lamps subject to this exemption that might be disposed during a year is not

known. As in the analysis of collective doses from routine use in Section 3.8.4.2, doses are

estimated assuming that 1 million lamps are disposed per year. It is further assumed that each

lamp contains 8 mg of thorium, so that 8×103

 g of thorium is assumed to be disposed per year. 

Finally, the dose estimates assume that 60% of the lamps are sent to landfills, 20% to

incinerators, and 20% to a metal smelter.

3.8.4.3.1 Disposal in Landfills

In the generic methodology described in Appendix A.2, doses are estimated for waste

collectors, workers at landfills, off-site individuals who reside near landfills during operations or

after closure, and individuals who might reside on the landfill sites after closure. In estimating

inhalation and ingestion doses to waste collectors and workers at landfills and in estimating

doses to off-site individuals during landfill operations, the thorium in the lamps is assumed to be

dispersed in amounts 10 times less than loose materials in trash, due to the physical form of

thorium in the lamps (see Section A.2.3.1.5). Similarly, for exposure to future on-site residents
3–141

at landfills, the amount of material inhaled or ingested is assumed to be a factor of 10 less than

for loose materials in the waste. However, for exposure to off-site individuals after closure of

landfills, the leaching of thorium from the lamps into groundwater is assumed to be the same as

the leaching of finely dispersed thorium in waste.

Based on the assumptions described above and the generic disposal methodology in

Appendix A.2, the following estimates of individual and collective doses from disposal in landfills

are obtained:

The EDE to individual waste collectors would be 1×105 mSv (0.001 mrem). The annual

EDE to a future on-site resident, to individual workers at landfills, off-site individuals

during landfill operations and after closure of the landfills would be less than 1×105

 mSv

(<0.001 mrem).

The collective EDE from 1 year’s disposals of lamps would be about 0.02 person-Sv

(2 person-rem), due almost entirely to exposure to future on-site residents for

1000 years after facility closure. If exposure to future on-site residents are not taken

into account, the collective EDE from exposure to waste collectors and workers at

landfills would be about 8×105

 person-Sv (0.008 person-rem).

3.8.4.3.2 Disposal in Incinerators

In the generic methodology described in Appendix A.2 for disposal in incinerators, doses are

estimated for waste collectors, workers at incinerators, and off-site individuals who reside near

the incinerators during operations. In estimating doses to waste collectors and workers at

incinerators, the thorium in the lamps is assumed to be 10 times less dispersible than thorium in

loose materials, as in the dose assessment for disposal in landfills, due to the physical form of

the lamps.

Based on the assumptions described above and the generic disposal methodology in

Appendix A.2, the following estimates of individual and collective doses from disposal in

incinerators are obtained:

The EDE to individual waste collectors would be 1×104

 mSv (0.01 mrem). The annual

EDE to individual workers at incinerators, and the dose to individual off-site residents

near incinerators would be less than 1×105

 mSv (<0.001 mrem).

The collective EDE from 1 year’s disposals of lamps would be 2×105

 person-Sv

(0.002 person-rem), due almost entirely to exposure to waste collectors.

3.8.4.3.3 Reclamation and Recycling

As indicated in Section 3.8.3, the AEC concluded that reclamation and recycling of thorium in

germicidal lamps, sunlamps, and lamps for outdoor or industrial lighting were very unlikely. 

However, given the much greater emphasis in recent years on reuse of discarded materials,

compared with the time when this exemption was established, and considering the presence of

significant amounts of reclaimable material in some of the types of lamps covered by this

exemption, an assessment of potential doses from reclamation and recycling has been

performed in this analysis.
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The dose assessment for reclamation and recycling of lamps containing thorium is based on

the generic methodology presented in Appendix A.2. Potential doses to individual workers at a

smelter and members of the public residing near a smelter are considered.

The extent to which the types of lamps covered by this exemption are recycled is not known. In

this assessment, lamps containing a total of 1600 g are assumed to be recycled.

Based on the assumptions described above, the estimates of individual doses from reclamation

and recycling of 1 million lamps per year are summarized as follows:

For individual workers at a smelter, the EDE would be 8×105

 mSv/yr (0.008 mrem/yr).

For off-site members of the public, the individual EDE would be less than 1×105

 mSv

(<0.001 mrem).

These dose estimates should be quite conservative, because it seems highly unlikely that as

many as 1 million lamps of the type covered by this exemption would be recycled per year.

3.8.4.4 Accidents

Two types of accident scenarios involving lamps are considered in this assessment: (1)

breakage of a lamp and (2) a fire in a home, small storage room, or warehouse. In each case,

the lamps are assumed to contain 8 mg of thorium for the average lamp or 2 g of thorium, the

maximum amount allowed under this exemption, and doses are estimated using the generic

methodology described in Appendix A.1. All dose estimates include the contributions from

220Rn although, in most cases, this contribution is negligible compared with the dose from other

radionuclides in the 232Th decay chain.

3.8.4.4.1 Breakage of a Lamp

In this assessment, a single lamp is assumed to be broken in a room in a home. An upper

bound estimate of the dose to an individual who is in the room when the lamp breaks and who

cleans up the broken lamp is obtained by assuming that the model developed in Appendix A.1

for the dose from a spill of material in powder form applies to this scenario. This assumption

should be conservative, because the amount of thorium transferred from a broken lamp into the

air or to the hands of an individual should be considerably less than the amounts for a spill of

material in powder form. In estimating the dose following breakage, the room volume is

assumed to be 30 m3

, the air ventilation rate is 1 h1, the breathing rate is 1.2 m3/h, and the

exposure time is 0.5 h.

Based on the assumptions described above, the EDE to an individual from inhalation and

ingestion following breakage of a lamp containing the average 8 mg would be 3×104

 mSv

(0.03 mrem). The EDE to an individual for breakage of a lamp containing 2 g of thorium would

be 0.07 mSv (7 mrem).

3.8.4.4.2 Fire in Home, Storage Room, or Warehouse

In the generic methodology described in Appendix A.1, doses are estimated for a fire in a

home, a small storage room, or a large warehouse. For each scenario, doses are estimated for
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individuals during the fire, either the occupant of a home or storage area or a firefighter during a

warehouse fire, as well as individuals involved in cleanup activities after the fire. Results of the

dose assessment based on the generic methodology are as follows.

For a fire in a home, a room (27 m3

) where the fire occurs is assumed to contain a single lamp

containing 8 mg of thorium. The EDE to an individual without respiratory protection in the room

during the fire would be 3×104

 mSv (0.03 mrem), and the dose during cleanup after the fire

would be less than 1×105

 mSv (<0.001 mrem). For a fire involving a lamp with 2 g of thorium,

the EDE would be 0.07 mSv (7 mrem).

For a fire in a small storage room, the room is assumed to contain 100 lamps. The EDE to an

individual during the fire would be 0.001 mSv (0.1 mrem), and the dose during cleanup after the

fire would be 2×105

 mSv (0.002 mrem). For a fire involving lamps containing 2 g of thorium,

the EDE to a firefighter would be 0.4 mSv (40 mrem). The dose estimate during the fire would

be lower if the individual uses effective respiratory protection.

For a fire in a large warehouse, the storage area is assumed to contain 10,000 lamps, or 5% of

the assumed distribution of 1 million lamps per year. This assumption should overestimate the

number of lamps likely to be stored in any warehouse. The EDE to a firefighter using

respiratory protection during the fire would be 3×105

 mSv (0.003 mrem). The dose to an

individual not using respiratory protection during cleanup after the fire would be 2×104

 mSv

(0.02 mrem). For a fire involving lamps containing 2 g of thorium, the EDE to a firefighter would

be 0.007 mSv (0.7 mrem).

3.8.5 Summary

This assessment has considered doses to the public from use and disposal of germicidal

lamps, sunlamps, and lamps for outdoor or industrial lighting that contain up to 2 g of thorium. 

Results of the assessment are presented in Table 3.8.1 and are summarized as follows:

Doses to individuals during distribution and transport, routine use, and disposal in

landfills or incinerators appear to be low. The low individual doses during routine use

are due primarily to the assumptions that individuals would not be located close to lamps

in a sports arena or factory and, for a sunlamp in a home, that individuals would use the

lamp for only short periods of time during the year and would not be located close to the

lamp for significant periods of time when it is not in use. However, the collective dose

could be the highest during routine use, due primarily to the assumptions about the

number of individuals who would be exposed in sports arenas and factories. The

estimated collective dose from disposal in landfills would be reduced by more than two

orders of magnitude if the dose to future on-site residents for 1000 years after facility

closure were not taken into account.

All estimates of individual and collective doses for routine exposure scenarios, except

doses to individuals during routine use, are based on an assumption that 1 million lamps

per year would be distributed, transported, used, disposed, or recycled. However, the

assumed number could be a considerable overestimate. If the actual number of lamps

containing thorium were known, the dose estimates for these scenarios could be

adjusted by the appropriate amount.
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This assessment also considered doses to individuals resulting from accidental breakage of a

lamp and a fire in a home, a small storage room, and a large warehouse. Potential doses from

breakage of a single lamp appear to be low, due to the likelihood that only a small fraction of

the thorium in a lamp would be inhaled or ingested. Potential doses from a fire in a home,

small storage room, or large warehouse also appear to be high, due to the amounts of thorium

that are assumed to be involved in each case and the assumption that individuals fighting a fire

in a warehouse would not use respiratory protection. In addition, the dose estimates for a

warehouse fire are likely to be conservative, because it seems unlikely that 5% of the total

annual distribution of lamps would be stored in a single warehouse.
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Table 3.8.1 Summary of Potential Radiation Doses From Use of Thorium-Containing

Germicidal Lamps, Sunlamps, and Lamps for Outdoor or Industrial Lighting a

Exposure Scenario

Individual Annual Effective

 Dose Equivalent

(mrem)c

Collective Effective

Dose Equivalentb

(person-rem)c

Distribution and transportb 0.007d 0.02

Routine use

 Sports arena

 Factory

 Sunlamp in home

<0.001

<0.001

<0.001

0.4e

Disposalb

 Landfills

 Incinerators

 Recycling

0.001f

0.01h

0.008i

2g

0.002

NAm

Accidents or misusej

 Lamp breakage

 Home fire

 Storage room fire

 Warehouse fire

0.03

0.03

0.1k

0.02l

NA

See following page for footnotes.
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Footnotes to Table 3.8.1

a

 Dose estimates are based on assumption that all lamps contain 8 mg of thorium for the

average lamp. Actual amounts of thorium in products are unknown. Dose estimates for

distribution and transport, routine use, and accidents or misuse assume that thorium has been

aged for 20 years after chemical separation. Dose estimates for disposal assume that all decay

products of 232Th are in activity equilibrium.

b

 Dose estimates are based on assumption that 1 million lamps containing thorium are

distributed, used, disposed, or recycled per year.

c

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

d

 Dose estimate applies to truck drivers; dose estimates for UPS terminal employees, retail

employees, or warehouse employees are somewhat less, and dose estimate for customers in

retail stores is considerably less (see Section 3.8.4.1 and Table 3.6.1).

e

 Dose estimate is based on assumed admixture of lamps in sports arenas, factories, and

homes described in Section 3.8.4.2.4.

f

 Dose estimate applies to waste collectors at landfills; dose estimates are considerably less for

workers at landfills, off-site members of the public, and future on-site residents (see

Section 3.8.4.3.1).

g

 Dose estimate applies to future on-site residents at landfills for 1,000 years after disposal due

to 1 year’s disposals; if dose to future on-site residents is not taken into account, estimate

annual collective dose to waste collectors and workers at landfills is 8×105

 person-Sv

(0.008 person-rem).

h

 Dose estimate applies to waste collectors at incinerators; dose estimates are considerably

less for workers at incinerators and off-site members of the public (see Section 3.8.4.3.1).

i

 Dose estimate applies to workers at smelters; dose estimate is considerably less for off-site

members of the public (see Section 3.8.4.3.3).

j

 Dose estimates apply to single occurrence of accident or misuse scenario; number of lamps

assumed to be involved in each scenario is described in Section 3.8.4.4.2.

k

 Dose estimate applies during firefighting without respiratory protection; dose estimate is

considerably less during cleanup after fire (see Section 3.8.4.4.2).

l

 Dose estimate applies during cleanup after fire; dose estimate is somewhat less during

firefighting using respiratory protection during fire (see Section 3.8.4.4.2).

m Not applicable.
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3.9 Rare Earth Products Containing Less Than 0.25% by Weight 

of Source Material

3.9.1 Introduction

In 10 CFR 40.13 (c)(1)(vi), any person is exempt from the requirements for a license, to the

extent that the person receives, possesses, uses, or transfers rare earth metals and

compounds, mixtures, and products containing not more than 0.25% by weight of thorium,

uranium, or any combination of these. This exemption was first established on March 20, 1947

(12 FR 1855). The exemption was deleted on January 14, 1961 (26 FR 284), but reinstated on

April 11, 1961 (26 FR 3063).

The original Federal Register notice contained no information on analyses of radiological

impacts on the public from use of the exempted rare earth materials. In the 1961 notice

reinstating the exemption, the Atomic Energy Commission referred to the radioactivity exposure

data furnished by the American Potash and Chemical Corporation and concluded that the

possession and use of source material in these materials can be conducted without an

unreasonable hazard to life or property. However, no quantitative estimates of radiological

impacts were given.

This exemption refers only to the weight percent of any combination of thorium and uranium. 

However, thorium is the primary radioactive element typically found in rare earth products. 

Naturally occurring and processed thorium normally consists primarily of 232Th (nearly 100% by

mass and 50% by activity) and 228Th (50% by activity). If it is assumed that 0.25% by weight of

a rare earth product is thorium and 232Th and 228Th are in equilibrium, the activity concentration

of 232Th and 228Th would be about 10 becquerel (Bq)/g (275 picocurie (pCi)/g) of each. This

concentration refers only to the amount of thorium in the material of interest, but the thorium

decay products also must be considered when estimating doses.

3.9.2 Description of Items

The rare earths are defined as a group of 17 elements composed of scandium, yttrium, and the

lanthanides (Hedrick, 1993). Scandium, atomic number 21, is the lightest rare earth. Yttrium,

atomic number 39, is chemically similar to the lanthanides and commonly occurs in the same

minerals. The lanthanides comprise a group of 15 elements with atomic numbers 57 through

71. Due largely to differing ionic radii, the rare earth elements are broadly classified into two

groups: (1) the light rare earth elements (LREE) or cerium subgroup, composed of the first

seven lanthanides (atomic numbers 57 through 63), and (2) the heavy rare earth elements

(HREE) or yttrium subgroup, composed of the remaining lanthanides (atomic numbers 64

through 71) and yttrium (atomic number 39). Yttrium is classed with the heavy subgroup

because of its occurrence, ionic radius, and other similar properties.

The rare earths are constituents of more than 100 minerals, but only a few are recovered for

commercial production. Monazite, bastnasite, xenotime, and rare earth-bearing clays are the

major sources of the world’s rare earth supply (Hedrick, 1993). The following subsections

briefly describe the primary sources of rare earth products, monazite and bastnasite, as well as,

key products derived from or containing rare earths.
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3.9.2.1 Monazite

Monazite, a rare earth phosphate, is one of the most abundant rare earth minerals. It is

classified as an LREE and is usually enriched in cerium (Hedrick, 1993). Heavy-mineral sands

are the source of most monazite mined in the world. Monazite is mined either as a primary

product or as a byproduct of titanium, zirconium, or tin minerals. Monazite also was recovered

from tailings at Jackson, NJ, which were reprocessed primarily for their zircon and ilmenite

content. 

From a radiological perspective, the mineral monazite is of particular importance because it

typically contains 5 to 7% thorium and 0.1 to 0.3% uranium (Hewson and Terry, 1995). In the

United States, the thorium content in monazite is about 4% (Phone call, J. Raiser, RGC

Minerals and Sands, Inc., Green Cove Springs, FL, November 1994). Assuming a 5% thorium

content, the 232Th and 228Th activity concentration in ore is about 204 Bq/g (5500 pCi/g). 

Assuming a 0.2% uranium content, the natural uranium (238U, 235U, and 234U) activity

concentration is about 50 Bq/g (1360 pCi/g).

The major monazite-producing countries are Australia, Brazil, China, India, Malaysia, Thailand,

and the United States (Hedrick, 1994). Worldwide demand for thorium-bearing rare earth ores

is low (Hedrick, 1997). Its popularity has diminished because of the potential waste problem

and possible radiological hazards. Domestic mine production of monazite ceased at the end of

1994, and monazite was not imported into the United States from 1992 to 1994 (Hedrick, 1997). 

In 1995 and 1996, monazite was imported into the United States from Australia and Malaysia

(Hedrick, 1997).

3.9.2.2 Bastnasite

Bastnasite, a light lanthanide (Ln) fluoride carbonate, occurs in an unusual type of magmaderived deposit in which the Ln elements have been enhanced (Kilbourn, 1992). The largest

recognized deposit, which is mixed with monazite and iron ores, occurs in a complex

mineralization at Bayan Obo in Inner Mongolia, China. The only other current commercial

bastnasite source is Molycorp’s Mountain Pass, CA, deposit, where the average Ln oxide

content of the ore is about 9% (Kilbourn, 1992). The U.S. deposit is the only resource mined

solely for its content of cerium and other lanthanides. 

The thorium content in bastnasite can range from about 0.02 to 0.1% by weight and the

uranium content is about 0.002% by weight (Phone call, W. Almas, Manager of Environmental

and Regulatory Affairs, Molycorp, Inc., Mountain Pass, CA, November 1994). Therefore, the

232Th and 228Th concentration in ore ranges from about 0.8 to 4 Bq/g (22 to 110 pCi/g) and the

natural uranium (238U, 235U, and 234U) activity concentration is about 0.5 Bq/g (14 pCi/g). 

Depending on the purity of the particular rare earth product, the thorium and uranium content

can vary from trace amounts up to about 0.25% by weight (primarily as thorium phosphate)

(Phone call, S. Trout, Molycorp, Inc., Fairfield, NJ, November 1996). 

3.9.2.3 Rare Earth-Derived Products

A wide variety of rare earth products are available for trade and industry, including

concentrates, individual and mixed compounds, and pure and alloyed metals (Hedrick, 1993). 

Rare earths are used in hundreds of applications, ranging from 19th-century traditional
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applications to high-technology uses. The applications encompass a wide range of products

from mixed rare earth compounds and alloys to ultrahigh-purity individual metals and

compounds. Quantities vary by application and range from tonnage quantities to milligrams. 

Table 3.9.1 provides a summary of the products that contain a rare earth element. The

approximate distribution of rare earths by use, based on information supplied by primary

processors and some consumers, is summarized in Table 3.9.2.

Individual rare earth oxides and other compounds, except promethium, are produced from

bastnasite, monazite, and xenotime ore (Hedrick, 1993). Purities from 96% up to 99.9999% are

available for most oxides. Compounds for nearly all of the rare earths are available as oxides,

acetates, carbonates, chlorides, fluorides, nitrates, oxalates, and sulfates. High-purity rare

earth metals are marketed in the forms of sponges, lumps, ingots, crystals, rods, wires, chips,

powders, sheets, foils, plates, sputtering plates, and custom cast and machined shapes. Alloys

such as mischmetal (mixtures of rare earth elements in metallic form), rare earth silicide,

ferrocerium, and other rare earth alloys are available in a variety of ingot shapes and sizes. 

Rare earth magnet alloys are marketed in ingot form and crushed ribbon, or may be purchased

as mixed oxides for powder metallurgical processes. 

Cerium is the most abundant member of the series of lanthanides. It is characterized

chemically by having two stable valence states, Ce4+ and Ce3+, and this property underlies

several technological uses (Kilbourn, 1992). Bastnasite and monazite are the two major

mineral sources for cerium. Of the worldwide production of 64,000 metric tons (MT) of

lanthanides, only a small fraction is produced as relatively pure individual Ln derivatives,

including cerium (Kilbourn, 1992). The majority of the lanthanides are used as concentrates

and like products. The various cerium-containing materials available commercially are

summarized in Table 3.9.3. Cerium oxide slurried in water is the most efficient glass polishing

agent, especially for glass compositions produced commercially in large volume (Kilbourn,

1992). Other major uses of cerium compounds include the decolorization of glass, production

of radiation-resistant glass (television glass faceplates and hot cell viewing windows), and

production of ultraviolet absorption glass (such as medical glassware and selected types of

automobile glass) (Kilbourn, 1992).

The scandium industry, for the most part, is separate and distinct from the lanthanide and

yttrium industry (Hedrick, 1993). Most scandium is produced as a byproduct of other ores. In

the United States, scandium has been recovered as a byproduct during processing of uranium

at a copper mine, from tailings from a mined-out fluorite mine, and from a tungsten byproduct

from a molybdenum operation (Hedrick, 1993). Tailings and concentrates from these sites

were shipped to processors for production of high-purity scandium oxide and other compounds. 

Most scandium is sold in oxide form. A small group of companies produces scandium metal,

primarily for lighting and research purposes.

Yttrium is often classed with the rare earth elements because it invariably occurs with them in

nature and has similar properties (Hedrick, 1993). Xenotime concentrate with a minimum 25%

yttrium oxide content is a major source of yttrium (Hedrick, 1993). Except for minor amounts of

yttrium contained in domestically produced bastnasite concentrates, essentially all purified

yttrium was derived from imported compounds (Hedrick, 1995).
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3.9.3 Summary of Previous Analyses and Assessments

When this exemption was initially established and then reinstated, little information was

provided on the radiological impacts on the public from use of the materials. The 1961 Federal

Register notice reinstating the exemption refers to the radioactivity exposure data furnished by

the American Potash and Chemical Corporation (APCC), but no quantitative estimates of

radiological impacts are given. However, the radioactivity exposure data for selected rare earth

processing workers at APCC’s West Chicago, Illinois, facility were obtained and are

summarized in this section. 

Other than the information available on ophthalmic glass, there is no known information

available assessing the potential radiological impact on members of the public (industrial

workers and other members of the public) from the use of rare earth products and materials. 

Most of the exposure data and studies have evaluated workers processing mineral sands. For

example, a number of studies were conducted in Australia to evaluate the radiological impacts

associated with the mineral sands industry. In these studies, the processing of mineral sands

to extract monazite and other minerals, which can contain between 5 to 7% by weight of source

material, would be licensed operations. However, this information can be extrapolated, where

applicable, to similar operations in which bulk rare earth products containing 0.25% by weight of

source material may be handled. Air concentrations, bioassay results, and doses to industrial

workers involved with the processing of mineral sands and subsequent extraction of heavy

minerals, such as monazite, are summarized. Also summarized are the radiological impacts

from the processing and handling of rare earth products derived from bastnasite ore, evaluated

in the Molycorp Mountain Pass facility characterization study. The potential on-site and off-site

radiological impacts from disposal of rare earth processing wastes were estimated by the EPA

(RAE–9232/1–2). 

3.9.3.1 Routine Exposures

This section presents a brief summary of the exposure data applicable to industrial workers and

off-site members of the public impacted by the routine handling and processing of monaziteand bastnasite-derived rare earth products, respectively. The processing and handling of

monazite-derived rare earth products are usually licensed operations; however, the exposure

data are useful when evaluating the handling of rare earth products in similar nonlicensed

operations.

3.9.3.1.1 Industrial Worker 

3.9.3.1.1.1 Monazite-Derived Rare Earth Products

In 1961, APCC’s West Chicago, Illinois, plant was the largest refiner in the United States of

thorium, rare earth chemicals, and yttrium. Lindsay Chemical Company, which previously

owned the West Chicago plant, was acquired by APCC in 1958. The rare earth feedstock for

the West Chicago plant was rare earth sodium sulfate, which contained 42% rare earth oxide

(REO). Most of the production was in the form of rare earth and cerium products, such as rare

earth chloride, rare earth oxide, rare earth fluoride, cerium hydrate, and various rare earth- and

cerium-based glass polishing compounds (McCoy, 1961).
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The radiation protection program at the APCC plant included film badges and air concentration

measurements. All operating personnel were required to wear badges. Many of these

employees transferred between the thorium (licensed) and rare earth (nonlicensed) sections of

the plant. Table 3.9.4 summarizes the annual doses from direct irradiation to personnel

working with crude rare earth products and purified rare earths in ion exchange columns. 

These doses were associated with activities that involved handling, preparation, and packaging

of rare earth materials, such as rare earth and cerium oxide, cerium hydrate, rare earth sodium

sulfate, rare earth chloride, and rare earth fluoride. The main effort of the air sampling program

was directed toward operations involving handling of thorium materials. However, there was a

record of selected air concentration results associated with the rare earth processing plant

(McCoy, 1961). The air concentration data were presented as a percent of the maximum

permissible level (% ML). The average % ML and derived air concentrations (DACs) for these

selected rare earth processing activities are summarized in Table 3.9.5.

In the mineral sands industry, work tasks involving dry particulate product handling pose a

particular risk because very high localized dust concentrations may be generated in the

breathing zone of workers. These tasks include bagging operations, manual cleaning methods

(e.g., shoveling, sweeping, and brushing), and certain maintenance activities. According to

Hewson and Fardy (1993), long-term workers may have been exposed to thorium ore dust at

significant levels of intake, i.e., on the order of the annual limit on intake. Based on this

possibility, bioassay studies were performed to complement estimates of radiation doses

derived from air sampling measurements. Table 3.9.6 provides a summary of the number of

personal air samples taken from designated industry workers (those with the potential to

receive an annual effective dose equivalent (EDE) greater than 5 millisieverts (mSv)

(500 mrem)) during the time period of 1986 to 1992, together with monazite production figures

and an estimate of the industry wide average alpha activity airborne concentration. According

to Hewson and Terry (1995), monazite preferentially concentrates in dust by a factor of

between 10 and 30.

Average alpha activity levels have decreased substantially since 1986, when the average daily

intake of alpha activity was estimated to be 8.5 Bq (235 pCi/g) (assuming 10 m3

 of air breathed

per work shift). Through a combination of engineering practices and improved work and

management practices, the average daily intake is now estimated to be 1.6 Bq (44 pCi)

(Hewson and Fardy, 1993). Analysis of the data also showed that the daily variability in the

alpha activity concentration is described by a geometric standard deviation (gsd) in the range of

2.0 to 2.5 for workers and supervisory staff and a gsd in the range of 3.0 to 3.5 for workers

involved in maintenance activities in the dry plant (Hewson and Fardy, 1993).

Serum and urine samples were collected from both unexposed workers and from 34 mineral

sands workers selected for the study (Hewson and Fardy, 1993). For just over half (56%) of

the mineral sands workers, the concentration of thorium in urine was above the upper end of

the range observed for unexposed workers. Thirty-one of the 34 mineral sands workers

recorded a concentration of thorium in urine well above the geometric mean of the background

concentration. 

The conclusions of this bioassay study indicated that the absorption of thorium is considerably

less than that predicted by the assessment methods. The discrepancies between measured

and expected bioassay values are due to protective effects of respirators, and also may be due
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to inappropriate model parameters such as lung solubility and clearance of large particles to

blood (Hewson and Fardy, 1993).

Marshman and Hewson (1994) compiled individual and collective radiation doses associated

with the Western Australian mineral sands industry between 1986 to 1993 for designated and

nondesignated employees. Designated employees are those workers directly involved in the

operation and maintenance of the processing plants. Nondesignated employees include

workers involved in mining, transport, administrative, and technical services. For the

designated employees, the 1992 annual mean external dose, as measured using

thermoluminescent dosimeters (TLDs) was 1.5 mSv (150 mrem). The 1992 annual mean

internal dose for designated employees was estimated to be 6.3 mSv (630 mrem). As stated in

Marshman and Hewson (1994), respiratory equipment use was not assumed. Also, the authors

cautioned that the internal dose estimates have significant uncertainty and should be

interpreted with caution. Based on 212 designated workers, the 1992 annual collective EDE

was estimated as 1.7 person-Sv (170 person-rem). 

Reductions in airborne radioactivity occurred across the industry from 1986 to 1993. The mean

airborne radioactivity concentrations decreased from 0.85 Bq/m3

 (23 pCi/m3) in 1986 to

0.18 Bq/m3

 (5 pCi/m3) in 1992. These mean airborne radioactivity concentrations were

obtained from personal air samples of workers involved in the operation and maintenance of

five mineral sands processing facilities. 

Hewson and Hartley (1990) reported levels of 222Rn and 220Rn in air in various mineral sands

mining and processing stages. The data were consistent with natural background levels, due to

the relatively low uranium content and the low emanation rate of radon. However, in Hewson

(1993), it is stated that routine monitoring for radon progeny may be necessary for monazite

processing, since grinding and cracking of the mineral matrix will increase the emanation rate of

radon. According to Hewson (1993), the realistic apportionment of dose between various

exposure pathways was 50% due to external irradiation and 50% due to internal radiation

(comprising 30% from long-lived alpha emitters and 20% from 222Rn and 220Rn progeny). 

3.9.3.1.1.2 Bastnasite-Derived Rare Earth Products

Radiation dosimetry data were collected at the Molycorp’s Mountain Pass California facility

during a characterization study (Bernhardt, 1996). TLDs were worn by plant workers and

placed in selected areas of the facility. The personnel TLD results for selected plant workers

are summarized in Table 3.9.7, and the area TLD results are summarized in Table 3.9.8. 

Radiation surveys for surface contamination and exposure rate measurements were also

conducted in selected nonproduction plant areas. The highest average exposure rate

measurement was 18 nanocoulomb (nC)/kg-h (70 microroentgen (R)/h) (including

background), observed on the bastnasite warehouse floor. Typically, the average exposure

rate measurements were between 3 and 10 nC/kg-h (13 to 40 R/h). The total surface

contamination activity ranged from zero (0) (chemical plant control room) to 440 disintegrations

per minute (dpm)/100 cm2

 (maintenance lunch room floor). The smearable contamination

ranged from zero (0) (mill lunch area) to 48 dpm/100 cm2

 (bastnasite warehouse floor).

The radionuclide air concentration data available for the Molycorp Mountain Pass facility was for

the lead and iron filter cake re-introduction processing project. The lead and iron filter cake is a

waste product that contains greater than 0.05% by weight of uranium. The lead and iron filter
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cake was added into the rare earth processing stream. During this re-introduction project, the

average monthly gross alpha air concentrations in the chemical processing and warehouse

areas were about 0.01 Bq/m3

 and 0.002 Bq/m3 (3×1013 Ci/cc and 4×1014 Ci/cc),

respectively. The gross alpha air concentrations did not exceed the action level of 0.06 Bq/m3

(1.5×1012 Ci/cc), which is about one-tenth of the DAC for natural uranium 0.7 Bq/m3

(2×1011 Ci/cc).

A radiation survey was conducted by Molycorp on a truck load containing three drums of

bastnasite concentrate and 14 pallets of cerium concentrate (Zapolski, 1985). Inside the

loaded trailer, the exposure rate readings next to the drums of bastnasite concentrate and

between the pallets of cerium concentrate were about 50 nC/kg-h (0.2 mR/h) and 100 nC/kg-h

(0.4 mR/h), respectively. Distance estimates between the source and the measurement points

were not provided. Exposure rate measurements taken in the truck cab and outside of the

loaded trailer ranged from 3 to 5 nC/kg-h (0.01 to 0.02 mR/h) (including background).

3.9.3.1.2 Members of the Public

Exposures to members of the public (other than industrial workers) result from the processing

of rare earths and from the commercial use of rare earth products. Environmental TLDs have

been used to assess off-site direct irradiation doses from rare earth processing facility

operations. Results of the environmental TLD programs are summarized in this subsection. 

The extent of off-site contamination resulting from past thorium and rare earth processing

waste operations are briefly highlighted. Other than the information available on ophthalmic

glass, there is no known reported information available for assessing the potential radiological

impact on the public from the use of rare earth products.

Environmental TLDs were placed in several work and nonwork areas in and around the

Molycorp Mountain Pass Facility (Bernhardt, 1996). The primary purpose of the environmental

TLD program was to evaluate the potential impacts from the Molycorp Mountain Pass lead/iron

filtercake project. However, the results of the environmental TLD program also include direct

irradiation levels associated with nonlicensed operations. Results of this survey are presented

in Table 3.9.9. The estimated annual EDE from direct irradiation at the plant boundary was

1.6 mSv (160 mrem). According to Bernhardt (1996), this annual EDE is essentially at ambient

background radiation levels. 

The Texas Department of Health, Bureau of Radiation Control (BRC), performs routine

environmental monitoring at sites throughout Texas, based on either the potential for

environmental release of radioactive materials or at sites that were contaminated from

operations conducted when controls were less stringent. Rhone-Poulenc in Freeport, TX is a

company that processes rare earth hydroxides. The company is licensed to possess and store

these materials prior to disposal. At one time, there were many drums containing radioactive

waste stored on-site, including waste from rare earth processing operations. Off-site shipment

of the waste products was still ongoing in 1992. Ambient gamma levels were monitored using

TLDs, and sediment, soil, vegetation, and surface water samples were collected around the

Rhone-Poulenc facility (BRC, 1992). The annual dose equivalent measured by the TLDs at the

fence line ranged from about 0.2 to 15 mSv (20 mrem to 1500 mrem). At two stations, the

annual radiation levels exceeded 5 mSv (500 mrem). However, occupancy at these stations

was low enough that no member of the public would likely receive an annual dose equivalent in
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excess of 5 mSv (500 mrem). No off-site contamination was detected at this facility (BRC,

1992).

A legacy of on-site and off-site contamination has resulted from past thorium and rare earth

extraction and processing operations. One of the sites still undergoing remediation is APCC’s

West Chicago, Illinois, facility. Wastes consisting of a tailings pile, five sedimentation ponds,

and a sludge pile of material dredged from sedimentation ponds are located in a residential

community, in some cases only 140 meters from occupied homes (Jenson, 1980). Currently,

there are four off-site Superfund remediation activities associated with the West Chicago facility

(Phone call, D. Seely, Project Manager, U.S. Environmental Protection Agency, Chicago, IL,

December 1996). The thorium and rare earth processing operations of the Maywood Chemical

Works (formerly the Stephan Company) also resulted in both on-site and off-site contamination

(Foley et al., 1989; Ring et al., 1989).

As mentioned in Section 3.2, the manufacture of ophthalmic glass frequently uses mixtures of

rare earths and zirconium oxides containing natural uranium and thorium. Buckley et al.

(NUREG/CR–1775) performed an analysis of radiological impacts on the public from

eyeglasses containing thorium. Analyses of doses from wearing of eyeglasses containing

0.05% by weight thorium considered exposure to the eye, primarily from irradiation by alpha

particles, and whole-body exposures from photons. Dose estimates reported by Buckley et al.

(NUREG/CR–1775) and the National Council on Radiation Protection and Measurements

(NCRP 95) assumed that an individual wears eyeglasses for 16 h/day and that 55% of the

population wears glasses. These dose estimates are summarized as follows:

The annual dose equivalent to the cornea from alpha particles is about 0.04 Sv (4 rem). 

Assuming a tissue weighting factor of 104

, the annual individual EDE from exposure to

the cornea was estimated to be 0.004 mSv (0.4 mrem).

The annual dose equivalent to the whole body of the individual from photons is about

2×104

 mSv (0.02 mrem).

NCRP 95 estimated that 50 million people in the United States wear eyeglasses with glass. 

Assuming an annual individual EDE of 0.004 mSv (0.4 mrem) from alpha particles and a wholebody dose equivalent from photons of 2×104

 mSv (0.02 mrem), the annual collective EDE to

the U.S. population would be about 200 person-Sv (20,000 person-rem). 

3.9.3.2 Disposal

Very little is known about the disposal of rare earth products. One source of information is an

inspection report (New York State, 1994) that identified elevated radiation levels associated with

slag from the production of additives rejected by two steel companies. Most of the available

information is on the disposal of rare earth processing wastes. For example, the EPA

(RAE–9232/1–2) conducted a generic dose assessment that evaluated the potential on-site and

off-site impacts from disposal of monazite-derived rare earth processing wastes. In addition, a

draft environmental impact report (ENSR, 1996) was prepared that describes the 232Th and 238U

activity concentrations in bastnasite-derived rare earth processing wastes. In the draft

environmental impact report, annual waste production estimates are given for the Molycorp

Mountain Pass facility. However, the initial processing of monazite would be licensed, and the
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initial processing of bastnasite would likely fall under 10 CFR 40.13 (a). Disposal of monazite

and bastnasite processing wastes would be similarly regulated. However, this information is

potentially useful for evaluating the EDEs attributed to the disposal of bulk rare earth products.

SKW Alloys of Niagara Falls, NY, produced additives for steel manufacturing (New York State,

1994). On April 26, 1994, a load of additives was rejected by a steel company because

radiation alarms were set off. On April 28, 1994, a second load was rejected from another

plant. These shipments were returned to SKW. Inspection of the slag from the first rejected

load showed elevated radiation levels (maximum reading of 77 nC/kg-h (0.3 mR/h) was

measured). SKW suspected that the problem derived from a previously prepared additive that

contained cerium, so radiation surveys were conducted. The highest radiation levels measured

in a room containing slag ranged from 1.7×104

 to 1.9×104 mSv/h (0.17 to 0.19 mrem/h). 

Samples of slag, metal, and cerium product were collected and analyzed. The 228Ra and 226Ra

activity concentrations from the slag were greater than that found in the cerium product or in the

metal additives, as shown in Table 3.9.10. The slag also contained lower concentrations of

224Ra as compared to 228Ra, indicating that the thorium decay chain was not in equilibrium. 

Thorium was not analyzed in any of the samples. Off-site samples contained only background

levels of 228Ra, 224Ra, and 226Ra radioactive materials.

Mineral processing facility waste is generally disposed in tailing ponds or used to construct

dams, dikes, or embankments. Small amounts of waste have been used off-site for backfill,

aggregate production, or road building (EPA, RAE–9232/1–2). The EPA (RAE–9232/1–2)

conducted an assessment of a generic REO mine and mill site located in a hypothetical site in

North Carolina, where ores associated with REO are composed of 50% monazite sands and

have elevated levels of thorium and uranium. The model site consists of commingled waste

rock, overburden, and a tailings pile of 1 hectare. The tailings have no further use other than

reprocessing for their mineral content. The assessment assumes 200,000 MT of tailings,

overburden, and wastes and that a 1-meter thick cover is placed over the waste when the site is

closed. At the generic site, the 232Th, 228Th and 228Ra activity concentrations were 74 Bq/g

(2000 pCi/g) and the 238U, 234U, 230Th, and 226Ra activity concentrations were 33 Bq/g

(900 pCi/g). The 210Pb, and 210Po activity concentrations were 23 Bq/g (630 pCi/g) and the 235U,

231Pa, and 227Ac activity concentrations were 1.7 Bq/g (45 pCi/g). This source term was

considered to be representative of rare earth processing waste streams (EPA, RAE–9232/1–2).

Annual individual EDEs are estimated for individuals residing on the disposal site. Annual

individual EDEs are also estimated for off-site individuals residing 100 meters from a disposal

site. The exposure pathways evaluated for the on-site individuals residing on the disposal site

are direct irradiation and dust inhalation. Ingestion of food grown on-site was not considered

since the rare earth wastes generally do not have the textural and nutritional properties, and

water-retention capabilities to support vegetation (EPA, RAE–9232/1–2). A number of potential

exposure pathways were evaluated for the individual residing 100 meters from a disposal site. 

The estimated annual individual EDEs for these on-site and off-site receptors are summarized

in Table 3.9.11.

Annual collective dose equivalents from disposal of rare earth processing waste estimated by

the EPA (1993) are summarized in Table 3.9.12. The population density for atmospheric

pathways was assumed to be 136 persons per square mile and the population using river water

as a drinking water supply was assumed to be 235,000 (EPA, RAE–9232/1–2).
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Molycorp commissioned a study to inventory and analyze various wastes, feedstock, and

products associated with the Mountain Pass California facility (ENSR, 1996). Table 3.9.13

shows the average concentrations of 232Th and 238U in the feedstock bastnasite ore, in

bastnasite concentrate, and in some of the wastes produced during rare earth processing. The

lead and iron filter cake contained source material in excess of the 0.05% by weight Nuclear

Regulatory Commission criterion (10 CFR 40.13(a)). Molycorp obtained an addendum to its

license for the possession and use of radioactive sealed sources to cover the management and

possession of uranium and thorium contained within the lead and iron filter cake (ENSR, 1996). 

Molycorp is required to feed stabilized lead and iron filter cake to the chemical process for the

lanthanide recovery or dispose of stabilized material at an approved disposal facility. Molycorp

generates small volumes of solvent extraction (SX) crud from the solvent extraction cells used

for the separation and recovery of lanthanides. SX crud is an insoluble organic complex

containing significant concentrations of both uranium and thorium, as shown in Table 3.9.13. 

Prior to 1984, Molycorp deposited lead and iron residues in three small surface impoundments. 

Molycorp is required to submit a plan for closure of these lead ponds. As a result of process

changes instituted by Molycorp, radionuclide concentrations within the lead sulfide concentrate

product are being held below 0.05% by weight for uranium and thorium combined (ENSR,

1996). The radiological impacts associated with the disposal of these rare earth processing

wastes were not provided in ENSR (1996).

3.9.4 Present Exemption Analysis

In the present exemption analysis, annual individual and collective EDEs are estimated for truck

drivers and industrial workers handling bastnasite and cerium concentrate. In addition, annual

individual and collective EDEs are estimated for the use of some consumer products that

contain rare earth products. These consumer products include pink ophthalmic glass, television

glass faceplates, and specialty automotive window glass. The generic disposal methodology is

used to estimate the individual and collective EDEs associated with the landfill disposal of

bastnasite-derived rare earth products. The generic accident methodology was used to

estimate EDEs associated with a warehouse fire containing cerium concentrate and a fire

during transport of cerium concentrate. 

3.9.4.1 Distribution and Transport

The results of the radiation survey described in Section 3.9.3.1.1.2 (Zapolski, 1985) are used in

the present assessment to estimate individual and collective annual EDEs to truck drivers

transporting bastnasite and cerium concentrate. The measured absorbed doses in the

Molycorp bastnasite warehouse area (Bernhardt, 1996) are used to estimate the annual EDEs

associated with the storage and distribution of bastnasite and cerium concentrate products. 

Bastnasite and cerium concentrates are assumed to be transported on pallets from California to

a product user located in the northeast. The direct irradiation exposure rate measurements

taken in the cab of a truck transporting 14 pallets (1350 kg per pallet) of cerium concentrate

(averaging about 0.25% by weight of thorium phosphate) and four drums of a bastnasite

concentrate (0.1% by weight of thorium phosphate) were 3 to 5 nC/kg-h (0.01 to 0.02 mR/h),

including background (Zapolski, 1985). If the natural radiation background is about 3 nC/kg-h

(0.01 mR/h), the exposure rate from shipment is assumed to be 3 nC/kg-h (0.01 mR/h). To

transport this shipment across the United States, it is assumed the truck driver is exposed for
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about 50 hours, with an estimated EDE of 0.005 mSv (0.5 mrem). If this driver makes 30

similar trips across the United States in 1 year, the estimated annual EDE is 0.2 mSv

(20 mrem). Industry sources indicate that approximately four shipments of product are

transported weekly to the eastern United States, resulting in about 200 annual shipments. 

Assuming all of these shipments are in a truck, the estimated annual collective EDE is

approximately 0.001 person-Sv (0.1 person-rem). These doses represent those near the

maximum allowed concentration which may also approximate the average use concentrations

for this product. 

To estimate the annual EDE to warehouse workers involved with the storage and distribution of

bastnasite and cerium concentrate, the measurements reported by Bernhardt (1996) are used

in the current assessment. According to Bernhardt (1996), the 40-hour absorbed dose was

estimated to be 0.02 milligray (mGy) (2 mrad) in the bastnasite bagging and warehouse area

(Table 3.9.8). It is assumed that the area TLDs were located in the main operational area. 

Bastnasite concentrate contains about 0.1% by weight of thorium phosphate. If workers spend

about 500 h/yr in the warehouse, the estimated annual EDE is 0.3 mSv (30 mrem).

Extrapolation of the 40-hour absorbed dose rate attributed to bastnasite concentrate (0.1% by

weight of thorium phosphate) to the storage of cerium concentrate (0.25% by weight of thorium

phosphate) results in an estimated 40-hour absorbed rate of 0.05 Gy (5 mrad). If a warehouse

worker spends about 500 h/yr in the warehouse, the estimated annual EDE is 0.6 mSv

(60 mrem).

Gross alpha airborne concentrations were sampled in warehouses A and B during the lead and

iron filter cake reintroduction project (Bernhardt, 1996). The measured average gross alpha air

concentrations were about 0.002 Bq/m3

 (4×1014 Ci/cc). Warehouses A and B are used to

store rare earth products (separate from the bagging operations) and lead and iron filter cake

that contained greater than 0.05% by weight of uranium. However, similar airborne alpha

activity concentrations were also measured within the guardhouse, which is considered to be a

background location. If it is assumed that the alpha activity is due to 232Th in equilibrium with its

progeny and that workers spend 500 h/yr in the warehouses that may contain bastnasite and

cerium concentrate, the estimated annual EDE due to dust inhalation is 0.46 mSv (46 mrem).

Therefore, the annual individual EDEs due to direct irradiation and dust inhalation to warehouse

workers handling bastnasite and cerium concentrates are estimated to be 0.7 and 1 mSv

(70 mrem and 100 mrem), respectively. It is unknown how many warehouse workers are

actually involved in rare earth product warehouse operations. If it is assumed that there are two

workers who spend 500 h/yr in a warehouse and there are about 100 facilities handling

bastnasite and cerium concentrates, the annual collective EDEs are assumed to be

0.1 person-Sv and 0.2 person-Sv (10 person-rem and 20 person-rem), respectively.

3.9.4.2 Routine Operations

In this subsection, the annual individual and collective EDEs from routine exposures to

industrial workers and other members of the public are estimated. For the industrial worker

scenario, the bagging operations associated with a bastnasite concentrate containing 0.1% by

weight of source material and a cerium concentrate containing 0.25% by weight of source

material, as thorium, are evaluated. This job activity is used to represent the upper bound of

potential industrial worker exposures, especially for workers handling bulk rare earth product. 
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Similar types of exposures may be possible for maintenance personnel. In addition, annual

individual and collective EDEs are calculated for the viewing of television with glass faceplates

and for the use of specialty automotive window glass. Individual and collective EDEs from pink

ophthalmic lenses are based on previous dose analyses. 

3.9.4.2.1 Industrial Workers

Industrial workers involved in bulk rare earth product handling are considered to represent the

upper bound of likely exposures. Similar types of exposures are possible for maintenance

personnel. In this assessment, the bagging operations associated with bastnasite and cerium

concentrates are evaluated. The two primary exposure pathways associated with the handling

of bulk rare earth materials are dust inhalation and direct irradiation.

For the purposes of this assessment, the 232Th activity concentration of 3 Bq/g (85 pCi/g), as

shown in Table 3.9.13, is used for the bastnasite concentrate and the 232Th activity

concentration of about 7.7 Bq/g (213 pCi/g) is used for the cerium concentrate. These activity

concentrations correlate to the 0.1% and 0.25% weight as thorium phosphate found in the

bastnasite and cerium concentrate products, as compared to the higher activity concentrations

associated with the weight percent source material as thorium (1% by weight in bastnasite

concentrate 4 Bq/g (110 pCi/g)) and 0.25% by weight in cerium concentrate 10 Bq/g

(275 pCi/g), respectively. As of 1995, the Threshold Limit Value–Time Weighted Average

(TLV–TWA) for respirable particulate (particulate not otherwise classified) was changed from

5 mg/m3

 to 3 mg/m3, which is lower than the dust loading value of 5 mg/m3 measured by

Hewson and Terry (1995). Assuming that the respirable dust loading was 3 mg/m3

 and that

232Th is in equilibrium with its progeny, the estimated annual EDE (exposure duration of 2000

hours, no respiratory equipment) is about 10 mSv and 30 mSv (1000 mrem and 3000 mrem),

respectively. If a filter-type dust respirator with an assigned protection factor (APF) of 10 is

worn (National Institute for Occupational Safety and Health (NIOSH), 1990), the annual EDEs

would be reduced to about 1 and 3 mSv (100 mrem and 300 mrem), respectively.

As shown in Table 3.9.7, the average 168-hour external absorbed dose associated with

bastnasite and cerium bagging operations was about 0.03 mSv (3 mrem) and 0.07 mSv

(7 mrem), respectively (Bernhardt, 1996). Assuming a 12-month worker exposure, the annual

individual EDE from direct irradiation in the bastnasite and cerium bagging areas would be

about 0.4 mSv (40 mrem) and 0.8 mSv (80 mrem), respectively.

Therefore, if a worker handled bastnasite concentrate for approximately 2000 h/yr under the

given conditions, the annual EDE from dust inhalation (without using respiratory equipment)

and direct irradiation would be about 10 mSv (1 rem). If a worker handled a cerium concentrate

product for about 2000 h/yr under the given conditions, the annual EDE from dust inhalation

(without using respiratory equipment) and direct irradiation would be about 30 mSv (3 rem). It

is unknown how many U.S. workers are performing similar operations; however, if 100 workers

are conducting bastnasite and cerium packaging (and like activities), the annual collective

EDEs would be about 1 person-Sv and 3 person-Sv (100 person-rem and 300 person-rem),

respectively.
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3.9.4.2.2 Other Members of the Public

Some of the numerous products containing rare earths that are likely to be used by the public

include pink ophthalmic glass, television faceplates, special application automotive window

glass, and nickel metal hydride batteries. Certain types of ophthalmic glass lenses may contain

cerium, such as glass lenses that have a pink tint (Phone call, W. Price, Product Assurance,

Corning, Inc., Corning, NY, November 1996). This product is manufactured in Brazil and

currently is not used much in the United States. Since cerium is used in glass to attenuate

ultraviolet radiation and prevents discoloration, it is used in television faceplates and specialty

automotive glass. Nickel metal hydride batteries may contain lanthanum, cerium, neodymium,

and praseodymium. However, only high-purity rare earth elements, containing very low

concentrations of thorium, are used in the manufacturing of batteries (Phone call, L. Houston,

Chemist, EverReady Energizer, Inc., Gainesville, FL, November 1996). For the purposes of the

current assessment, exposures from the use of pink ophthalmic glass, television faceplates,

and automobile windshields containing cerium are evaluated. 

3.9.4.2.2.1 Ophthalmic Lenses

Regardless of the origin of the radioactive material, whether it is in the zirconium oxide or

cerium oxide, the ophthalmic glass industry has a radiation standard to which it adheres

voluntarily. The evaluations conducted by Buckley et al. (NUREG/CR–1775) and the NCRP

(NCRP 95), which are described in Section 3.2, incorporated the radiation limits of this standard

and are therefore considered to be applicable to this assessment. The individual annual dose

equivalent to the cornea from alpha particles was estimated to be 0.04 Sv (4 rem) and the

annual whole-body dose equivalent from photons was estimated to be 2×104

 mSv (0.02 mrem)

(NUREG/CR–1775; NCRP 95). In NCRP 95, a tissue weighting factor of 104

 was used to

derive the EDE from the alpha particle dose equivalent to the cornea. In this assessment, the

annual EDE due to ophthalmic glass eyewear is based only on the photon whole-body dose

equivalent. These doses assume 0.05% thorium in the glass, the estimated individual EDE for

this section would be five times greater for the maximum allowed thorium content of 0.25%. As

exact average content is not known, the maximum content allowed will be used for pink lenses. 

In Section 3.2, it was established that ten million people wore corrective eyewear that contained

glass. If it is assumed that 1% of the glass eyewear population used pink ophthalmic glass

(about 100,000 people), the hypothetical annual collective EDE due to photon irradiation is

0.1 person-Sv (10 person-rem).

3.9.4.2.2.2 Television Faceplates

Television glass faceplates contain about 0.2 to 0.3% cerium oxide (Phone call, D. Johnson,

Consumer Electronics Manufacturers Association, Electronics Industry Association (EIA),

Arlington, VA, December 1996). Television faceplates are subjected to bombardment by

high-energy electrons, particularly with the high tube voltages needed for color displays. This

bombardment over time tends to cause discoloration, or browning, of the glass due to the

creation of color centers. This unwanted effect is suppressed by the addition of cerium oxide to

the glass (Kilbourn, 1992). If 0.3% cerium oxide is used in television faceplates and there is

0.25% by weight of thorium in the cerium product, then the 232 Th activity concentration is about

0.03 Bq/g (0.8 pCi/g) in the television faceplate. 
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MicroShield (Computer Codes, Grove Engineering, 1996) was used to estimate the EDE rate to

television viewers from gamma radiation. Assuming the dimensions of a 68.6-cm (27-inch)

television faceplate are 53.34 cm × 43.2 cm × 0.3 cm and there is an average distance of

1.5 meters between the television faceplate and the viewer, the estimated EDE rate is

6×109

 mSv/h (6×107 mrem/h). If a television viewer spends about 3 h/day watching television

(EPA/600/P–95/002Fa), the annual EDE is calculated to be about less than 1×105

 mSv

(<0.001 mrem). In 1995, approximately 23 million color televisions were sold (EIA, 1996). If

250 million televisions were viewed for 3 hours per day at the given geometric dimensions, the

annual collective dose equivalent is estimated to be 2 person-Sv (200 person-rem).

It is unclear how much, if any, cerium oxide is present in computer monitor faceplates. It is

assumed that these faceplates contain 0.2 to 0.3% cerium oxide as an upper bound estimate. 

Further, if 13 million personal computers were sold in 1997 and if they are assumed to last for

5 years, taking into account built-in obsolescence, there would be about 65 million monitors in

use at any time. Taking into account the amount of time an average person might spend in

front of a computer at home or at work (e.g., 2000 hours) and the closer distance to a monitor

compared with a television set, about 20 inches, the hypothetical individual annual EDE would

be 3×105

 mSv (0.003 mrem). The maximum annual EDE for this product would be

proportional to the total time spent but it is unreasonable to estimate above twice this value. 

The hypothetical collective EDE would be 65 million monitors times the individual annual EDE

or 2 person-Sv (200 person-rem).

3.9.4.2.2.3 Automotive Window Glass

Not all automotive glass contains cerium. Cerium is used in specialty automotive glass

primarily to attenuate UV radiation and reduce solar heating (Phone call, M. Purvis, LibbeyOwens-Ford Company, Toledo, OH, November 1996). The market for this type of glass is

primarily in higher priced or luxury automobiles. It was widely used in 1988 and 1989, but the

use of cerium in automotive window glass has been reduced since then due to economic

considerations. It is possible that up to about 1.5% of the automotive glass composition may be

cerium (Phone call, P. Higby, Libbey-Owens-Ford Company, Toledo, OH, November 1996). If

there is 0.25% by weight of thorium in the cerium product and 1.5% of the glass composition is

cerium, then the 232 Th activity concentration is about 0.15 Bq/g (4 pCi/g) in the automotive

glass.

MicroShield (Computer Codes, Grove Engineering, 1996) was used to estimate the potential

EDE rate to motorists from gamma radiation from the glass. Assuming that the windshield

glass dimensions are about 157 cm × 83.8 cm × 0.64 cm and that the distance between the

driver and windshield is 0.6 meter, the EDE rate is calculated to be about 1.4×106

 mSv/h

(1.4×104

 mrem/h). If the side window glass dimensions were 83.8 cm × 61 cm × 0.64 cm and

the distance between the driver and the side window was 0.3 meter, the EDE rate is estimated

to be 3×106

 mSv/h (3×104 mrem/h). If an individual spends 80 minutes per day in an

automobile (EPA/600/P–95/002Fa), the annual EDE is estimated to be 0.002 mSv (0.2 mrem). 

It is unknown how many vehicles contain this type of glass. Assuming there are 800,000

vehicles with this type of window glass, the annual collective dose equivalent is estimated to be

approximately 1 person-Sv (100 person-rem). This dose is based on the maximum allowed

thorium content as no date is available regarding industry practices.
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A maximum annual EDE would be to a commercial driver and is 0.007 mSv (0.7 mrem) based

on 2000 h/yr driving, e.g., taxi driver. Assuming 200,000 vehicles of this type, the annual

collective dose equivalent is estimated also to be about 1 person-Sv (100 person-rem).

3.9.4.3 Disposal

To evaluate the disposal of a rare earth product containing 0.25% by weight of source material,

the generic disposal methodology, as described in Appendix A.2, is used to estimate the annual

individual and collective EDEs from landfill disposal. The quantity disposed and the possible

composition of these products when disposed is unknown. For the purposes of this

assessment, it is assumed that 900 MT (9×105

 kg) of cerium concentrate (0.25% by weight of

source material as thorium) is disposed annually in the United States. As cerium concentrate

contains about the maximum allowed concentration of 0.25% by weight, average and maximum

EDEs will be about the same value. Assuming there are about 3,500 sanitary landfills in the

United States, this equates to about 270 kg of product disposed annually in each landfill. 

Summarized in Table 3.9.14 are the annual individual and collective EDEs from disposal of

cerium concentrate in municipal landfills.

Cerium may be used in certain metal and alloy additives. During the production of additives,

radium appears to concentrate in the slag. As noted in Section 3.9.3.2, the measured dose

rates ranged from 1.7×104

 to 1.9×104 mSv/h (0.17 to 0.19 mrem/h) in a room where slag was

stored. Assuming that a slag worker is exposed 24 hours in a year (Hill, et al., see references

1995), the annual individual EDE is estimated to be 0.05 mSv (5 mrem). Background

radionuclide concentrations were indicated in off-site soil samples.

3.9.4.4 Accidents

In this evaluation, the warehouse and transport accident scenarios were selected to represent

upper bound doses associated with a product that contains 0.25% by weight of source material. 

It is assumed that 1000 bags of cerium concentrate (45-kg bags) are stored in a warehouse or

are being transported. If 0.25% by weight source of material is thorium, the 232Th activity

concentration is 10 Bq/g (275 pCi/g). It is difficult to imagine a fire causing a significant fraction

of cerium concentrate to become airborne; however, fire fighting activities could disperse the

material and clean up workers could encounter contamination. Using the generic accident

methodology, as described in Appendix A.1, the EDE to a firefighter combating a warehouse

fire is about 0.05 mSv (5 mrem) and the EDE to a firefighter combating a transportation fire is

about 0.3 mSv (30 mrem). The estimated EDEs to spill cleanup workers are 0.2 mSv

(20 mrem) for the warehouse fire and 2 mSv (200 mrem) for the transportation fire. No

respiratory equipment is assumed for the cleanup workers, unlike for the firefighters. However,

some type of respirator would be worn by cleanup personnel. If dust-type respirators with an

APF of 10 were used, the estimated EDEs for spill cleanup workers at a warehouse fire and

transportation fire would be 0.1 mSv (10 mrem) and 0.2 mSv (20 mrem), respectively. 

3.9.5 Summary 

Numerous products may contain rare earth materials. Depending on the rare earth material

used, negligible amounts up to 0.25% by weight of source material may be present. Selected

rare earth products and byproducts were used to represent the upper bound of doses that could
3–162

be attributed to rare earth products. The annual and collective EDEs estimated in the current

assessment are summarized in Table 3.9.15.

Based on this analysis, the following general conclusions about the radiological impacts on

industrial workers and other members of the public were obtained:

Upper bound doses to industrial workers involved in routine handling operations,

especially during bagging operations, may be significant. This is due primarily to the

dust inhalation pathway. However, based on current information, average doses to

other industrial workers handling bulk rare earth products are likely to be lower than

those estimated for the bagging operation (and like activities).

The primary exposure pathways for industrial workers appear to be dust inhalation and

direct irradiation. Radon emanation (222Rn and 220Rn) from rare earth products appear

to be low except during certain processing operations and in certain waste streams. 

There was very little known information associated with radon exposures.

It is unknown how many U.S. industrial workers (and exposure durations) may be

involved with the handling of bulk rare earth products that contain 0.25% by weight of

source material and whether respiratory protection is routinely used.

The annual individual EDEs to the public from the glass used in television faceplates

and automotive glass is primarily from direct irradiation.

Ophthalmic glass is subject to Optical Industry Association alpha particle emission rate

standards. These standards are applicable to alpha particle emission rates regardless

of source origin (e.g., zirconium oxide or cerium oxide). In addition, pink ophthalmic

glass is used very little in U.S. eyewear. 

Other observations made during completion of this radiological assessment include:

The primary rare earth producers and many users, such as the battery and glass

manufacturers, appear to be well aware and/or have taken steps to reduce the thorium

content of their products. Lower grade rare earth products, such as bastnasite or

cerium concentrates, are more likely to contain thorium, whereas higher grade products

contain small or trace amounts of thorium.

Additional information on certain areas is necessary to accurately assess the potential

radiological impacts associated with this exemption. These areas are (1) rare earth

product use and corresponding thorium content, (2) radionuclide airborne activity

concentrations both within the work place setting and at off-site receptor locations, and

(3) current disposal practices.
3–163

Table 3.9.1 Products Derived From Rare Earth Elements a

Element Product

Cerium Glass, metal, and gemstone polishing

Radiation shielding glass

Glass colorization and decolorization

Discolorization stabilizer

Photochromic glass opacifier

Ceramic pigment

Carbon arc lighting

Incandescent lamp mantle

Trichromatic fluorescent lighting phosphor (yellow green)

Pyrophoric alloys, including lighter flints

Cast iron alloys, aluminum alloys, magnesium alloys, superalloys

Automotive catalytic converter monolith

Heat stabilizer in alumina catalyst

Fluid cracking catalyst

Hydrolysis of phosphoric acid esters

Oxidation of naphthalene

Polymerization catalyst for olefins

Corrosion inhibitor

Getter in azide manufacture, getter in vacuum tubes

Developing agent for aniline black

Heavy metal soaps for lubricating greases and flatting agents

Waterproofing and mildew proofing 

Dying and printing textiles

Ceramic capacitors

Semiconductors

Cathodes

Desulfurizing catalysts

High-strength glass

Catalyst for ammonia production

Illuminant in photography

Radiation dosimeter

Diluent in plutonium nuclear fuels

Fluorinating agent

Oxidant for organic compounds

Antiknock fuel additive

Oxidizer in self-cleaning ovens

Pharmaceutical for sea sickness and morning sickness

Dysprosium Permanent magnet addition (NdFeB)

Nuclear reactor control (DyP)

Measurement of neutron flux

Atomic weight determination

High-temperature sensing phosphor
3–164

Table 3.9.1 Products Made From Rare Earth Elementsa

 (continued)

Element Product

Erbium Fiber-optic amplifier

Glass colorizing of crystal and glassware (light pink)

Additive for colorization and nonreflective coating for sunglasses

Infrared-absorbing glass

Colorization of simulant gemstones

Lasers

Burnable poison in nuclear reactor

Chemical catalyst for n-hexane

Permanent magnet additive

Europium Cathode-ray tube phosphor (reddish orange)

Trichromatic fluorescent phosphor (reddish orange and blue)

X-ray intensifying screen phosphor (violet)

Metal halide lamp phosphor (reddish orange)

High-and low-temperature sensing phosphors

Nuclear reactor neutron absorber

Nuclear reactor control rods

Ceramic sensor

Organic shift reagent in nuclear magnetic spectroscopy 

Dopant in laser crystals

Tagging agent in postage stamp glue

Gadolinium Trichromatic fluorescent lighting phosphor (yellow green)

X-ray intensifying screen phosphors (yellow green)

Nuclear control rods

Nuclear reprocessing

Neutron-absorbing paint

High refractive index paint

Synthetic garnet for computer memory (GGG)

Scavenger for oxygen in titanium production

Dielectric ceramics

Laser crystal

Cryogenic refrigeration

Pharmaceutical radionuclide in scanning devices to detect

 osteoporosis (153Gd)

Magneto-optical recording

Holmium Dopant in laser crystal

Getter in vacuum tubes

Refractories
3–165

Table 3.9.1 Products Made From Rare Earth Elements a (continued)

Element Product

Lanthanum Petroleum fluid cracking catalyst

Glass additive to increase refractive index and decrease

 dispersion (eyeglasses, cameras, binoculars, rifle scopes, 

 and telescope lenses)

Nickel hydride rechargeable batteries

Hydrogen storage alloys

Optical and laser glass

Solid oxide fuel cell cathode and interconnect

Trichromatic fluorescent lighting phosphor (green)

X-ray intensifying screen phosphor (violet, blue, yellow-green

 and blue green)

Low-temperature sensing phosphor

Phosphor lamp coating

Calcium lighting

Lighting electrode

Tungsten welding electrode

Cast iron alloys, steel alloys, superalloys

Aromatization of n-hexane

Fluoride determination

Conductive ceramic heating elements (chromites)

Dielectric ceramic capacitors (titanates)

Electro-optical ceramics

Dimerization of methane to ethylene

Antiseptic

Electronic thermo-emissive cathode (boride)

Insulation

Reducing agent to convert nitrogen oxides to nitrogen

Low-temperature superconductor

Cryogenic refrigerant using hydrogen (LaNi5)

Superconductors (LaAs, LaSb, LaP)

Lutetium High-temperature sensing phosphor

Nuclear technology

Neodymium Permanent magnets (NdFeB)

Glass colorization (violet)

Dopant in laser crystals and glass

Ceramic capacitors

Ceramic glaze (pinkish violet)

Magnesium alloys

Welder’s protective glass

Carbon arc lighting electrodes

Catalyst for polymerization of olefins

Fiber-optic temperature sensor
3–166

3.9.1 Products Made From Rare Earth Elements a (continued)

Element Product

Praseodymium Ceramic pigment (yellow)

Glass colorization (light green)

Refractory ceramic heating element

Laser crystal

Aluminum alloys

Magnesium alloys

Permanent magnets

Oxidative dimerization of methane to ethylene

Cryogenic refrigeration (PrNi5)

Promethium Fluorescent lighting starter

Coatings for self-luminous watch dials

Beta particle source for thickness gauges

Miniature atomic batteries

Semiconductor battery 

X-ray source (147Pm)

Samarium Permanent magnets (SmCo)

Dopant in laser crystals

Neutron absorber

Nuclear reactor control rod

Catalyst in dehydrogenation of ethyl alcohol

Catalyst in oxidative dimerization of methane to ethylene

Infrared absorbing glass

Scandium Metal halide lamps to improve color appearance and produce

 high output

Laser crystal (GSGG)

Aluminum alloys 

Dual-anode tube in X-ray spectrometers

Semiconductors

Tracer material

Collimators in neutron lenses

Terbium Trichromatic fluorescent lighting phosphor (yellow green)

X-ray intensifying screen phosphors (yellow green, violet, blue)

Magnetostrictive alloys for transducers (TbFeNi)

Dopant in laser crystals

Thulium X-ray source (170Tm)

Dopant in laser crystal

Ferrites 
3–167

Table 3.9.1 Products Made From Rare Earth Elements a (continued)

Element Product

Ytterbium Portable X-ray source for blood irradiation

Dialectic ceramics

Carbon roads for industrial lighting

Dopant in synthetic garnet

Dopant in laser crystal

Additive in special glass

Yttrium High-temperature superconductor (YBaCuO)

Incandescent lamp mantle

Phosphor for cathode-ray tubes (reddish orange)

Trichromatic fluorescent lighting phosphor (reddish orange)

Metal halide lamp phosphor (reddish orange)

High-temperature sensing phosphor

Oxygen sensor ceramic

Laser crystals (YAG, YLF)

Gemstone simulants (CS and YAG)

Microwave ferrite (YIG)

Cast iron alloys and superalloys

Magnesium castings

Dispersion strengthened nickel and titanium alloys

Aromatization of n-hexane

Prostheses

Stabilizer for silicon nitride and zirconia ceramics

Microwave filter

Acetylene lights

Deoxidizer for vanadium and other nonferrous metals

Solid oxide fuel cell electrolyte and anode

Refractory crucibles and coatings

High-temperature furnace electrodes

____________________________________________________________________________

a

 Hedrick, 1993.
3–168

Table 3.9.2 Distribution of Rare Earth Materials a

Distribution Percentage of Use

Catalysts in petroleum, chemical, and

 pollution control

34

Metallurgical uses as iron and steel additives

 and as alloys

27

Ceramics, glass polishing compounds, and glass

 additives

26

Miscellaneous uses in phosphors, electronics,

 permanent magnets, lighting, and research

13

a

 Hedrick, 1993.
3–169

Table 3.9.3 Cerium-Containing Materials and Their Applications a

Nature of Ce Content in Material Example Commercial Application

Major component of mixed-

 lanthanide composition

Rare earth chloride,

 mischmetal

FCC catalysts,b

Fe metallurgy

Minor component of mixed-

 lanthanide composition

Lanthanum concentrate,

 La-Ln chloride

FCC catalystsa

Dominant element in oxide-type

 composition

Cerium concentrate Glass polishing, glass

 decolorizing

Relatively pure compound (>90%) Oxide, nitrate, metal Auto-emission catalysts

Highly pure compound (>99%) Oxide, salts Luminescence, catalysts

a

 Kilbourn, 1992.

b

 Fluid Catalytic Cracking (FCC).
3–170

Table 3.9.4 Annual Exposures to American Potash and Chemical Corporation

Personnel Working With Rare Earth Products a

Operator Activity Year

Annual Dose-Direct Irradiation (Beta/Gamma)

(mrem)b

Rare earth hydrate operator 1960

1959

455

210

Fluoride dryer-grinder 1960

1959

660

205

Cerium helper 1960

1959

760

500

Grinder operator 1960

1959

590

225

Bastnasite operator 1960

1959

535

125

Cerium operator 1960

1959

930

600

Cerium sifter 1960

1959

135

 60

Rare earth fluoride operator 1960

1959

265

455

Rare earth hydrate dryer 1960

1959

580

160

Ion exchange operator 1960

1959

200

605

Ion exchange operator 1960

1959

660

210

Ion exchange operator 1960

1959

815

465

Ion exchange operator 1960

1959

1,155 

 665

Ion exchange operator 1960

1959

785

870

a

 McCoy, 1961.

b

 1 mrem = 0.01 mSv.
3–171

Table 3.9.5 Air Concentrations of Thorium in Rare Earth Processing Plant a

Rare Earth Material

Activities in Areas

Sampled

% Weight of

Thorium

Content % ML

Thorium Activity

Concentrationb

(Ci/mL)c

Glass polishes

 bastnasite, cerium

 oxides, other

 products

Polishing materials

 are prepared,

 calcined, sifted and

 screened, and

 packed

<0.05–0.2 14 7×1011

Cerium hydrate Drying in manually

 loaded and unloaded

 tray dryers and wet

 separation of cerium

 from rare earth

 mixture

0.18 7 4×1011

Rare earth oxide Closed room with

 continuous rotary

 calciners and where

 material is screened

 and packed 

0.22 24 1×1010

Purified rare earths Ion exchange <0.05% 11 6×1011

Rare earth sodium

 sulfate

Filter cake is dried in

 rotary dryer and

 handled in bulk

0.11 0 0

Rare earth chloride NAd <0.05 3.5 2×1011

Rare earth fluoride Preparation, drying,

 pulverizing, and

 packing

0.22 11.5 6×1011

a

 McCoy, 1961.

b

 Estimated air concentration from % ML (maximum permissible level) as described in

McCoy (1961). 

c

 1 Ci/mL = 0.037 Bq/mL.

d

 No applicable information.
3–172

Table 3.9.6 Airborne Radioactivity Levels Across All Mineral Sands Industry a

Reporting Year

Monazite

Production

(kT/yr)

Number of

Designated

Workers

Number of PAS

Measurementsc

AA Indexb

(Bq/m3)

d

 1986 10.9 266 485 0.85

 1987 11.2 287 1,045 0.60

 1988 10.6 301 1,509 0.51

 1989 12.0 331 2,048 0.49

 1990 8.5 287 1,458 0.27

 1991e 6.9 194 1,204 0.16

a

 Hewson, 1993.

b

 A representative industry AA index value for each site has been obtained by averaging the

alpha activity concentration across each dry plant work category (i.e., operational, maintenance,

and supervisory categories).

c

 The number of personal air sampling (PAS) measurements relates to the number of samples

used to derive the alpha activity (AA) index.

d

 1 pCi/m3 = 0.037 Bq/m3. e

 The AA index for 1991 also reflects significantly curtailed production capacity at most of the

mineral sands sites.
3–173

Table 3.9.7 1992 Molycorp Personnel TLD Results

Locationa Period of Evaluationa

Exposurea

(mrem/168-h)b

Estimated Average

Annual Effective Dose

Equivalent

(mrem)b

Roaster operator 2/1–2/29

 3/1–3/31

 4/1–4/30

 2

 1

 8

 

 44

Bastnasite packaging 2/1–2/29

 3/1–3/31

 4/1–4/30

 3

 0

 6

 

 36

 

Cerium dryer operator 2/1–2/29

 3/1–3/31

 4/1–4/30

 4

 4

 8

 

 64

Cerium packaging 2/1–2/29

 3/1–3/31

 4/1– 4/30

 --

 3

 11

 

 84 

 

Ce-96 operator 2/1–2/29

 3/1–3/31

 4/1–4/30

 4

 0

 10

 

 56

Ce-96 helper 2/1–2/29

 3/1–3/31

 4/1–4/30

 2

 6

 10

 

 72 

 

a

 Bernhardt, 1996.

b

 1 mrem = 0.01 mSv.
3–174

Table 3.9.8 Molycorp Plant Area TLD Results a

Location

Period of

Analyses

Absorbed

Dose Rate

(rad/h)

Absorbed Dose

for 40 Hours

(mrad)

Estimated

Annual Dose

Equivalentb

(mrem)

Cerium bagging 1/9–2/12/92

2/11–4/15/92

 40

 40

 2

 2

80

80

Solvent extraction area

 SX–1 raff cell

 SX–1 sump pump

 SX–1 feed tanks

 

 SX–1 raff cell

 SX–1 sump pump

 SX–1 feed tanks

1/9–2/12/92

2/11–4/15/92

550

100

420

570

150

510

22

 4

17

23

 6

21

1,100

200

840

1,140

300

510

Bastnasite bagging and

 warehouse 

1/9–2/12/92

2/11–4/15/92

--

 60

 --

 2

---

120

a

 Bernhardt, 1996. Absorbed dose rate and 40-hour absorbed dose data rounded when

applicable.

b

 Annual exposure duration is estimated to be 2,000 hours and assumes that 0.01 mGy

(1 mrad) is equal to 0.01 mSv (1 mrem).
3–175

Table 3.9.9 Environmental TLD Measurements Associated With Molycorp

Lead and Iron Filter Cake Project a

Estimated Monthly Dose Equivalent

Date

Administration

Office

(mrem)b

Guardhouse

Background

(mrem)b

Fenceline 

Near School

(mrem)b

10/95–1/96 7 8 8

1/96–4/96 19 24 lost

7/96–10/96 17 19 17

Average 15 17 13

Annual averagec 180 200 160

a

 Bernhardt, 1996. Estimated monthly dose equivalents are based on full-time occupancy and

include background radiation levels as illustrated by estimated monthly dose equivalent at the

guardhouse.

b

 1 mrem = 0.01 mSv.

c

 Estimated annual effective dose equivalent (mrem).
3–176

Table 3.9.10 Radium Concentrations in Slag, Metal Additives,

and Cerium Oxide Product a

Radium Concentration

(pCi/g)b

Sample 228Ra 224Ra 226Ra

Cerium oxide product 113 110 11

Metal additive (9–11% cerium) 21 38 1

Slag pile from production of metal additive 104 12 14

Metal additive from first rejected load 9 7 1

Slag from first rejected load 310 29 38

Metal additive from second rejected load 1 2

Slag from second rejected load 592 59 63

Off-site soil 0.6 0.6 0.8

a

 New York State, 1994.

b

 1 pCi/g = 0.037 Bq/g.
3–177

Table 3.9.11 Individual Effective Dose Equivalents From Storage or Disposal

of Rare Earth Byproducts (Monazite Based) a

Exposure Scenario

Annual Effective Dose Equivalent

(mrem)b

ON-SITE INDIVIDUAL

Direct gamma

Dust inhalation

Total

0.01

70

70

OFF-SITE INDIVIDUAL

Direct gamma

Dust inhalation

Well water

Food contaminated by well water

Food contaminated by dust deposition

Total

0.004

5×105

3

5×104

6×107

3

a

 EPA, RAE–9232/1–2.

b

 1 mrem = 0.01 mSv.
3–178

Table 3.9.12 Annual Collective Effective Dose Equivalent From Storage 

or Disposal of Rare Earth Byproducts (Monazite Based) a

Exposure Scenario

Annual Collective Effective 

Dose Equivalentb

(person-rem)c

Exposure to resuspended particulates

River water contaminated by groundwater

River water contaminated by surface runoff

Total

0.03

0.02

3

3

a

 EPA, RAE–9232/1–2.

b

 The collective effective dose equivalent is due to 1 year of intake.

c

 1 person-rem = 0.01 person-Sv.
3–179

Table 3.9.13 Average Concentration of Radionuclides in Selected Materials a

Material

232Th

(pCi/g)b

238U

(pCi/g)b

Annual Amount Produced or Disposedc

(MT)

Bastnasite ore 25 7 408,150

Bastnasite

 concentrate

 85 4 20,090

Stabilized lead

 iron filter cake

 32 1,150 NAc

Pond lead iron

 residue

 42 1,540 NAc

Lead sulfide

 concentrate

 5 74 590c

SX crud 15 760 113d

a

 ENSR, 1996.

b

 1 pCi/g = 0.037 Bq/g.

c

 Stabilized lead iron residue and pond lead residue are no longer produced as a result of

operations.

d

 Estimate based on the past 3 years of production of SX crud.
3–180

Table 3.9.14 Summary of Annual Effective Dose Equivalent Due to Landfill Disposal

of Cerium Concentrate (0.25% by Weight of Source Material as Thorium) a

Exposure

Scenario

Individual Annual Effective

Dose Equivalent

(mrem)c

Collective Effective 

Dose Equivalentb

(person-rem)c

WASTE COLLECTOR

Direct gamma

Dust inhalation

Soil ingestion

Total

0.5

0.1

0.06

0.7 2

ON-SITE WORKER

Direct gamma

Dust inhalation

Soil ingestion

Total

0.06

0.02

0.01

0.09 2

ON-SITE INDIVIDUALd

Direct gamma

Inhalatione

Soil ingestion

Total

<0.001

<0.001

<0.001

<0.001 420

OFF-SITE INDIVIDUAL

Dust inhalation <0.001 <0.001

Well waterd <0.001

a

 Refer to the generic disposal methodology in Appendix A.2.

b

 To estimate the annual collective effective dose equivalent (EDE), the following populations

were considered: For the waste collector, there is assumed to be 1 collector for each of 3,500

landfills. The landfill worker population is 17,500 (5 workers per landfill multiplied by 3,500

landfills); the on-site population is 34,000; the off-site population associated with atmospheric

releases is 1 million people within an 80-km radius of each landfill; and the off-site population

associated with ground water emissions is 700,000.

c

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

d

 For the on-site individual, the annual individual and collective EDE is estimated by assuming

the disposal of 270 kg of cerium concentrate per year per landfill, which is accumulated for

30 years until landfill closure and residents begin to live on-site 30 years after landfill closure.

For the off-site individual EDE it is assumed that 270 kg of cerium concentrate is disposed in

1 year per landfill and the travel time to the groundwater well 100 meters from the landfill

boundary is about 1.5×105

 years. e

 Annual individual and collective EDE for inhalation exposure pathway takes into account 220Rn

and its short-lived decay products during indoor residence.
3–181

Table 3.9.15 Summary of Doses to Industrial Workers and Other Members of the Public

From the Handling and Use of Rare Earth Products Containing Source Material

 

Exposure Scenario

Individual Annual Effective

Dose Equivalent

(mrem)a

Collective Effective

Dose Equivalent

(person-rem)a

DISTRIBUTION AND TRANSPORTb

Driver

 Cerium and bastnasite

 concentrates

Warehouse worker

 Bastnasite concentrate

 Cerium concentrate 

20

70

100

0.1

10

20

ROUTINE OPERATIONS

Industrial workers

 Bastnasite concentratec

 Cerium concentratec

Other members of the publicd

 Pink ophthalmic glass

 (photon irradiation)

 Television faceplates

 Personal computer

 faceplates

 Automobile glass

1000

3000

 0.1

 <0.001

 0.003

 0.006

100

300

10

200

200

6

DISPOSALe

Landfill

 Waste collector

 On-site worker 

 On-site individual 

 Off-site individual

 -Dust inhalation

 -Well water

Slag worker

0.7

0.09

<0.001

<0.001

<0.001

5

2

2

400

<0.001

 

NAg

ACCIDENTS

Warehouse fire

 Firefighter

 Cleanup personnelf

 

Transportation fire

 Firefighter

 Cleanup personnelf

5

2

30

20

NA

NA

See following page for footnotes.
3–182

Footnotes to Table 3.9.15

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 The truck driver is assumed to make 30 trips across the United States per year and there are

200 shipments annually. For the warehouse worker, it is assumed that each worker spends

500 h/yr in the warehouse, there are two workers per warehouse, and there are about 100

facilities in the United States storing bastnasite or cerium concentrates.

c

 For the industrial worker, the annual effective dose equivalent (EDE) assumes a respirable

dust concentration of 3 mg/m3

, no dust respiratory protection, and 2,000-hour exposure

duration. For the industrial worker, the annual collective EDEs assume a worker population of

100.

d

 The hypothetical populations used to estimate annual collective EDE for the public are as

follows: 100,000 wearers of pink ophthalmic glass, 250 million color televisions, and 1 million

drivers with solar-controlled glass (800,000 public and 200,000 commercial). 

e

 For the landfill operators, a worker population of 17,500 is assumed. For the off-site

population used to estimate annual collective EDEs there are 500,000 people who live within

80 km of each municipal landfill (atmospheric releases), and 700,000 people drink from off-site

wells. 

f

 Respiratory protection is also assumed for warehouse and transportation spill cleanup

personnel. An assigned protection factor of 10 (APF=10) is assumed.

g

 Not applicable.
3–183

3.10 Thorium in Personnel Neutron Dosimeters

3.10.1 Introduction

In 10 CFR 40.13(c)(1)(vii), any person who receives, possesses, uses, transfers, or acquires

personnel neutron dosimeters containing thorium is exempted from licensing requirements for

source material, provided each dosimeter does not contain more than 50 milligrams (mg) of

thorium. This exemption was proposed on June 24, 1976 (41 FR 26032), and issued as a final

rule on February 3, 1977 (42 FR 6579).

The first Federal Register notice cited above includes estimates obtained by the Nuclear

Regulatory Commission (NRC) of the radiation dose to individuals. A complete description of

this product as originally envisioned and a more detailed discussion of potential radiation doses

to individuals resulting from its use are given in the environmental impact statement for this

exemption (NUREG–0137).

In the 1980s, the personnel neutron dosimeters of choice used detectors made of CR39 (allyl

diglycol carbonate) plastic for fast neutrons and so-called neutron albedo dosimeters for

thermal and intermediate-energy neutrons (Griffith, 1988). The neutron albedo dosimeters use

thermoluminescent detectors made of nonradioactive materials such as lithium fluoride (LiF),

and the CR39 plastic detectors are damaged by neutron-recoil hydrogen ions (protons)

produced in radiator foils made of nonradioactive plastic materials (Luszik-Bhadra et al., 1990).

This product is not believed to be currently manufactured or in use. Past distribution, if any, is

not known.

3.10.2 Description of Items

The personnel neutron dosimeters containing thorium are simple in construction and have no

moving parts. A dosimeter consists basically of three components: (1) a radiator with a

thorium-containing foil or coating; (2) a detector that is a suitable medium for recording the

damage tracks due to the reaction products from the radiator; and (3) a case that encloses the

detector and radiator. The case is designed for attachment to the clothing of the dosimeter

wearer.

In the dosimeter design used as a basis for the assessment (NUREG–0137), the radiator

consisted of a 5-mg/cm2 -thick (0.0008-cm–thick) film of thorium fluoride (ThF4) coated, by

vapor deposition, onto an aluminum substrate. This substrate was a 0.2-cm-thick disk with a

radius of 1.8 cm, and each coating contained 50 mg of thorium that was assumed to have aged

20 years since purification. A single detector of polycarbonate plastic, about 5.0 cm in diameter

by 0.0001 cm thick, contacted the radiator in each dosimeter. Both the detector and radiator

were enclosed in an aluminum case with a wall thickness of 0.2 cm, and the case was designed

so that the portion covering a detector could be removed during detector replacement.

When fast neutrons with energies greater than approximately 1.5 MeV impinge on a radiator

of thorium, fission reactions occur that are characterized by the splitting of a thorium nucleus

into at least two very energetic nuclei (fission fragments) and by the release of a relatively

large amount of energy (about 200 MeV). These very energetic fission fragments will produce
3–184

damage tracks in the detector that can be counted after appropriate chemical etching of the

detectors (Tommasino and Harrison, 1985). The number of tracks in the detector can be

counted visually, by automatic optical scanning, or by electric spark counting techniques.

The personnel neutron dosimeters that contain thorium would be issued to and worn by

persons who work in or visit facilities where they may be exposed to fast neutrons

(NUREG–0137). These facilities might include nuclear power plants, research reactors, particle

accelerators, military installations, nuclear research laboratories, and other establishments

containing neutron-producing equipment or neutron-emitting materials.

3.10.3 Summary of Previous Analyses and Assessments

A variety of routine, accident, and misuse exposure scenarios have been considered previously

by the NRC (NUREG–0137). Scenarios for routine aspects of the life cycle for personnel

dosimeters include: 

Exposures to radiation workers and members of the public during distribution and

transport.

Exposures to radiation workers while wearing dosimeters or changing detectors used in

dosimeters.

Exposures to a wearer’s family or members of the public in the vicinity of a dosimeter

wearer.

Exposures to disposal workers at landfills.

The exposure scenarios involving accidents or misuse of personnel dosimeters that have been

considered include:

Release of thorium in a fire.

Release of thorium to water following accidental dumping of dosimeters into a reservoir

used to supply drinking water.

For exposure scenarios involving routine aspects of the life cycle of the neutron dosimeters, it

was assumed that ingestion or inhalation of thorium would not occur normally and external

irradiation would be the only significant exposure pathway. However, the previous assessment

used an older dosimetry system (International Commission on Radiation Protection (ICRP) 2)

and simply reported radiation doses to a number of critical organs, including the total body.” 

Because the radiation doses were from external irradiation involving high-energy photons, it is

possible to equate the total body” doses to effective dose equivalent (EDE) (ICRP 26). Thus,

the dose estimates for routine exposure scenarios obtained by the NRC (NUREG–0137), based

on a thorium content at the exemption limit of 50 mg, an annual distribution of 75,000

dosimeters, and a useful lifetime of 4 years for the dosimeters, are summarized as follows:
3–185

Distribution and transportation workers would receive annual EDEs in the range of

2×106 to 0.002 millisievert (mSv) (2×104 to 0.2 mrem), and the annual collective EDE

would be 2×105

 person-Sv (2×103 person-rem).

Radiation workers, including wearers of dosimeters and personnel changing detectors,

would receive annual EDEs of 2×106 to 0.013 mSv (2×104

 to 1.3 mrem), and the

annual collective EDE would be 2 person-Sv (200 person-rem).

Members of families of dosimeter wearers and other members of the public would

receive an annual EDE of 0.008 mSv (0.8 mrem) or less, and the annual collective EDE

would be 0.008 person-Sv (0.8 person-rem).

Disposal workers at a landfill, handling an estimated 75 dosimeters per year, would

receive annual EDEs of about 2×108

 mSv (2×106 mrem), and the annual collective

EDE would be about 2×106

 person-Sv (2×104 person-rem).

Dose estimates obtained by the NRC (NUREG–0137) for the accident and misuse scenarios

are summarized as follows:

If a fire involving a shipping carton of 200 dosimeters is assumed to discharge all of the

thorium uniformly in a moderately sized storage room, an individual breathing the

contaminated air for 1 hour could receive radiation doses of 5 mSv (0.5 rem) to the total

body, 200 mSv (20 rem) to the bone, and 3 mSv (3 rem) to the lungs. However, actual

doses for the postulated accident probably would be only small fractions of these values

because of such factors as ventilation of the room, accumulation of the dispersed

thorium near the storage carton for the dosimeters, rapid deposition of airborne thorium

onto the floor, and use of respiratory protection equipment by the individual entering the

room. More importantly, it is highly improbable that a fire would be sufficiently hot to

vaporize all of the thorium.

If a shipping carton containing 200 neutron dosimeters is dumped into a reservoir and all

of the thorium is dispersed uniformly in the water, the radiation doses to the individuals

drinking the water would be about 7×106

 mSv (7×104 mrem) to the total body and

6×105 mSv (0.006 mrem) to the bone. Because such an accident is highly unlikely, it is

eliminated from further consideration in this report.

3.10.4 Present Exemption Analysis

This hypothetical assessment is based on the previous NRC assumptions (NUREG–0137) of a

useful lifetime of 4 years for the dosimeters, based on normal wear and tear on the case of the

dosimeters, and an annual distribution of 75,000 dosimeters, with each dosimeter containing

the maximum allowed 50 mg of natural thorium. Thus, the total amount of natural thorium

distributed annually in the 75,000 dosimeters would be 3.75 kg. Collective doses were not

determined as this product is not believed to be currently manufactured or in use.
3–186

3.10.4.1 Distribution and Transport

Shipments would contain the number of dosimeters ordered by a customer, probably between

20 and 1000 dosimeters. Hence, the NRC (NUREG–0137) assumed a typical shipment might

contain 200 dosimeters, and 375 such shipments per year would be required for an assumed

annual distribution of 75,000 dosimeters.

For this assessment, it is assumed that the dosimeters are shipped primarily by a

parcel-delivery system, and a local parcel-delivery driver picks up the dosimeters and takes

them to a local terminal for shipment to customers. It is further assumed that semi-trucks are

used to transport the dosimeters between local terminals, and the dosimeters pass through an

average of four regional terminals before reaching their final destination. Radiation doses to

workers at both local and regional terminals are assumed to be the same as those estimated

for workers at a large warehouse (see Appendix A.3).

Based on the above assumptions and the generic methodology of Appendix A.3, the individual

receiving the largest radiation dose is the local driver who is assumed to pick up 1 or 2

shipments each day from the same supplier for a total of 375 shipments during a year. The

annual EDE to this individual is estimated to be 0.005 mSv (0.5 mrem). Individual doses to

other drivers, terminal workers, and members of the public along the truck routes would be less. 

The maximum EDE would be five times this value to accommodate the maximum number of

dosimeters in a shipment.

3.10.4.2 Routine Use

Persons touching or wearing a dosimeter could be exposed to both beta particles and photons

from decay of the natural thorium contained in the dosimeters (NUREG–0137). Beta particles

will penetrate the aluminum case (approximately 26% of the beta particles have energies

greater than 1.26 MeV), and the beta-particle dose rate was calculated to be 6.5×104

 mSv/h

(0.065 mrem/h) at the center of the front surface to the aluminum case of a dosimeter

(NUREG–0137). Photons from decay of natural thorium account for most of the external dose

from the dosimeters, and the photon dose rate at 1 meter from the dosimeter was calculated to

be 5.5×108

 mSv/h (5.5×106 mrem/h) (NUREG–0137). The photon attenuation was neglected

in this calculation. The natural thorium was assumed to be 20 years old so that the decay chain

was about 87% of equilibrium.

3.10.4.2.1 Dosimeter Wearers

The NRC (NUREG–0137) considered individual doses from wearing or touching dosimeters for

three different groups: (1) dosimeter wearers, (2) members of the wearer’s family, and (3) other

persons who might be near dosimeter wearers. Results of these previous NRC considerations

can be summarized as follows.

Wearers are assumed to carry two dosimeters attached to their clothing over the chest for

2,000 h/yr and to touch a dosimeter with their hands for 1 min/day (4 h/yr). Such a person,

could receive an annual dose equivalent to the hands of 0.004 mSv (0.4 mrem) from both beta

particles and photons. The annual EDE to dosimeter wearers from whole-body irradiation by

photons could be 0.009 mSv (0.9 mrem).
3–187

Members of the wearers’ families (four other members assumed) might be exposed for

12 h/day and 365 days/yr (4380 h/yr) to a dosimeter stored at an average distance of 6 meters

from each member. Each family member might also handle a dosimeter for a total of 0.7 h/yr,

and thereby receive an annual dose equivalent to the hands of 7×104

 mSv (0.07 mrem) from

both beta particles and photons. The annual EDE to a family member from whole-body

irradiation by photons could be less than 1×105

 mSv (<0.001mrem). 

Other persons who might be near dosimeter wearers (e.g., in restaurants, public buildings, and

transportation vehicles) could be exposed to the photons from the dosimeters. If such a person

should be exposed daily to a single dosimeter for 1 hour and if the dosimeter and person were

separated at an average of 3 meters, the person could receive an annual EDE from whole-body

irradiation by photons of less than 1×105

 mSv (<0.001mrem).

3.10.4.2.2 Detector Changers

The NRC (NUREG–0137) assumed that each customer assigned a trained employee (e.g., a

health physics technician) the task of replacing periodically the detectors in used dosimeters. 

The detector changers were assumed to handle 400 dosimeters per month for 5 seconds with

the case closed and 3 seconds with the case open. This monthly activity would require the use

of 375 detector changers to service the 150,000 dosimeter wearers. The changer also is

assumed to be exposed at 4.5 meters from an average of 200 dosimeters for 2000 h/yr. Such

exposures could result in an annual dose equivalent to the hands of 0.2 mSv (20 mrem) from

both beta particles and photons. The annual EDE to a detector changer from whole-body

irradiation by photons could be 8×104

 mSv (0.08 mrem).

Because opening of the dosimeter cases is assumed during detector replacement, any thorium

that might be detached from the radiators could be released to work areas and either ingested

or inhaled (NUREG–0137). The likelihood of such releases is considered small, because

(1) ThF4 is a relatively stable compound that should not oxidize or undergo destructive chemical

changes during normal conditions of use, (2) radiators should not be subjected to destructive

mechanical stresses in dosimeter cases, and (3) any significant detachment of ThF4 from a

radiator should be visible to and recognized by a trained worker who would initiate procedures

to prevent the spread of any detached thorium. Hence, potential releases of thorium during

detector replacement are considered unimportant if trained workers are used for detector

replacement, and the potential radiation dose from ingestion of detached thorium on a radiator

is shown to be negligibly small in Section 3.10.4.4 of this report.

3.10.4.3 Disposal

To estimate potential individual and collective doses to the public from disposal of neutron

dosimeters containing thorium in landfills and by incineration, the generic disposal methodology

in Appendix A.2 was used. The amount of natural thorium in 1 year’s distribution of 75,000

personnel neutron dosimeters is 3.75 kg, and it is in a form that is not readily dispersible in air

or readily accessible to infiltrating groundwater at landfills. Thus, reduction by a factor of

10 was assumed in the following dose-to-source ratios for inhalation and ingestion in

Appendix A.2: (1) waste collectors at both landfills and incinerators, (2) workers at landfills,

(3) off-site members of the public exposed to airborne releases during landfill operations and

releases to groundwater following disposal in landfills, and (4) future on-site residents at

landfills.
3–188

For disposal at landfills, the annual individual EDE would be about 1×105

 mSv (0.001 mrem) to

waste collectors. The annual individual doses to workers at landfills, off-site members of the

public, and future on-site residents would be substantially less. 

For disposal by incineration, the annual EDE would be 4×105

 mSv (0.004 mrem) to waste

collectors. The annual individual dose to workers at incinerators and off-site members of the

public would be substantially less. 

3.10.4.4 Accidents and Misuse

In the case of misuse, exposure to a person who finds and vandalizes a personnel neutron

dosimeter containing 50 mg of natural thorium is considered. It is assumed that the person

deliberately destroys the dosimeter out of curiosity and handles the thorium radiator foil in the

dosimeter for 30 minutes before discarding both the radiator foil and other dosimeter parts. For

the skin dose from beta particles from the radiator foil, a dose rate of about 0.054 mSv/h

(5.4 mrem/h) was calculated using VARSKIN MOD2 (Computer Codes, Durham, 1992). Thus,

the dose equivalent to the hands from touching the active surface of the thorium radiator for 

30 minutes could be 0.03 mSv (3 mrem). For potential ingestion of material due to touching of

the active surface of the thorium radiator, the generic accident methodology developed in

Appendix A.1 for spills of radioactive materials in the form of a powder was used. It was

assumed, first, that 10% of the loose material on the radiator is deposited on the body and,

second, that 0.1% of this deposited material is ingested before it is removed from the body by

washing. Thus, the estimated EDE from ingestion of material from the natural thorium radiator

could be 3×105

 mSv (0.003 mrem).

In the case of accidents, the following scenarios are considered: (1) a residential fire involving a

single dosimeter, and (2) a transportation fire involving an typical shipment of 200 dosimeters. 

A release fraction of 0.1% is assumed for the 50 mg of natural thorium contained in each

dosimeter. Based on these assumptions and the generic accident methodology in

Appendix A.1, the estimates of individual dose from fires involving dosimeters containing

thorium are summarized as follows:

For a person trying to escape from a residential fire or a neighbor trying to rescue a

person from a residential fire, the individual dose from a single dosimeter containing

50 mg of natural thorium could be 4×105

 mSv (0.004 mrem).

For a firefighter wearing a respirator at a residential fire, the individual dose from a

single dosimeter containing 50 mg of natural thorium could be less than 1×105

 mSv

(<0.001 mrem). The individual dose could be less than 1×105

 mSv (<0.001 mrem) to a

worker who is involved in the cleanup following the fire and who does not wear a

respirator.

For a firefighter wearing a respirator at a transportation fire, the individual dose from

200 dosimeters containing a total of 10 g of natural thorium could be 3×105

 mSv

(0.003 mrem). The individual dose could be 2×104

 mSv (0.02 mrem) to a worker who is

involved in the cleanup following the fire and who does not wear a respirator.
3–189

3.10.5 Summary

Table 3.10.1 presents the results of this hypothetical assessment of radiation doses to the

public from the distribution, routine use, and disposal of personnel neutron dosimeters

containing thorium. These results are based on an assumed 1 year’s distribution of 75,000

neutron dosimeters with an exemption limit of 50 mg of natural thorium each. For routine use of

the dosimeters, including distribution and disposal, the annual EDE to the most highly exposed

individual (wearer of a dosimeter) is estimated to be 0.009 mSv (0.9 mrem). For accidents and

misuse, the individual EDEs were estimated to be 2×104

 mSv (0.02 mrem) or less, and these

results suggest there is very little potential for significant exposure to members of the public

from accidents or misuse involving the thorium-containing foils in the personnel neutron

dosimeters.

An annual distribution of 75,000 personnel neutron dosimeters was assumed in the previous

assessment by the NRC (NUREG–0137) and also used as a basis for this assessment. No

demand has developed, and there is no known distribution or use.
3–190

Table 3.10.1 Potential Radiation Doses From Personnel Neutron Dosimeters

Containing Thorium

Exposure Pathway

Individual Annual

Effective Dose Equivalent

 (mrem)a

Distribution and transport 0.5b

Routine use

 Dosimeter wearers

 Detector changers

0.9c

0.08d

Disposal

 Landfills

 Incinerators

0.001e

0.004f

Accidents or misuse

 Fire

 Vandalism of dosimeter unit

0.02g

0.003h

a

 1 mrem = 0.01 mSv.

b

 Dose estimate applies to local parcel-delivery driver; dose estimates are considerably less for

terminal workers, long-haul semi-truck drivers, and members of the public along truck routes

(see Section 3.10.4.1).

c

 Dose estimate applies to dosimeter wearer; dose estimates are considerably less for

members of wearers’ family and other casually exposed members of the public (see

Section 3.10.4.2.1).

d

 Dose estimate applies to whole-body irradiation of detector changer; dose estimate for hands

of detector changer is 0.2 mrem/yr (20 rem/yr) (see Section 3.10.4.2.2).

e

 Dose estimate applies to waste collector at landfill; dose estimates are less for workers at

landfills, off-site members of the public, and future on-site residents at landfills (see

Section 3.10.4.3).

f

 Dose estimate applies to waste collector at incinerators; dose estimates are less for workers at

incinerators and off-site members of the public (see Section 3.10.4.3).

g

 Dose estimate applies to worker during cleanup following transportation fire; dose estimates

are less for firefighters and others involved in transportation or residential fires (see Section

3.10.4.4).

h

 Dose estimate applies to ingestion of natural thorium by a person who finds and destroys a

personnel neutron dosimeter containing a thorium radiator foil; dose estimate for beta-particle

dose to hands from touching active surface of the radiator foil is 0.03 mSv (3 mrem) (see

Section 3.10.4.4).
3–191

3.11 Glazed Ceramic Tableware

3.11.1 Introduction

In 10 CFR 40.13(c)(2)(i), glazed ceramic tableware is exempted from licensing requirements for

source material, provided that the glaze does not contain more than 20% by weight of source

material. A general exemption for ceramic products was included in the original 10 CFR 40

published on March 20, 1947 (12 FR 1855). The present exemption for glazed ceramic

tableware was proposed on September 7, 1960 (25 FR 8619), and issued as a final rule on

January 14, 1961 (26 FR 284).

Neither the original 1947 Federal Register notice nor the later notices from 1960 and 1961 cited

above provide information on radiological impacts on the public from use of glazed ceramic

tableware. The later notices stated that the exemption would not result in an unreasonable

hazard to life or property but do not present a supporting dose analysis.

Potential radiological impacts on members of the public from use of glazed ceramic tableware

are associated with beta-particle irradiation during handling tableware, beta-particle and

gamma-ray irradiation while near tableware, and ingestion of uranium leached into foodstuffs

that contacted tableware. Older data on and analyses of exposure and dose rates near glazed

ceramic objects and leaching of uranium from glazes by water and acidic solutions have been

summarized by Buckley et al. (NUREG/CR–1775) and the National Council on Radiation

Protection and Measurements (NCRP 95). These summaries and more recent data are

presented in Section 3.11.3.

3.11.2 Description of Items

An item of glazed ceramic tableware is a piece of ceramic formed into a plate, dish, bowl, cup,

or saucer and coated with a layer of glass-like material called glaze. The purposes of the glaze

are to seal the ceramic piece and to provide color to the item. Uranium oxide (U3O8) or sodium

uranate (Na2UO2) has been added to glazes to produce a variety of attractive colors (e.g., red,

orange, yellow, blue, and chocolate brown). Although such glazes may contain as much as

20% by weight of uranium, the highest percentage reported is 14%. (Ceramic glazes normally

contain uranium, not thorium.)

Glazes are applied to the surface of a fired ceramic piece by brushing on or dipping into

liquefied glazing material. The coated piece is then fired at a high temperature (e.g., 1430C)

to set the glaze. The resulting product is glass-like in appearance but is not as resistant as

glass to environmental stresses. The quantities of uranium-containing glazes applied to various

objects are not well documented, but Buckley et al. (NUREG/CR–1775) estimated a plate to

contain 4.5 g of uranium, and Piesch et al. (1986) reported a red-glaze thickness of 0.2 mm.

Uranium-containing glazes were applied to fine china and earthenware, items used in homes,

as overglazes, which are less resistant to wear and leaching than are the color-containing

underglazes used on commercial tableware. The use of uranium-containing glaze on porcelain

enamel and in pottery products was rare.
3–192

Domestic production of glazed ceramic tableware containing uranium apparently ceased about

20 years ago. The number of pieces of such tableware still in use is unknown. Buckley et al.

(NUREG/CR–1775) estimated production of at least 2 million pieces of dinnerware between

1959 and 1969, approximately 200,000 pieces per year. Although possibly still used for dining

purposes, pieces of such tableware are now considered collectible items (Landa and Councell,

1992) and likely would be displayed as such. The use of uranium in ceramic-tableware glazes

ceased primarily because of the availability of cheaper substitutes, such as mixtures of

antimony and iron, that yield acceptable glazes. Other factors that may have influenced the use

of uranium in glazes are: (1) publicity regarding the presence of radioactive materials in the

tableware probably affected sales adversely, and (2) as revealed by a general review of

commercial consumer-product literature, manufacturers are striving to remove from their

products all materials that are perceived to be hazardous. The above considerations, however,

do not preclude the possibility that uranium is present in glazes on imported products.

3.11.3 Summary of Previous Analyses and Assessments

Previous studies were concerned with measurement of exposure or dose rates near glazed

ceramic objects and determination of rates at which uranium leaches into various liquids. Dose

assessments largely were restricted to doses from handling tableware and from ingesting acidic

liquids that were in contact with the glazes.

3.11.3.1 External Exposures

Menczer (1965), using film badges, measured absorbed dose rates of between 5 and

200 microgray (Gy)/h (0.5 and 20 mrad/h) at the surfaces of various items of glazed ceramic

tableware due to gross beta and gamma radiation. He estimated that a person who handles

tableware for 1.5 h/day could receive a dose equivalent to the hands in the range of 0.02 to

0.1 sievert (Sv)/yr (2 to 10 rem/yr). The assumed handling time, 1.5 h/day, likely is an upper

limit value for most persons.

Simpson and Shuman (NUREG/CP–0001) reported exposure rates from a typical place

setting” of uranium-containing dishes to be 4 microcoulomb (C)/kg (15 milliroentgen (mR))

(130 Gy or 13 mrad) per hour at contact or 0.8 C/kg (3 mR) (30 Gy or 3 mrad) per hour at

2.5 cm. They estimated that a person seated before this place setting could receive 0.8 C/kg

(3 mR) (30 Gy or 3 mrad) per hour to the hands and arms and approximately 0.08 C/kg

(0.3 mR) (3 Gy or 0.3 mrad) per hour to the torso. They also report measured exposure rates

of between 1 and 2 C/kg (5 and 7 mR) (40 and 60 Gy or 4 and 6 mrad) per hour at contact

with 0.2 C/kg (0.7 mR) (6 Gy or 0.6 mrad) per hour at 25 cm from Fiestaware®.

Simpson and Shuman (NUREG/CP–0001) also used a beta-particle point-source formulation to

calculate an absorbed dose rate of 0.024 Gy/yr (2.4 rad/yr) from a 45.6 day/yr (3 h/day)

exposure at 7.6 cm from a glazed ceramic surface in which all uranium daughters are in

radioactive equilibrium. This estimate likely is high by a factor of about 2 because all uranium

daughters would not achieve radioactive equilibrium during the useful life of a piece of

tableware. Also, a 3 h/day exposure at 7.6 cm is unrealistically conservative.

Buckley et al. (NUREG/CR–1775) cite gamma-ray exposure-rate measurements taken near

various ceramic products by the Bureau of Radiological Health. For plates, exposure rates
3–193

were between 0.1 and 4 C/kg (0.5 and 15 mR) (4 and 130 Gy or 0.4 and 13 mrad) per hour

at the surface and between 0.5 and 77 nC/kg (0.002 and 0.3 mR) (or 0.02 and 3 Gy or

0.002 and 0.3 mrad) per hour at 1 meter.

Piesch et al. (1986) measured a dose rate of 0.32 mSv (32 mrem) per hour at the surface of a

teacup covered with a red glaze. Assuming daily use of the teacup (an actual exposure time

was not given), they estimated an individual could receive 4 mSv (400 mrem) to the lips and

12 mSv (1200 mrem) to the fingers.

McCormick (1992) counted a variety of Fiestaware® items (plates, bowls, and pitchers) in an ion

chamber. Reported exposure rates ranged between 0.5 and 3 C/kg (2 and 10 mR) (20 and

90 Gy or 2 and 9 mrad) per hour.

3.11.3.2 Internal Exposures

Kendig and Schmidt (1972) contacted red-glazed plates with 4% acetic acid solutions for time

periods between 24 and 60 hours. Uranium contents in the solutions were 8.6, 1.8, 8.2, and

8.6 ppm (or mg/L) after 24 hours of contact for two bowls, a saucer, and a plate, respectively. 

After 60 hours of contact, uranium contents in solution were 55 and 41 ppm for a saucer and

plate, respectively.

Landa and Councell (1992) contacted uranium-bearing glazes on two objects (a plate and a

gravy boat) with 20 mL of three different solutions: (1) water, (2) 4% acetic acid, and

(3) 1 molar nitric acid. After 24 hours of contact, uranium concentrations in the solutions

ranged from 3.9 to 10.6 g/L in water; 470 to 31,800 g/L in acetic acid; and 96,100 to

304,000 g/L in nitric acid.

Landa and Councell (1992) noted the possibility of a memory effect on the leaching process. 

(The amount of uranium leached from an item per period of contact will decrease as the

number of contact periods increases.) Such an effect would be expected as the amount of

uranium available for leaching decreases after each leaching operation. For example, 20 mL of

4% acetic acid at 31,800 g/L contains approximately 640 g of uranium. If the uranium

content in 20 mL of acetic acid were removed from an item each day for 1 year, approximately

0.23 g of uranium would be removed from the item. If the uranium content of 0.5 L of acetic

acid (e.g., 8.5 fluid ounces of orange juice) were removed from an item each day for 1 year,

approximately 5.8 g of uranium would be removed. This exceeds or is a substantial portion of

the total uranium content of many pieces of tableware (see Section 3.11.2 and Table 3.11.1).

The above studies report neither the uranium concentrations in, nor the surface areas of, the

glazes that were in contact with the leaching solutions. Normalizing the above results with

respect to uranium concentration and contacted surface area could yield a better measurement

for comparing leach rates from a variety of objects (i.e., micrograms of uranium per liter of

leachate, per hour of contact, per square centimeter of glaze). 
3–194

3.11.4 Present Exemption Analysis

3.11.4.1 General Information

Most pieces of glazed ceramic tableware are distributed in sets of four, six, or eight place

settings. Other pieces that may be available separately include specialty items such as a

platter, a gravy dish, and serving bowls. Individual pieces may be distributed as replacement

items; however, the potential radiation doses associated with distribution of individual pieces

are a small subset of the doses associated with distribution of place settings. Therefore,

distribution of individual pieces is not evaluated in this section.

As noted in Section 3.11.2, the availability of tableware coated with uranium-bearing glazes

cannot be quantified because domestic manufacture of such products ceased at least 20 years

ago and because import of such tableware, if it occurs at all, could not be verified. The present

analysis conservatively assumes an annual distribution similar to that reported in Buckley et al.

(NUREG/CR–1775), 216,000 pieces of tableware. This tableware is assumed to be divided into

6000 sets, each of which contains 36 pieces in six place settings. Each place setting consists

of a 25-cm- (10-inch-) diameter plate, a 20-cm- (8-inch-) diameter dish, a 25-cm- (10-inch-)

diameter bowl, a 15-cm- (6-inch-) diameter bowl, a cup, and a 10-cm- (4-inch-) diameter

saucer. Analyses in this section include: (1) individual and collective doses due to distribution

of the tableware; (2) individual doses due to home use of a set of tableware; (3) individual and

collective doses due to display of a place setting as a collectible; (4) individual and collective

effective dose equivalents (EDEs) due to disposal; and (5) individual doses due to an

unintended use of a plate as a radiography source.

Measurements of several pieces of tableware indicate that representative thicknesses of

ceramic are between 0.64 and 1.3 cm for plates, dishes, saucers, and bottoms of bowls and

cups and approximately 0.64 cm for the walls of cups and bowls. Piesch et al. (1986) report a

glaze thickness of about 0.02 cm. A glaze thickness of 0.025 cm is used in the following

analyses.

Geometries and materials used to represent the various pieces and arrays of tableware used in

this assessment are listed in Table 3.11.1. Each item is represented by two cylinders of glaze

separated by 0.64 cm of ceramic. The thicknesses of the glaze cylinders were obtained by

dividing the total volume of glaze on an item by twice the top surface area of the cylinder. 

Arrays of tableware are represented by cylinders of a composite material, which is a

homogeneous mixture of glaze and ceramic. Array representations were derived in a way that

kept the exposed (projected) surface area constant while preserving the total mass of ceramic,

glaze, and uranium in the items that make up the arrays (see Table 3.11.1).

The glaze is modeled as a material having the composition of Pyrex® glass plus 20% by weight

of uranium, a density of 2.7 g/cm3

, and an atomic number of 11.1. The ceramic is represented

by a material with the composition of concrete and a density of 1.2 g/cm3

. The composite

material is modeled as a material having a density of 1.3 g/cm3

 and an atomic number of 9.7.
3–195

3.11.4.1.1 External Exposures

External exposures to glazed ceramic tableware involve exposures to both beta particles and

gamma rays. Beta-particle exposure is especially important during handling of tableware.

Absorbed dose rates in tissue due to beta-particle irradiation were calculated using VARSKIN

MOD 2 (Computer Codes, Durham, 1992). Dose rates were calculated at a depth of 7 mg/cm2

in tissue from contact with and from exposures at selected distances from a 25-cm-diameter

glazed ceramic disk. Dose rates, which vary between 240 Gy (24 mrad) (0.24 mSv or

24 mrem) per hour at contact and 0 Gy (0 mrad) per hour at 760 cm, are listed in

Table 3.11.2. Use of the 25-cm disk as a surrogate for all tableware geometries maximizes the

calculated doses, except doses from contact with the disk. (Dose rates from a 25-cm diameter

disk are between two and three times higher than dose rates from a 15-cm diameter disk.)

The contribution of beta-particle irradiation to the EDE rate at a selected distance from a piece

of glazed tableware is obtained by multiplying the calculated dose rate by the skin weighting

factor (0.01) and the fraction of skin under irradiation. For handling a plate, the dose rate to the

palms and fingers is 0.24 mSv/h (24 mrem/h). The skin of the palms and fingers constitutes

about 1% of the total skin area of the body. Therefore, the contribution of beta-particle

irradiation to the EDE rate is 2.4×105

 mSv/h (0.0024 mrem/h). For exposures at less than

90 cm, one-fourth of the total skin area is assumed to be irradiated; for the remaining exposure

distances, one-half of the total skin area is assumed to be irradiated. Resulting dose rates are

presented in Table 3.11.2.

Dose rates due to gamma and bremsstrahlung irradiation were calculated using CONDOS II

(Computer Codes, O’Donnell et al., 1981). Tables 3.11.3 and 3.11.4 are lists of calculated EDE

rates at selected distances from the items described in Table 3.11.1.

3.11.4.1.2 Internal Exposures

Reported uranium concentrations for various contact durations (see Section 3.11.3.2) in water

are 3.9 and 10.6 g/L/24-h of contact. Reported concentrations in 4% acetic acid are 470,

10,000, and 31,800 g/L/24-h; 55,000 g/L/50-h; and 105,000 g/L/60-h. Normalizing these

reported concentrations with respect to their reported leach durations yields effective

concentration rates of 0.2 and 0.4 g/L/h in water and 20, 420, 1325, 1100 and 1750 g/L/h in

4% acetic acid.

In the following assessment, maximum 24-hour contact concentrations, 10.6 g/L in water and

31,800 g/L in 4% acetic acid solutions, are used for liquids that contact glaze for 24 hours or

more (e.g., liquids not consumed in one sitting). For liquids in contact with glaze for less than

24 hours, the normalized concentrations corresponding to the above are used, 0.4 g/L/h in

water and 1325 g/L/h in acetic acid.

Based on EPA (EPA/600/P–95/002Fa) studies, a typical daily intake of liquids is approximately

2 L. This intake may include 1.4 L of water and water-based beverages, 0.11 L of fruit juices,

0.046 L of soups, and 0.55 L of food-derived liquids. These liquids are designated as water-like

and acid-like as follows: water and water-based beverages, 100% (1.4 L) water-like; fruit juices,

100% (0.11 L) acid-like; soups, 75% (0.035 L) water-like and 25% (0.012 L) acid-like; and

food-derived liquids, 75% (0.41 L) water-like and 25% (0.14 L) acid-like. Thus a typical
11 (54 L of 30-min acidic × 1325 g/L/h × 0.5 h) + (260 L of 30-min water-like × 0.4 g/L/h ×

0.5 h) + (5.5 L of 24-h acidic × 31,800 g/L) + (29 L of 24-h water-like × 11 g/L) = 2.1×105 g.

3–196

individual could ingest approximately 1.8 L/day (670 L/yr) of water-like liquids and 0.26 L/day

(95 L/yr) of acid-like liquids.

The above liquids normally are consumed out of glasses, not out of ceramic tableware, and

most contact times are much less than 24 hours. However, the following assumptions were

used to bound the analysis: 

Water and water-based liquids are 100% (1.4 L/day) water-like; 75% (1.1 L/day) are

consumed from glassware and 25% (0.35 L/day) from ceramic tableware; and 90% of

the liquids are in contact with the container for less than 30 minutes and 10% for

24 hours or more. Thus, 0.79 L/day are in contact with glassware and 0.315 L/day are

in contact with ceramic ware for 30 minutes, and 0.035 L/day are in contact with ceramic

ware for 24 hours.

Fruit juices are 100% (0.11 L/day) acid-like; 90% (0.099 L/day) are consumed from

glassware and 10% (0.011 L/day) from ceramic tableware; and 100% of the liquids are

in contact with the container for less than 30 minutes. Thus, 0.099 L/day are in contact

with glassware and 0.011 L/day are in contact with ceramic ware for 30 minutes.

Soups are 75% (0.035 L/day) water-like and 25% (0.012 L/day) acid-like; 100% are

consumed from ceramic tableware; and 90% are in contact with the container for

30 minutes and 10% for 24 hours or more. Thus, 0.032 L/day of water-like and

0.0109 L/day of acid-like soups are in contact with ceramic ware for 30 minutes, and

0.0035 L/day of water-like and 0.0012 L/day of acid-like soups are in contact with

ceramic ware for 24 hours.

Food-derived liquids are 75% (0.41 L/day) water-like and 25% (0.14 L/day) acid-like;

100% are consumed from ceramic tableware; and 90% are in contact with the container

for 30 minutes and 10% for 24 hours or more. Thus, 0.37 L/day of water-like and

0.126 L/day of acid-like food-derived liquids are in contact with ceramic ware for

30 minutes, and 0.041 L/day of water-like and 0.014 L/day of acid-like food-derived

liquids are in contact with ceramic ware for 24 hours.

Thus, considering only liquids consumed from ceramic ware, an individual may consume about

0.72 L/day (260 L/yr) of water-like and 0.15 L/day (54 L/yr) of acid-like liquids that were in

contact with glaze for 30 minutes plus 0.08 L/day (29 L/yr) of water-like and 0.015 L/day

(5.5 L/yr) of acid-like liquids that were in contact with glaze for 24 hours or more. Based on

these consumption rates and the leaching factors discussed above, an individual could ingest

approximately 0.21 g of uranium during 1 year.11 Thus, given an ingestion dose factor of

1.9×106

 mSv/g (1.9×104 mrem/g), this individual could receive an annual EDE of about

0.4 mSv (40 mrem) from ingestion of uranium leached from glazed ceramic tableware.

The derivation of the leaching factors was made without correlation to the uranium

concentration of the glaze. However, because of other assumptions regarding contact time and

usage, the results are considered conservative for a maximum exposed individual. Unless
3–197

glazed ceramic tableware is used as primary dinnerware, any actual dose would be

substantially less.

3.11.4.2 Distribution

As noted above, domestic production of tableware coated with uranium-containing glazes has

ceased and import of such tableware, if occurring, cannot be quantified. However, based on

past production, this assessment assumes an annual distribution of 216,000 pieces of

tableware in 6,000 cartons, each of which contains 36 pieces (six 6-piece place settings) of

tableware. Each of 2 manufacturers is assumed to ship 3000 sets of tableware (cartons), of

which 1000 are distributed through 10 wholesalers, 1000 through 20 large retailers, and 1000

through 50 small retailers.

The following model is used in this assessment. (See Table 3.11.5 for a list of the model steps,

the number of facilities involved in each step, and the number of cartons handled at each

facility.) During the year, all cartons (3000) from each of two manufacturers are loaded into a

small express-delivery truck and transported to a parcel-delivery center. At each center, 

(1) 100 cartons are loaded into a large local-delivery truck for transport to a wholesaler;

(2) 50 cartons are loaded into each of two large local-delivery trucks for transport to two large

retail stores; (3) 20 cartons are loaded into each of five small express-delivery trucks for

transport to five small retail stores; and (4) 900 cartons are loaded into each of three large

regional-delivery trucks for transport to three intermediate truck terminals. At each intermediate

truck terminal, (1) 100 cartons are loaded into one large local-delivery trucks for transport to a

wholesaler; 

(2) 50 cartons are loaded into each of two large local-delivery trucks for transport to two large

retail stores; (3) 20 cartons are loaded into each of five small express-delivery trucks for

transport to five small retail stores; and (4) 300 cartons are loaded into each of two large

regional-delivery trucks for transport to two final regional truck terminals. At each final truck

terminal, (1) 100 cartons are loaded into one large local-delivery truck for transport to a

wholesaler; (2) 50 cartons are loaded into each of two large local-delivery trucks for transport to

two large retail stores; and (3) 20 cartons are loaded into each of five small express-delivery

trucks for transport to five small retail stores.

The carton source representation given in Table 3.11.1 was used in conjunction with the

exposure conditions and calculation methods given in Appendix A.3.3 to calculate individual and

collective EDEs for each step in the model. Results of the calculations are presented in

Table 3.11.5. The highest calculated individual annual EDE was approximately 0.08 mSv

(8 mrem), to a worker in the initial parcel warehouse. The total collective EDE for distribution

was about 0.5 person-Sv (50 person-rem), almost entirely due to exposures at retail

establishments. Because of the number of cartons carried in each truck, the average truck

driver exposure conditions were used in the calculations (see Appendix A.3.3.). These dose

estimates are based on calculations using the maximum allowed concentration of 20% by

weight. Using an average 14% by weight, the workers dose would be 0.06 mSv (6 mrem) and

the collective EDE would be 0.3 person-Sv (30 person-rem).
3–198

3.11.4.3 Routine Use

3.11.4.3.1 Use as Dinnerware

Members of a household can be exposed to the uranium contained in glazed ceramic tableware

both externally and internally. External exposures can occur while dining; moving tableware

before and after dining, washing and storing tableware, and merely being in the home,

especially in the kitchen where arrays of tableware are stored in cabinets. Internal exposure

can occur via ingestion of foodstuffs in contact with the uranium-containing glaze. The

following model, which is described partially in Table 3.11.6, the dose rates given in

Tables 3.11.2 through 3.11.4, and the uranium concentration rates given in Section 3.11.4.1.2

are used to estimate EDEs and dose equivalents to skin. External exposure durations were

based on studies of homemakers’ activities (Steidl and Bratton, 1968; EPA/600/P–95/002Fa);

exposure distances are based on observation. Liquid intakes are derived from the Exposure

Factors Handbook (EPA/600/P–95/002Fa).

A person designated as a homemaker in Table 3.11.6 is assumed to wash all dishes and to

spend the most time at home and in the kitchen. These activities could result in an EDE of

about 0.07 mSv/yr (7 mrem/yr). Other family members could receive EDEs similar to the home

maker. From Table 3.11.6, these persons could receive an EDE of about 0.06 mSv/yr

(6 mrem/yr) from external exposures. However, the doses from external exposures are small

with respect to the 0.4 mSv/yr (40 mrem/yr), that could be received from intakes of liquids that

were in contact with uranium-bearing glazes (see Section 3.11.4.1.2). 

3.11.4.3.2 On Display as Collectibles

Collectibles may be displayed in many ways, both in homes and in public facilities, such as

museums and stores that sell collectibles. Display in a public facility is the largest contributor to

collective dose.

Due to space limitations in a home display cabinet, one six-piece place setting is assumed to be

on permanent display in a glass-enclosed cabinet that eliminates the beta dose component

except during handling of tableware. It is assumed that an individual (1) handles individual

pieces (25-cm-plate is assumed) from a distance of 30 cm for 6.1 h/yr, (2) views the display

from a distance of 91 cm for an additional 10 h/yr, and (3) is in other rooms at average

distances of 610 and 762 cm for 200 and 6,000 h/yr, respectively. This individual could receive

an EDE of 6×104

 mSv/yr (0.06 mrem/yr). This person also could receive an annual dose

equivalent to the skin of the hands of about 1.4 mSv (140 mrem). Collective EDEs to a family

of four, with only one individual handling the items, could be 1.7×106

 person-Sv/yr

(1.7×104

 person-rem/yr) and 3.4×105 person-Sv (3.4×103 person-rem) over 20 years. If a

place setting from all 6000 sets of tableware is put on display in 6000 homes, the 20-year

collective EDE could be 0.2 person-Sv (20 person-rem).

An open display of six place settings in a public facility (e.g., museum, shop, etc.) is assumed to

be viewed by 500 persons/day (0.18 million/yr). Each person is assumed to spend 5 minutes

viewing the display from 91 cm away, 30 minutes at an average distance of 460 cm from the

display, and 1 hour at an average distance of 760 cm from the display. Each person could

receive an EDE of less than 1×105

 mSv (<0.001 mrem). For purposes of estimating collective

doses, the calculated value of 1.8×106

 mSv (1.8×104 mrem) based a 14% by weight glaze has
3–199

been used rather than the less than value. Collective EDEs per display could be

3×104 person-Sv/yr (0.03 person-rem/yr), 0.006 person-Sv (0.6 person-rem) over 20 years. If

6,000 such displays were available, the 20-year collective EDE could be 40 person-Sv

(4,000 person-rem).

3.11.4.4 Accidents and Misuse

3.11.4.4.1 Accidents

It is inconceivable that an accident involving ceramic tableware could have radiological

consequences greater than those from routine distribution and use. Even in the event of a fire,

a uranium-bearing glaze is unlikely to become airborne, because temperatures used to set

glazes, approximately 1430C, exceed temperatures normally achieved during a vehicular or

structure fire. Thus, any radiological consequences of an accident involving glazed ceramic

tableware would be due to handling the residue of the accident. Such handling is unlikely to

result in an EDE greater than the 0.08 mSv (8 mrem) that could be received by a parceldelivery terminal worker who handles all 6000 cartons in 1 year.

3.11.4.4.2 Misuse

Some pieces of glazed ceramic tableware have been used in a manner unintended by the

exemption—as crude radiography sources. In such instances, a person could handle a plate

for 1 minute (hands-on contact and body at 30 cm), be an average of 91 cm from the plate for

5 minutes while placing and removing the object to be radiographed on the plate and covering

and uncovering both plate and object with undeveloped film, and be 610 cm from the plate for

8 hours. Such exposures could result in reception of about 0.004 mSv (0.4 mrem) to the skin of

the hand and an EDE of less than 1×105

 mSv (<0.001 mrem).

3.11.4.5 Disposal

Under normal circumstances, glazed ceramic tableware would be disposed of as ordinary,

non-combustible household trash. The following assessment assumes discard of 1.8 Mg of

uranium in 6000 sets of tableware in 1 year based on the average uranium mass per set

presented in Table 3.11.1. This assumption omits consideration of random discard of individual

pieces of tableware prior to the year of discard. Individual doses will be maximized if all pieces

are disposed of at once, and collective doses will be affected little, if at all.

Using the assumptions of the generic disposal methodology (see Appendix A.2) for disposal of

2.5 Mg of natural uranium with 20% being incinerated, the highest calculated individual EDE is

0.004 mSv (0.4 mrem), to a waste collector at a municipal incinerator. No other individual is

estimated to receive an EDE greater than 0.001 mSv (0.1 mrem). The total collective EDE to

all workers and potentially exposed members of the public could be about 0.08 person-Sv

(8 person-rem).

3.11.5 Summary

This assessment has considered potential doses to the public from use of uranium as a

coloring agent in glazes applied to ceramic tableware. All direct exposure calculations were
3–200

based on the maximum concentration (20% by weight) of uranium allowed in the glazes and an

assumption that the tableware is still manufactured. However, there is no evidence that this is

occurring. The highest concentration reported in actual pieces of tableware is 14% by weight;

the collective dose for dinnerware is based on weight percent. Internal doses were estimated

based on the measured concentration of uranium in various liquids simulating different type

foods. Results of the assessment are summarized in Table 3.11.7.

The highest potential EDEs, approximately 0.04 mSv/yr (40 mrem/yr), are to persons who might

drink liquids that were in contact with and had leached uranium from the glazes. Although such

leaching does occur, the amount of leaching assumed in this assessment likely is excessive for

a 20-year useful life. No measurements were made of the uranium content of the glaze

(relative to the limit of 20% by weight). However, because of other assumptions regarding

usage, the estimated doses are considered conservative. Unless glazed ceramic tableware is

used as primary dinnerware, any actual dose would be substantially less.

Calculated doses due to distant external exposures to beta particles also are overstated due to

the exclusive use of a 25-cm-diameter disk to represent all tableware geometries. Using

smaller source diameters could reduce external dose estimates due to dining by a factor of 3 or

more. 

Since currently domestically manufactured as decorative or ornamental objects only, the most

likely exposure to an individual would be external from display. Therefore, the average

individual annual EDE is estimated to be 6×104

 mSv (0.06 mrem) for the 20% by weight limit

and 4×104

 mSv (0.04 mrem) at an average 14% by weight.
3–201

Table 3.11.1 Source Representations of Ceramic Tableware

 Representation

Cylinder Separator

Item

Radius

(cm)

Depth

(cm) Material

Uranium

Massa

(g)

Thickness

(cm) Material

10” plate 13 0.026 Glaze 14 0.64 Ceramic

8” dish 10 0.027 Glaze 9.3 0.64 Ceramic

4” saucer 5.1 0.028 Glaze 2.5 0.64 Ceramic

3.5” cup 4.4 0.11 Glaze 7.7 0.64 Ceramic

10” bowl 13 0.046 Glaze 25 0.64 Ceramic

6” bowl 7.6 0.059 Glaze 12 0.64 Ceramic

6, 10” plates 5.8 20 Composite 87 0 None

6, 8”dishes 5.2 16 Composite 56 0 None

6, 4” saucers 3.7 7.9 Composite 15 0 None

2 cups 5.0 4.7 Composite 15 0 None

6, 10” bowls 8.3 6.6 Composite 150 0 None

6, 6” bowls 6.4 13 Composite 70 0 None

Carton 12 23 Composite 424 0 None

a

 Based on 20% by weight glaze. Multiply by 0.7 for 14% by weight average.
3–202

Table 3.11.2 Absorbed Beta Dose Rates at 7 mg/cm2 in Tissue at Selected Distances

From the Surface of a 25-cm- (10-in-) Diameter Disk of 0.025-cm-Thick Ceramic Glaze

Containing 20% by Weight of 20-Year-Old Natural Uranium

Distance

(cm)

Absorbed Dose Rate

(mrad/h)a

Fraction of

Skin Exposed

Contribution to

Effective Dose Equivalent

(mrem/h)a

0 24 0.01 0.0024

30 0.84 0.25 0.0021

40 0.49 0.25 0.0012

91 0.089 0.50 0.00045

183 0.017 0.50 8.5×105

457 0.00040 0.50 2.0×106

610 0.000013 0.50 6.6×108

762 0 0.50 0

a

 Assume 1 mrad/h = 1 mrem/h. 1 mrem/h= 0.01 mSv/h.
3–203

Table 3.11.3 Effective Dose Equivalent Rates Due to Photon Irradiation at Selected

Distances From Pieces of Ceramic Tableware Coated With Glaze

Containing 20% by Weight of 20-Year-Old Natural Uranium

Distance

in Air

(cm)

Dose Equivalent Rate

(rem/h)a

 Plate

13 cm (10")

Dish

10 cm (8")

Saucer

5.1 cm (4")

Cup

4.4 cm (3.5")

Bowl

13 cm (10")

Bowl

7.6 cm (6")

1 1.8×105 1.6×105 1.2×105 4.1×105 3.0×105 3.1×105

30 6.5×107 4.4×107 1.2×107 3.7×107 1.1×106 5.5×107

40 3.8×107 2.5×107 6.8×108 2.1×107 6.6×107 3.2×107

91 7.7×108 5.1×108 1.3×108 4.1×108 1.3×107 6.2×108

183 1.9×108 1.3×108 3.3×109 1.1×108 3.3×108 1.5×108

457 3.0×109 2.0×109 5.2×1010 1.6×109 5.3×109 2.5×109

610 1.7×109 1.1×109 2.9×1010 9.0×1010 3.0×109 1.4×109

762 1.1×109 7.1×1010 1.9×1010 5.7×1010 1.9×109 8.7×1010

a

 1 rem/h = 0.01 Sv/h.
3–204

Table 3.11.4 Effective Dose Equivalent Rates Due to Photon Irradiation at Selected

Distances From Arrays of Ceramic Tableware Coated With Glaze

Containing 20% by Weight of 20-Year-Old Natural Uranium

Distance

in Air

(cm)

Dose Equivalent Rate

(rem/h)a

6 Plates

13 cm (10")

6 Dishes

10 cm (8")

6 Saucers

5.1 cm (4")

3 Stacks 

2 Cups

6 Bowls

13 cm (10")

6 Bowls

7.6 cm (6") Carton

1 8.6×105 7.7×105 4.6×105 4.1×105 9.2×105 7.7×105 1.1×104

30 1.8×106 1.3×106 4.5×107 5.9×107 3.5×106 1.9×106 7.7×106

40 1.1×106 8.0×107 2.7×107 3.4×107 2.1×106 1.1×106 4.8×106

91 2.5×107 1.8×107 5.6×108 7.1×108 4.7×107 2.4×107 1.1×106

457 1.1×108 7.8×109 2.4×109 2.9×109 2.1×108 1.1×108 5.1×108

610 6.3×109 4.4×109 1.3×109 1.6×109 1.2×108 6.0×109 2.9×108

762 4.0×109 2.8×109 8.4×1010 1.0×109 7.6×109 3.8×109 1.9×108

a

 1 rem/h = 0.01 Sv/h.
3–205

Table 3.11.5 Summary of the Distribution Model and Potential Individual and

Collective Effective Dose Equivalents a

Step Representation

Cartons

per

Facility

Number

of

Facilities

Highest

Individual

Effective Dose

Equivalent

(mrem)b

Collective

Effective Dose

Equivalent

(person-rem)b

To parcel

 delivery center

Express

 delivery, small

 truck

3,000 2 2.3 0.0050

At parcel

 delivery center

Medium

 warehouse

3,000 2 7.5 0.087

To regional

 truck center 1

Regional

 delivery, large

 truck

900 6 0.19 0.0013

At regional

 truck center 1

Large

 warehouse

900 6 0.53 0.024

To regional

 truck center 2

Regional

 delivery, large

 truck

300 12 0.062 0.00084

At regional

 truck center 2

Large

 warehouse

300 12 0.18 0.016

To wholesaler Local delivery,

 large truck

100 20 0.14 0.0031

At wholesaler Large store 100 20 4.0 13

To large store Local delivery,

 large truck

50 40 0.070 0.0031

At large store Large store 50 40 2.0 13

To small store Local delivery,

 small truck

20 100 0.034 0.00041

At small store Small store 20 100 1.0 22

Total: 50

a

 Based on 20% by weight limit. Multiply by 0.7 for doses using average 14% by weight.

b

 1 mrem = 0.01 mSv; 1 person-rem=0.01 person-Sv.
3–206

Table 3.11.6 Summary of Home Use Model and Potential Individual

Effective Dose Equivalents

Person Activity

Duration

(h/yr)

Distance

(cm) Source

Effective

Dose

Equivalent

(mrem/yr)a

Home

 maker

In house 5,214 760 All arrays in

cabinet

0.1

In kitchen 986 460 All arrays in

cabinet

0.1

Washing dishes 260

6.1

91

Hand contact

1 place setting

1 item (10-in

plate)

0.8

0.02

Dining 365 40 1 place setting 3.3

180 1 place setting 0.2

91 2 place

settings

2.2

Total 7

Others In house 4,780 760 All arrays in

cabinet

0.1

In kitchen 620 460 All arrays in

cabinet

0.07

Dining 365 40 1 place setting 3.3

180 1 place setting 0.2

91 2 place

settings

2.2

Total 6

a

 1 mrem/yr = 0.01 mSv/yr.
3–207

Table 3.11.7 Summary of Potential Radiation Doses From Uranium-Containing,

Glazed Ceramic Tableware a

Exposure Scenario

Individual Annual Effective

Dose Equivalent Rate

(mrem)c

Total Collective Effective

Dose Equivalentb

(person-rem)c

Distribution 6 30

Routine used

 As dinnerware

 On display

 50e

 0.06f 4,000g

Disposal 0.4 8

Misuse 0.4 NAh

a

 External dose estimates are based on the assumption that the glaze contains the maximum

amount of uranium allowed under the exemption, 20% by weight. Collective doses are based

on an annual distribution of 6,000 sets of tableware. The collective dose for dinnerware on

display is based on 14% uranium, the maximum observed. 

b

 Refer to text for time period for collective dose calculations.

c

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

d

 Tableware is assumed to be used or on display for an average of 20 years.

e

 Represents maximum exposed individual, typical usage would yield substantially lower doses. 

Refer to 3.11.4.1.2.

f

 On display in the home.

g

 Six thousand displays with six place settings each on display in a public place over 20 years.

h

 Not applicable.

3–209

3.12 Piezoelectric Ceramic Containing Not More Than 

2 Percent by Weight of Source Material

3.12.1 Introduction

In 10 CFR 40.13(c)(2)(ii), any person who receives, possesses, uses, transfers, or delivers

piezoelectric ceramics containing source material is exempt from licensing requirements

for source material, provided that the ceramic material does not contain more than 2% by

weight of source material. This exemption was proposed on December 10, 1969

(34 FR 19511), and issued as a final rule on April 18, 1970 (35 FR 6313). The ceramic material

normally contains uranium but not thorium.

The first Federal Register notice cited above includes estimates obtained by the Atomic Energy

Commission (AEC) of the radiation dose to individuals. A complete discussion of this product

as originally envisioned and a detailed discussion of the dose estimates for members of the

public are not available. Thus, the validity of the AEC’s analysis of the radiological impact on

the public is difficult to evaluate in terms of current applications and disposal of piezoelectric

ceramics as discussed in Section 3.12.3.

3.12.2 Description of Items

The prefix piezo” comes from the Greek word for pressure (Shackelford, 1992). When

pressure is applied to certain classes of crystalline materials, the crystalline structure produces

a voltage proportional to the pressure. The converse also holds true; a mechanical stress is

produced when an electric field is applied to one of these materials. Such materials have found

use as electromechanical transducers (Cook and Ehrlich, 1993). Piezoelectric properties occur

naturally in some crystalline materials and can be produced in other polycrystalline materials. 

Most contemporary applications of piezoelectricity use polycrystalline ceramics such as barium

titanate, lead metaniobate, or lead zirconate titanate (Haertling, 1986; Berlincourt, 1971;

Harrison and Moratis, 1970). Lead zirconate titanate is commonly referred to as PZT,” a trade

name of the Electro Ceramic Division of Morgan Matroc, Inc. (formerly the Clevite Corporation

and also Vernitron, Inc.). The density of these widely used and manufactured PZT materials

ranges from about 7.45 to 7.75 g/cm3

 and averages approximately 7.6 g/cm3 (Harrison and

Moratis, 1970).

Piezoelectric ceramics are manufactured in many different shapes and sizes for use in a variety

of consumer products that require an electromechanical coupling device (Lancaster, 1993;

Dance, 1993; Haertling, 1986; Adler and Desmares, 1975). Such coupling devices are used in

products such as pacemakers; electronic telephone ringers and tone dialers; microphones;

loudspeakers, particularly tweeters, and super-thin, flat speakers; intermediate frequency filters

in AM and FM radios; gas igniters for stoves, heaters, patio grills, camping lanterns, and

cigarette lighters; flash bulb actuators; depth-finders and fish-finders for pleasure boats;

phonograph pickup cartridges; remote controls for TVs, stereos, VCRs, and air conditioners;

games and toys; audio alerts and alarms in smoke detectors, automobiles, fire protection and

security systems, household appliances, watches, cameras, calculators, and computer

keyboards; nebulizers in ink-jet printers; humidifiers; thickness gauges; accelerometers and

vibration sensors; automotive knock sensors; and gyroscopes.
3–210

The manufacturing process used in the preparation of piezoelectric ceramics is the mixed-oxide

process (Haertling, 1986). The mixed-oxide process consists of (1) weighing the oxides,

(2) blending in a liquid medium, (3) drying to completeness, (4) calcining the powder at

approximately 900o

C for 1 hour, (5) milling in a liquid medium, (6) binder addition, (7) cold

pressing a specific shape, (8) high-temperature firing (sintering) at approximately 1275o

C for

several hours, (9) shaping, (10) electroding, and (11) final poling. The final poling gives the

ceramic its piezoelectric properties. The orientation of the d.c. poling field determines the

orientation of the mechanical and electrical axes. The poling field can be applied so the

ceramic exhibits piezoelectric responses in various directions or combinations of directions

(Haertling, 1986).

From the moment the activated ceramic material is removed from the poling apparatus, the

material properties undergo changes. This process of change is referred to as aging.” Aging

of the ceramic occurs very rapidly in the first few hours, but the material properties tend to

change less rapidly with time after a few days (Berlincourt, 1971). The aging process can be

attributed normally to depolarization due to the relaxation of dipoles in the material. 

Depolarization of the material can result very rapidly, however, if the material is exposed to

excessive heat, a strong electric depoling field, high mechanical stresses, or any combination

thereof (Haertling, 1986). The temperature at which a piezoelectric material is totally

depolarized is defined as the curie point.”

The piezoelectric properties of a ceramic material can be optimized for specific application by

compositional adjustment (Haertling, 1986; Berlincourt, 1971). In PZT-type materials, calcium,

strontium, or barium can be substituted for a fraction of the lead, and tin may be used for the

zirconium, resulting in a piezoelectric ceramic with increased permittivity, but a lower curie

point. This piezoelectric ceramic, designated as PZT-4 material, has wide application in the

ultrasonics industry. In 1969, the Clevite Corporation (now the Electro Ceramic Division of

Morgan Matroc, Inc.) also proposed the addition of uranium oxide to piezoelectric ceramics

made of lead zirconate titanate (34 FR 19511). This piezoelectric ceramic, designated as

PZT-7 material, is used primarily in applications that require good mechanical stress

capabilities, high curie points, and low aging rates. Other companies also were licensed to use

uranium in piezoelectric materials such as Sprague Electric Company in Massachusetts. 

Activity using this material at Sprague Electric Company ceased over a decade ago.

The PZT-7 material was originally envisioned to be useful in such applications as electric wave

filters, phonograph pickups, delay lines, and ultrasonic transducers in flaw detectors, but the

PZT-7 material is expensive and out of the mainstream. Currently, it is used in the following

types of applications: (1) gyroscopes for military applications, accelerometers, and other-type

sensors for aerospace applications, and (2) high frequency delay lines used in the broadcasting

industry to convert U.S. TV signals to European TV signals, and vice versa. There are no

restrictions, however, that prevent the use of PZT-7 material in more common consumer

products such as telephones, remote controls and other electronic equipment, household

appliances, and so forth.

There is no indication in the literature that thorium has ever been used in making piezoelectric

ceramics.
3–211

3.12.3 Summary of Previous Analyses and Assessments

The only known published data on radiological impacts on the public from routine use and

disposal of piezoelectric ceramics is the AEC’s analysis in the first Federal Register notice cited

previously. In the AEC’s analysis, routine exposure scenarios were evaluated for external and

internal exposure during normal use of the material and normal disposal of broken or defective

pieces. The results of the AEC’s analysis of routine exposure scenarios can be summarized as

follows:

Due to the low levels of external radiation for piezoelectric ceramic material that does

not contain more than 2% by weight of source material and the short periods of time that

an individual would use or be near the ceramic materials, annual dose equivalents would

not exceed a few percent of the dose limit for members of the public.

Normal use of piezoelectric ceramic material should not result in a significant internal

dose from inhalation exposure, because degassing of volatile materials is accomplished

during manufacturing.

Normal disposal of broken or defective transducer elements is highly unlikely to result in

any significant radiation problem, and it is unlikely that transducer elements or

piezoelectric ceramic material would be reclaimed and thereby result in any addition of

uranium to other products.

During the time of the AEC’s analysis for external exposure from normal use, the limit on the

annual dose equivalent for members of the public was 5 millisieverts (mSv) (500 mrem). 

Therefore, an annual dose equivalent that would not exceed a few percent of the dose limit

could be 0.1 mSv (10 mrem) or higher.

The one accident of misuse scenario considered by the AEC involved crushing or fracturing of

piezoelectric ceramic. The AEC concluded that the potential inhalation hazard is not significant

because the airborne particles would be too large to be respirable.

3.12.4 Present Exemption Analysis

Annual production of all types of piezoelectric ceramics in the United States is estimated at

several hundreds of tons (1 ton equals 2000 pounds or 908 kg), but the PZT-7 ceramic material

is expensive and out of the mainstream, as discussed previously. The present annual

production of PZT-7 is only 0.25 ton (500 pounds or 227 kg) with less than 1% by weight of

uranium (5 pounds or 2.3 kg) (Phone call, Craig Neir, Electro Ceramics Division, Morgan

Matroc, Inc., Bedford, OH, January 1995). The annual distribution is estimated to be about

200,000 devices containing 12 mg of uranium each. 

To arrive at the estimated annual distribution of 200,000 devices containing the PZT-7 ceramic

material, it is assumed that a typical PZT-7 device such as that used in an accelerometer can

be represented as a disk with a diameter of 1 cm and a thickness of 2 mm. These disks are

estimated to have a mass of 1.2 g based on a density of 7.6 g/cm3

 (see Section 3.12.2) and to

contain about 12 mg of uranium (1.2 g per device times 1% uranium by weight). Further, it is
3–212

assumed that the uranium in the PZT-7 ceramic devices is natural uranium and that the useful

lifetime of the products or instruments containing the devices is 10 years.

Because the typical PZT-7 device defined above may grossly underestimate the potential

radiation dose to an individual routine user, also considered are the dose rates from a much

larger transducer element in the shape of a circular cylinder with a diameter of 10 cm and a

length of 15 cm. This device has a total mass of 9 kg and contains 90 g of natural uranium

(9 kg times 1% by weight uranium). Hence, individual doses to routine users are considered to

be controlled by this device, but all other individual doses resulting from exposure to multiple

devices and, thereby, all collective doses are controlled by exposures to typical PZT-7 devices

containing only 12 mg of natural uranium.

3.12.4.1 Distribution and Transport

This section estimates potential radiation doses to the public from the distribution and transport

of piezoelectric devices containing natural uranium, based on the generic distribution

methodology of Appendix A.3.

In applying the methodology of Appendix A.3, it is assumed that these devices are shipped

primarily by a parcel-delivery service, and that a local driver in a large van picks up the devices

and takes them to a local terminal for shipment to other local terminals for delivery to

customers. A typical shipment is assumed to consist of 800 devices (i.e., the annual

distribution of 200,000 devices divided by 250 working days per year). It is further assumed

that semi-trucks are used to transport the devices between local terminals, and that the devices

pass through an average of four regional terminals before reaching their final destination. The

radiation doses to workers at both local and regional terminals are assumed to be the same as

those estimated for large warehouse workers in Appendix A.3.

Based on the above assumptions and the generic distribution methodology of Appendix A.3, the

individual receiving the largest radiation dose is the local driver, who is assumed to pick up an

average of 800 piezoelectric devices containing uranium from the same manufacturer each day

(250 day/yr). The annual effective dose equivalent (EDE) to this individual is estimated to be

5×105

 mSv (0.005 mrem). Individual doses to other drivers, terminal workers, and members of

the public along truck routes would be less. The annual collective EDE to all parcel-delivery

drivers, terminal workers, and members of the public along truck routes is estimated to be

1×106

 person-Sv (1×104 person-rem).

3.12.4.2 Routine Use

This section estimates potential radiation doses to routine users of piezoelectric devices

containing natural uranium. It is assumed that ingestion and inhalation of natural uranium from

these piezoelectric devices can be ignored during routine use because the piezoelectric

ceramics are degassed by high-temperature firing to as much as 1275o

C during manufacturing

(Haertling, 1986), and these dense ceramic materials are chemically inert and immune to

moisture and other atmospheric conditions (Berlincourt, 1971). Also, it is assumed that these

devices are mounted inside other products or instruments so there is no potential for a betaparticle dose to the skin from touching the piezoelectric devices, and the radiation dose to

routine users results entirely from external irradiation of the whole body by photons from the

natural uranium in the piezoelectric ceramic material.
3–213

To estimate the potential collective dose to routine users, it is assumed that one individual is

exposed at an average distance of 1 m from a typical piezoelectric device for 1000 h/yr and two

individuals also are exposed at an average distance of 2 m for 1000 h/yr. For distances of

1 and 2 m from a typical device containing 12 mg of natural uranium, the EDE rates as

calculated with MicroShield (Computer Codes, Grove Engineering, 1996) were about 0.5 pSv/h

(50 prem/h) and 0.1 pSv/h (10 prem/h), respectively. The self-absorption of photons in the

piezoelectric device and instrument containing the device were ignored in these calculations,

but the natural uranium was decayed for 20 years so that the photons from the short-lived

progeny were included in the dose estimates (see Section 3.1). For these conditions, the

estimated annual EDE to the individuals exposed at 1 m and at 2 m would be less than

1×105

 mSv (<0.001 mrem). For purposes of collective dose estimates, the calculated

individual EDEs of 5×107

 mSv (5×105 mrem) at 1 m and 1×107 mSv (1×105 mrem) at 2 m

are used instead of the less than value. Thus, the annual collective EDE for the 200,000

devices distributed per year is (5×107

 mSv/device) × (200,000 devices) × (1 person) +

(1×107

 mSv/device) × (200,000 devices) × (2 persons), or about 1×104 person-Sv

(1×102

 person-rem), and the total collective EDE over the 10-year useful lifetime of the devices

could be 0.001 person-Sv (0.1 person-rem).

To estimate the maximum potential dose to a routine user, it is assumed that the piezoelectric

device might be a large transducer element in the shape of a circular cylinder with a diameter of

10 cm and a length of 15 cm. This cylinder would have a mass of 9 kg based on a density of

7.6 g/cm3

 (see Section 3.12.2) and contain 180 g of natural uranium (9 kg times 2% by weight

uranium). If the self-absorption of photons within this device is taken into account, the dose

rate at 1 m is about 2×106

 mSv/h (2×104 mrem/h). In these calculations with MicroShield

(Computer Codes, Grove Engineering, 1996), the natural uranium was decayed for 20 years

and the composition of the PZT-7 ceramic material was taken to be 65% by weight PbO, 2% by

weight UO2, and 17% by weight ZrO2 and 17% by weight of TiO2. Thus, the maximum annual

EDE to a routine user is estimated to be 0.002 mSv (0.2 mrem) if the individual is exposed at an

average distance of 1 m from this device for 1000 h/yr.

3.12.4.3 Disposal

This section estimates potential radiation doses to the public from disposal of piezoelectric

devices containing natural uranium based on the generic disposal methodology in

Appendix A.2.

In applying this methodology, it is assumed that 1 year’s distribution of piezoelectric devices

contains 2.3 kg of natural uranium, and it is in a form that is not readily dispersible or readily

accessible to groundwater. Thus, the following adjustments are made to the dose-to-source

ratios in Appendix A.2: (1) there is no exposure by inhalation or ingestion for waste collectors at

either landfills or incinerators or for workers at landfills, (2) there is no exposure to off-site

members of the public from airborne releases at landfills, (3) there is a reduction by a factor of

10 in the exposure to off-site members of the public from groundwater releases, and (4) there is

a reduction by a factor of 10 in the exposure to future on-site residents by inhalation and

ingestion.

For landfill disposal, the annual EDE to waste collectors would be less than 1×105

 mSv

(<0.001 mrem). The annual individual doses to workers at landfills, off-site members of the

public, and future on-site residents are less than that to the waste collector. The total collective
3–214

dose could be about 6×105

 person-Sv (0.006 person-rem), due almost entirely to exposure to

future on-site residents for 1000 years after loss of institutional controls over the landfill sites. 

For disposal by incinerator, the annual EDE to waste collectors would be less than 1×105

 mSv

(<0.001 mrem). Annual doses to workers at incinerators and off-site members of the public are

less than to the waste collector. The total collective EDE could be 2×107

 person-Sv

(2×105

 person-rem), due mainly to exposures to waste collectors at incinerators and off-site

members of the public from airborne releases during incinerator operations.

3.12.4.4 Accidents and Misuse

This sections considers: (1) the external radiation dose to a repairperson who routinely carries

at least one piezoelectric device in the pants pocket of his or her coveralls, and (2) the internal

radiation dose to individuals from a transportation accident involving a fire.

In the case of misuse, it is assumed a repairperson carries at least one small piezoelectric

ceramic disk in the pants pocket of his or her coveralls for 2000 h/yr. Characteristics of this

disk are assumed to be as follows: (1) it is a circular disk with a diameter of 1 cm and thickness

of 2 mm (see Section 3.12.4), (2) the density is 7.6 g/cm3

 (see Section 3.12.2), (3) the amount

of natural uranium contained in the device is 1% by weight (see Section 3.12.4), and (4) the

natural uranium has decayed for 20 years so the short-lived progeny are included in the dose

estimates (see Section 3.1). Using the modeling described in Appendix A.4, the source is

assumed to be located 1 cm from the skin in the pocket, which has a thickness of 0.7 mm and a

density of 0.4 gm/cm3

. Based on calculations using VARSKIN MOD2 (Computer Codes,

Durham, 1992), the dose equivalent rate from beta particles to a small area of skin under the

pants pocket of the coveralls is estimated to be 2×103

 mSv/h (0.2 mrem/h). Based on

calculations with MicroShield (Computer Codes, Grove Engineering, 1996), the EDE rate from

whole-body irradiation by photons is estimated to be 5×108

 mSv/h (5×106 mrem/h). The latter

dose rate is calculated at a body depth of 10 cm, which is considered to be a reasonable

approximation for the average depth of the body organs relative to a small source on the

surface of the body. Thus, the annual dose equivalent to a small area of skin from beta

particles could be 4 mSv (400 mrem). Because of the small exposed skin area (1 cm2

assumed), the contribution to the EDE is negligible (less than 1×105 mSv (<0.001 mrem)). The

annual EDE from whole-body irradiation by photons could be 1×104

 mSv (0.01 mrem), if a

repairperson carried this disk in the coveralls pants pocket for 2000 h/yr.

The above estimates are based on a device with 1% uranium by weight. At the exemption limit

of 2%, dose estimates would be twice these values.

In the case of an accident, it is assumed that a transportation fire involves a typical shipment of

800 piezoelectric devices containing a total of 9.6 g of natural uranium (see Section 3.12.4.1). 

It is also assumed that the release fraction for the natural uranium in the devices is 0.01% the

same as for devices in protective coverings. (See Appendix A.1). Based on these assumptions

and the generic accident methodology of Appendix A.1, the radiation doses are estimated to be

less than 1×105

 mSv (<0.001 mrem) to a firefighter who wears a respirator during the fire and

8×105

 mSv (0.008 mrem) to a worker who is involved in the cleanup following the fire and who

does not wear a respirator.
3–215

For a transportation fire involving a single large transducer with 90 g uranium, the EDE to the

firefighter would be less than 1×105 mSv (<0.001 mrem) and 7×104 mSv (0.07 mrem) for

clean-up.

3.12.5 Summary

Table 3.12.1 presents the results of this assessment of potential radiological impacts on the

public from the distribution, routine use, and disposal of piezoelectric devices containing

uranium. These results are based on an assumed annual distribution of 200,000 piezoelectric

devices containing 1% (12 mg) of natural uranium each and a useful lifetime of 10 years for the

products containing these devices. Maximum individual dose for routine uses and accident

(fire) is based on exposure to a single large transducer containing 90 g of uranium.

For routine use of these piezoelectric devices, including distribution and disposal in landfills or

by incineration, the annual EDE to the most highly exposed individuals (routine users of a

piezoelectric device) is estimated to be 0.001 mSv (0.1 mrem). The total collective dose is

estimated to be about 2×103

 person-Sv (0.2 person-rem), due almost entirely to exposure to

routine users of the piezoelectric devices. 

For accidents involving fire, the individual EDEs were estimated to be 7×104

 mSv (0.07 mrem)

or less. For misuse by a repairperson routinely carrying a piezoelectric device with 1% uranium

by weight in the coveralls pants pocket, the estimated annual EDE was 1×104

 mSv (0.01mrem)

and an annual dose equivalent from beta particles to a small area of skin beneath the pants

pocket of the coverall of as much as 4 mSv (400 mrem).
3–216

Table 3.12.1 Potential Radiation Doses From Piezoelectric Devices 

Containing Uranium

Exposure Pathway

Individual Annual Effective

Dose Equivalent

(mrem)b

Collective Effective

Dose Equivalenta

(person-rem)b

Distribution and transport

Routine use

Disposal

 Landfills

 Incineration

Accidents and misuse

 Fire

 Carrying in pocket

0.005c

0.2d

<0.001e

<0.001f

0.07g

0.01h

1×104

0.1

6×103

2×105

a

 Collective doses are based on an assumed annual distribution of 200,000 piezoelectric

devices containing 12 mg of natural uranium per device and an assumed useful lifetime of

10 years for products or instruments containing the piezoelectric devices.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 Dose estimate applies to local parcel-delivery driver; dose estimates are less for terminal

workers, long-haul semi-truck drivers, and members of public along truck routes (see

Section 3.12.4.1).

d

 Dose estimate applies to a user who is routinely exposed to a large piezoelectric device

containing 90 g of natural uranium (see Section 3.12.4.2).

e

 Dose estimate applies to waste collectors at landfills; dose estimates are less for workers at

landfills, off-site members of the public from groundwater releases, and future on-site residents,

and dose estimates are essentially zero (0) for off-site members of the public from airborne

releases during landfill operations (see Section 3.12.4.3).

f

 Dose estimate applies to waste collectors at incinerators; dose estimates are less for workers

at incinerators and off-site members of the public (see Section 3.12.4.3).

g

 Dose estimate applies to a worker who is involved in the cleanup following a fire involving a

single large transducer (90 g uranium) and who does not wear a respirator; dose estimate is

less for a firefighter who is assumed to wear a respirator during the fire (see Section 3.12.4.4).

h

 Dose estimate applies to whole-body irradiation of a repairperson carrying a 12 mg uranium

(1%) device in a coveralls pocket; dose estimate for the annual dose equivalent from beta

particles to a small area of skin beneath the pants pocket of the coveralls could be 4 mSv

(400 mrem). At the 2% exemption limit, the estimated doses would be twice these values. (See

Section 3.12.4.4).
3–217

3.13 Glassware

3.13.1 Introduction

In 10 CFR 40.13(c)(2)(iii), the receipt, possession, use, transfer, ownership, and acquisition of

glassware containing source material are exempted from licensing requirements, provided that

the glassware does not contain more than 10% by weight of source material. The exemption

does not apply to commercially manufactured glass brick, pane glass, ceramic tile, or other

glass or ceramic used in construction. The glassware normally contains uranium rather than

thorium. An exemption for unspecified glass products was first established on March 20, 1947

(12 FR 1855). The present exemption for glassware was proposed on September 7, 1960

(25 FR 8619), and issued as a final rule on January 14, 1961 (26 FR 284).

Neither the initial 1947 notice in the Federal Register nor the later notices from 1960 and 1961

cited above quantify the radiological impacts on the public from use of the exempted glassware. 

The 1960 notice states that possession and use of such glassware would not result in an

unreasonable hazard to life or property, but does not present a supporting dose analysis.

Potential radiological impacts on the public from use of uranium-containing glassware are

associated with beta-particle irradiation while handling glassware, beta-particle and gamma-ray

irradiation while near glassware, and ingestion of uranium leached into foodstuffs that had been

in contact with glassware.

3.13.2 Description of Exempt Items

Source material (depleted or natural uranium) is added to glassware as a permanent coloring

agent. Such glassware is formed by blending specific quantities of sodium diuranate (Na2U2O7)

or sodium uranyl carbonate (2Na2Ca3U·O2CO3) with other glass ingredients and smelting the

blend to produce a new glass. As a result, the uranium is dispersed uniformly throughout the

glass, as uranium oxide (U3O8), and becomes an integral part of the glass. The uranium

content of the glass is variable, ranging between 0.26 and 10% by weight. Lower uranium

contents appear to be more common. Buckley et al. (1980) estimated the uranium content of a

decorative item to be 0.91 g. A variety of glass colors can be produced in this manner. The

most common colors are yellow and yellow-green; other colors include orange, vermillion red,

and white. The observed color may depend on the type of illumination. For example, a glass

that appears yellow-green in transmitted light may appear emerald green in reflected light and

fluorescent green in ultraviolet light. 

A variety of objects, both decorative and practical, can be formed from uranium-containing

glasses. Known objects include various designs of drinking glasses, wine glasses, tumblers,

candy dishes, vases, pitchers, goblets, ash trays, candlestick holders, and other ornamental

and decorative objects. Based on information supplied by one of the few known domestic

manufacturers of uranium-bearing glassware, current production of uranium-bearing glass is

limited to the manufacture of decorative and ornamental objects. In addition, this manufacturer

declined to make current production figures available.

Buckley et al. (NUREG/CR–1775) estimated production of at least 4,160,000 pieces of

decorative glassware over a 21-year period (1958 to 1978), approximately 200,000 pieces per
3–218

year, and 15,000 drinking glasses over a 5-year period (1968 to 1972), approximately 3,000 per

year. 

The number of previously produced pieces still in use is unknown. Although possibly still used

for serving food and beverages, previously produced pieces of glassware are now considered

collectible items (Landa and Councell, 1992) and likely would be displayed as such. Currently

produced decorative items also would be displayed in homes, shops, and museums. The

above considerations, however, do not preclude the possibility that uranium is present in

imported glassware, including other than decorative items.

3.13.3 Summary of Previous Analyses and Assessments

Previous studies were concerned almost exclusively with determination of the rates at which

uranium leaches from glassware into various liquids. External dose rate measurements were

performed to establish that the glass was radioactive, but are, at best, qualitative. Landa and

Councell (1992) contacted a variety of items of uranium-bearing glassware with 20 mL of three

different solutions: (1) water, (2) 4% acetic acid, and (3) 1 molar nitric acid. After 24 hours of

contact, uranium concentrations in the solutions ranged from <0.03 (the lower limit of detection)

to 0.63 g/L in water; <0.03 to 30.1 g/L in acetic acid; and 0.1 to 29.7 g/L in nitric acid. 

Average concentrations of uranium leached from items of glassware designed to hold

beverages were 0.052 g/L for water and 5.9 g/L for acetic acid. Landa and Councell (1992)

noted the possibility of a memory effect on the leaching process. (The amount of uranium

leached from an item per period of contact will decrease as the number of contact periods

increases.) Such an effect would be expected as the amount of uranium available for leaching

decreases after each leaching operation. The above study reports neither the uranium

concentrations nor the surface areas of the glasses that were in contact with the leaching

solutions. Normalizing the above results with respect to uranium concentration and contacted

surface area could yield a better measurement for comparing leach rates from a variety of

objects (i.e., micrograms of uranium per liter of leachate, per hour of contact.

3.13.4 Present Exemption Analysis

3.13.4.1 General Information

All currently manufactured items of glassware that contain uranium are intended for use as

decorative items. Therefore, most pieces of currently manufactured glassware are sold singly

or in pairs. Most older pieces of such glassware were designed not only for decorative uses,

but for use as drinking glasses and food-containing accessories as well. No items designed for

use as dinnerware have been identified (Landa and Councell, 1992; NUREG/CR–1775). 

Therefore, the use of glassware that contains uranium as dinnerware is not evaluated. 

As noted in Section 3.13.2, the availability of glassware containing uranium cannot be

quantified. The present analysis assumes an annual distribution similar to that reported in

Buckley et al. (NUREG/CR–1775), 200,000 pieces of decorative glassware and 3,000 drinking

glasses. The analyses in this section include: (1) individual and collective doses due to the

distribution of the glassware, (2) individual and collective doses due to the routine use of

drinking glasses, (3) individual and collective doses due to the display of decorative pieces in

the home, (4) individual and collective doses due to the display of decorative pieces in a public
3–219

viewing area, and (5) individual and collective doses due to disposal. For assessment

purposes, a piece of decorative glassware is represented as an 18-cm-diameter glass disk with

a thickness of 0.64 cm, whereas a drinking glass is represented as an 18-cm-diameter disk with

a thickness of 0.32 cm. The glass is modeled as a material having the composition of Pyrex

glass, density of 2.4 g/cm3

, and an atomic number of 10.2. Individual doses have been

estimated using the exemption limit of 10% uranium by weight; distribution, disposal and all

collective doses use an assumed average concentration of 5%.

3.13.4.1.1 External Exposures

External exposures to uranium containing glassware involve exposures both to beta particles

and gamma rays. Beta particle exposure is especially important during the handling of

glassware. Absorbed dose rates in tissue due to beta particle irradiation were calculated using

VARSKIN MOD 2 (Computer Codes, Durham, 1992). Dose rates were calculated at a depth of

7 mg/cm2

 in tissue from contact with and from exposures at selected distances from the disks

described above. Dose rates decrease from 0.27 millisievert (mSv)/h (27 mrem/h) at contact to

0.032 nSv/h (3.2 nrem/h) at 610 cm (see Table 3.13.1). Dose rates for both source

representations are essentially the same. The contribution of beta particle irradiation to the

effective dose equivalent (EDE) rate at a selected distance from a piece of glassware is

obtained by multiplying the calculated dose rate by the skin weighting factor (0.01) and the

fraction of skin under irradiation. For handling a plate, the dose rate to the palms and fingers is

0.27 mSv/h (27 mrem/h). The skin of the palms and fingers constitute about 1% of the total

skin area. Therefore, the contribution of beta particle irradiation to the EDE rate is

2.7×105

 mSv/h (2.7×103 mrem/h). For exposures at less than 90 cm, one fourth of the total

skin area is assumed to be irradiated; for exposure distances greater than 90 cm, one half of

the total skin area is assumed to be irradiated. Dose rates due to gamma and bremsstrahlung

irradiation were calculated using the PC version of CONDOS II (Computer Codes, O’Donnell 

et al., 1981), which uses current dosimetry data and dose conversion factors in performing the

same dose calculations as those performed using CONDOS II. Resulting dose rates are

presented in Table 3.13.1. In addition, Table 3.13.2 presents a list of calculated EDE rates at

selected distances from a drinking glass and decorative piece.

 

3.13.4.1.2 Internal Exposures

After a 24-hour soak, average reported concentrations of uranium leached from glassware

designed for drinking purposes were 0.052 g/L into water and 5.9 g/L into 4% acetic acid

(see Section 3.13.3). Normalizing these reported concentrations with respect to their reported

leach durations yields effective concentration rates of 0.0022 g/L/h in water and 0.25 g/L/h in

4% acetic acid. In the following assessment, the average 24-hour contact concentrations,

0.052 g/L in water and 5.9 g/L in 4% acetic acid solutions, are used for liquids that contact

glassware for 24 hours or more (e.g., liquids not consumed in one sitting). For liquids in contact

with glassware for less than 24 hours, the normalized concentrations, 0.0022 g/L/h in water

and 0.25 g/L/h in acetic acid, are multiplied by the assumed contact duration. 

Based on Environmental Protection Agency (EPA) (EPA/600/P–95/002Fa) studies, the typical

daily intake of liquids is approximately 2 L. In Section 3.11.4.1.2, this intake is broken down into

intakes of water-like and acid-like liquids that contacted glassware and ceramic ware for two

time periods, 30 minutes and more than 24 hours. Using the breakdown as for tableware and

considering only liquids drunk from glassware, an individual might hypothetically use glassware
12 (36 L of 30-min acidic×0.25 g/L/h×0.5 h) + (290 L of 30-min water-like×0.0022 g/L/h×0.5 h)

+ (96 L of 24-h water-like × 0.052 g/L) = 9.7 g.

3–220

containing uranium to consume about 290 L/yr of water-like and 36 L/yr of acid-like liquids that

were in contact with glassware for 30 minutes, plus 96 L/yr of water-like liquids that were in

contact with glassware for 24 hours or more. Based on these consumption rates and the

leaching factors discussed above, an individual could ingest approximately 9.7 g of uranium

during 1 year.12 Thus, given an ingestion dose factor of 1.9×106

 mSv/g (1.9×10-4 mrem/g),

this individual could receive an annual EDE of about 1.8×105

 mSv (1.8×10-3 mrem) from

ingestion of uranium leached from glassware.

3.13.4.2 Distribution

As noted above, domestic production of glassware containing uranium is attributed to only one

of a few manufacturers, and import of such glassware, if occurring, cannot be quantified. 

However, based on past distribution, this assessment is based on an annual distribution of

3,000 drinking glasses by one manufacturer and 200,000 pieces of decorative glassware by a

second manufacturer.

3.13.4.2.1 Drinking Glasses

Drinking glasses are packed six per carton; therefore, shipment of 3,000 glasses requires 500

cartons. One manufacturer ships 400 cartons of glasses to a nearby truck-parcel distribution

center (large warehouse) and 100 cartons to an air-parcel distribution center (air freight

terminal). (See Table 3.13.3 for a list of the model steps, the number of facilities involved in

each step, and the number of cartons handled at each facility.) The manufacturer loads 400

cartons into a small express-delivery truck that transports them to a truck parcel-delivery center. 

At the center, (1) 4 cartons are loaded into each of five large local delivery trucks and

transported to five small retail stores; (2) 10 cartons are loaded into each of two large localdelivery trucks and transported to two large retail stores; and (3) 90 cartons are loaded into

each of four large regional-delivery trucks and transported to four intermediate truck terminals. 

At each intermediate truck terminal, 18 cartons are loaded into each of five large regionaldelivery trucks and transported to five final regional truck terminals. At each final truck terminal,

(1) 4 cartons are loaded into each of two large local-delivery trucks and transported to two small

retail stores; and (2) 10 cartons are loaded into a large local-delivery truck and transported to a

large retail store.

The exposure conditions, calculational methods, and dose factors given in Appendix A.3.3 are

used to calculate individual and collective EDEs for each step in the model. The results of the

calculations are presented in Table 3.13.3. The highest calculated individual EDE was

approximately 4×104

 mSv (0.04 mrem), to a worker in a large retail store. The total collective

EDE for distribution was about 0.002 person-Sv (0.2 person-rem), almost entirely from

exposures at retail establishments. Due to the number of cartons carried in each truck, the

average truck driver exposure conditions were used in the calculations (see Appendix A.3.3).

The manufacturer also loads 100 cartons into a small express-delivery truck that transports

them to an air-parcel-delivery center. At the center, 20 cartons are loaded into each of five
3–221

airplanes and transported to five receiving airports. At each receiving airport, 4 cartons are

loaded into each of five large local-delivery trucks for transport to five small retail stores.

The exposure conditions, calculational methods, and dose factors given in Appendix A.3.3 are

used to calculate individual and collective EDEs for each step in the model. Results of the

calculations are presented in Table 3.13.3. The highest calculated individual EDE was

approximately 1×104

 mSv (0.01 mrem), to a worker in a small retail store. The total collective

EDE for distribution was about 8×104

 person-Sv (0.08 person-rem), almost entirely due to

exposures at retail establishments. As noted above, average truck driver exposure conditions

were used in the calculations.

3.13.4.2.2 Decorative Items

Decorative items are packed 10 per carton; therefore, shipment of 200,000 items requires

20,000 cartons. One manufacturer ships all cartons of decorative items to a nearby truckparcel distribution center (large warehouse). (See Table 3.13.4 for a list of the model steps, the

number of facilities involved in each step, and the number of cartons handled at each facility.)

The manufacturer loads 20,000 cartons into a small express-delivery truck that transports them

to a truck-parcel-delivery center. At the center, (1) 20 cartons are loaded into each of 50 large

local-delivery trucks and transported to 50 small retail stores; (2) 50 cartons are loaded into

each of 20 large local-delivery trucks and transported to 20 large retail stores; and (3) 450

cartons are loaded into each of 40 large regional-delivery trucks and transported to 40

intermediate truck terminals. At each intermediate truck terminal, 90 cartons are loaded into

each of 50 large regional-delivery trucks and transported to 50 final regional truck terminals. At

each final truck terminal, (1) 20 cartons are loaded into each of twenty large local-delivery

trucks and transported to twenty small retail stores; and (2) 50 cartons are loaded into each of

10 large local-delivery trucks and transported to ten large retail stores.

The exposure conditions, calculational methods, and dose factors given in Appendix A.3.3 are

used to calculate individual and collective EDEs for each step in the model. The results of the

calculations are presented in Table 3.13.4. The highest calculated individual EDE was

approximately 0.04 mSv (4 mrem), to the initial small express-delivery truck driver. The total

collective EDE for distribution was about 0.5 person-Sv (50 person-rem), almost entirely due to

exposures at retail establishments. As noted above, the average truck driver exposure

conditions were used in the calculations (see Appendix A.3.3).

3.13.4.3 Routine Use

3.13.4.3.1 Drinking Glasses

Members of a household can be exposed to the uranium contained in glassware both externally

and internally. External exposures can occur while dining, moving glassware before and after

dining, washing and storing glassware, and merely being in the home, especially in the kitchen

where arrays of glassware are stored in cabinets. Internal exposure can occur via ingestion of

liquids in contact with the uranium-containing glass. The following model, which is described

partially in Table 3.13.5, uses the dose rates given in Tables 3.13.1 and 3.13.2, and the

uranium concentration rates given in Section 3.13.4.1.2, to estimate EDEs and dose

equivalents to skin from use of 500 sets of glassware by 500 families of four persons. External
3–222

exposure durations were based on studies of homemakers’ activities (Steidl and Bratton, 1968;

EPA/600/P–95/002Fa); exposure distances are based on observation. Liquid intakes are as

described above.

A highly exposed person is designated as a homemaker in Table 3.13.5. This person is

assumed to wash all dishes and to spend the most time at home and in the kitchen, as well as

dining. These activities could result in an EDE of about 0.02 mSv/yr (2 mrem/yr) from external

exposures. The dose from ingestion of uranium that has leached from the glasses into liquids,

2×105

 mSv/yr (2×103 mrem/yr)), is negligible with respect to the EDE from external exposures. 

Thus a highly exposed individual could receive a total EDE of approximately 0.02 mSv/yr

(2 mrem/yr) of glassware use. Other family members could receive EDEs of about 0.01 mSv/yr

(1 mrem/yr). At an assumed average 5% uranium by weight, doses would be one-half the

above values.

The collective EDE to a family of four could be approximately 5×105

 person-Sv/yr

(0.005 person-rem/yr) and 0.001 person-Sv (0.1 person-rem) per 20 years. For use of 500 sets

of drinking glasses by 500 families of four for 20 years, the collective EDE could be

approximately 0.6 person-Sv (60 person-rem).

3.13.4.3.2 On Display as Collectibles

Collectibles may be displayed in many ways, both in homes and in public facilities, such as

museums and stores that sell collectibles. This assessment is based on display of four items in

each of 25,000 homes and 10 items in each of 10,000 public places. The highly exposed

individual likely would be a person who displays glassware in the home. Display in a public

facility is the largest contributor to collective dose.

Four pieces of decorative glassware are assumed to be scattered about a home. A highly

exposed individual is a family member who handles individual pieces for 6.1 h/yr, views them

from a distance of 91 cm for 10 h/yr, and is in other rooms at average distances of 610 and

762 cm for 200 and 6000 h/yr, respectively. This individual could receive an EDE of

0.002 mSv/yr (0.2 mrem/yr). Using 5% uranium by weight as an average concentration, the

collective EDE for 25,000 homes with a family of 4 would be 1 person Sv (100 person-rem).

Ten pieces of glassware (5% uranium) on display in a public facility are viewed by 500

persons/day (0.18 million persons/year). Each person spends 5 minutes at 91 cm from the

display, 30 minutes at 460 cm from the display, and 3 hours at an average distance of 762 cm

from the display. With an individual EDE estimated as 3×106

 m Sv (3×104 mrem), the

collective EDEs per facility could be 0.01 person-Sv (1 person-rem) over 20 years. If 100

decorative items each were on display in 1000 such facilities, the 20-year collective EDE could

be approximately 100 person-Sv (10,000 person-rem).

3.13.4.4 Accidents and Misuse

It is inconceivable that an accident involving glassware could have radiological consequences

much greater than those from routine distribution and use. Even in the event of a fire, glass is

unlikely to become airborne. Thus, any radiological consequences of an accident involving

glassware would be from handling the residue of the accident. Such handling is unlikely to
3–223

result in an EDE greater than the 0.04 mSv (4 mrem) to the driver who delivers 20,000 cartons

of decorative tableware.

Misuses of glassware might include use as containers for miscellaneous items. However, these

misuses should not yield radiation doses greater than those calculated for distribution, use, and

disposal. Thus, it is unlikely that an EDE from misuse would exceed 0.04 mSv (4 mrem).

3.13.4.5 Disposal

Under normal circumstances, glassware would be disposed of as ordinary, noncombustible

household trash. The following assessment assumes discard of 4 Mg of uranium in 215,000

pieces of glassware in 1 year. This assumption omits consideration of random discard of

individual pieces of glassware prior to the year of discard. This is not an important omission

because some of the broken pieces may be replaced, individual doses will be maximized if all

pieces are disposed of at once, and collective doses will be affected little, if at all.

Using the assumptions of the generic disposal methodology (see Appendix A.2) for disposal of

4 Mg of natural uranium, the highest calculated individual EDE is 0.008 mSv (0.8 mrem), to a

waste collector at a municipal incinerator. No other individual is estimated to receive an EDE

greater than 0.001 mSv (0.1 mrem). The total collective EDE to all workers and potentially

exposed members of the public could be about 0.2 person-Sv (20 person-rem).

3.13.5 Summary

This assessment has considered potential doses to the public from use of uranium as a

coloring agent in glassware. Individual dose calculations were based on the maximum

concentration (10% by weight) of uranium allowed in the glass. Distribution disposal, and all

collective doses were estimated using an assumed average uranium concentration of 5% by

weight. Actual concentrations of uranium are reported to be as low as 0.26% by weight. The

only known domestic manufacturer of uranium-bearing decorative and ornamental glassware

claims to use uranium concentrations much lower than the limit. Also, many of the assumptions

used in the dose calculations are on the conservative side. Thus, the doses reported herein

may be much higher than the doses based on actual uranium concentrations in glass.

As summarized in Table 3.13.6, the highest hypothetical EDE, approximately 0.04 mSv/yr

(4 mrem/yr), is associated with the transport of glassware from a manufacturer to a truck

distribution center. The total collective dose due to distribution, use, and disposal could be as

high as 100 person-Sv (10,000 person-rem) if 100,000 pieces of decorative glassware were to

be placed on public display (e.g., 10 pieces in each of 10,000 museums) for 20 years.
3–224

Table 3.13.1 Absorbed Dose Rates at 7 mg/cm2 in Tissue at Selected Distances From

the Surface of a 0.64-cm-Thick, 25-cm- (10-in-) Diameter Glass Disk Containing

10% by Weight of 20-Year-Old Natural Uranium

Distance

(cm)

Absorbed Dose Rate

(mrad/h)a

Fraction of Skin

Exposed

Contribution to EDE

(mrem/h)b

0 27 0.01 0.0027

2.5 20 0.01 0.0020

30 1.1 0.25 0.0027

40 0.59 0.25 0.0015

91 0.095 0.50 4.7×104

183 0.015 0.50 7.3×105

457 0.00017 0.50 8.7×107

610 0.0000032 0.50 1.6×108

762 0 0.50 0

a

 Assume 1 mrad/h = 1 mrem/h; 1mrad/h = 10 microgray (Gy)/h.

b

 1 mrem/h = 0.01 mSv/h
3–225

Table 3.13.2 Effective Dose Equivalent Rates Due to Photon Irradiation at Selected

Distances From Pieces of Glassware Containing 10% by Weight

of 20-Year-Old Natural Uranium

Distance in Air

(cm)

Dose Equivalent Rate

(rem/h)a

Drinking Glass Decorative Piece

1 4.3×105 7.8×105

30 9.0×107 1.8×106

40 5.2×107 1.0×106

91 1.0×107 2.0×107

183 2.5×108 5.0×108

457 4.0×109 7.9×109

610 2.2×109 4.4×109

762 1.4×109 2.8×109

a

 1 rem/h = 0.01 Sv/h.
3–226

Table 3.13.3 Distribution Model and Potential Individual and Collective Effective Dose

Equivalents (EDEs) for Drinking Glasses a

Step Representation

Cartons

per

Facility

Number

of

Facilities

 Individual

Annual

EDE

(mrem)b

Collective EDE

 (person-rem)b

DISTRIBUTION VIA TRUCK-Parcel Delivery

To parcel-delivery

 center

Express delivery,

 small truck

400 1 0.02 0.00002

At parcel-delivery

 center

Large warehouse 400 1 0.02 0.0001

To intermediate

 truck terminal Regional delivery, large truck

90 4 0.001 0.000006

At intermediate

 truck terminal

Large warehouse 90 4 0.004 0.0001

To final truck

 terminal Regional delivery, large truck

18 20 0.0004 0.000006

At final truck

 terminal

Large warehouse 18 20 0.0008 0.0001

To large store Local delivery, large truck10 22 0.001 0.00002

At large store Large store 10 22 0.04 0.1

To small store Local delivery,

 large truck

4 45 0.0004 0.00002

At small store Small store 4 45 0.01 0.1

Total 0.2

DISTRIBUTION VIA AIR-Parcel-delivery

To airport 1 Express delivery, small truck100 1 0.006 0.00003

At airport 1 Air freight terminal 100 1 0.004 0.00001

To airport 2 Airplane 20 5 0.004 0.0004

At airport 2 Air freight terminal 20 5 0.0006 0.00001

To small store Local delivery,

 large truck

4 25 0.0004 0.00001

At small store Small store 4 25 0.01 0.08

Total 0.08

a

 Based on assumed average 5% uranium by weight.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.
3–227

Table 3.13.4 Distribution Model and Potential Individual and Collective Effective Dose

Equivalents (EDEs) for Decorative Objects a

Step Representation

Cartons

per

Facility

Number

of

Facilities

 Individual

Annual

EDE

(mrem)b

Collective

EDE

(person-rem)b

DISTRIBUTION VIA TRUCK-Parcel-delivery

To parceldelivery

 center

Express delivery,

 small truck

20,000 1 4 0.004

At parcel-delivery

 center

Large warehouse 20,000 1 3 0.02

To intermediate

 truck terminal

Regional delivery,

 large truck

450 40 0.02 0.001

At intermediate

 truck terminal

Large warehouse 450 40 0.07 0.002

To final truck

 terminal

Regional delivery,

 large truck

90 200 0.005 0.001

At final truck

 terminal

Large warehouse 90 200 0.02 0.02

To large store Local delivery,

 large truck

50 220 0.02 0.004

At large store Large store 50 220 0.6 20

To small store Local delivery,

 large truck

20 450 0.007 0.004

At small store Small store 20 450 0.3 30

Total 50

a

 Based on assumed average 5% uranium by weight.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.
3–228

Table 3.13.5 Summary of Home Use Model and Potential Individual Effective

Dose Equivalents a

Person Activity

Duration

(h/yr)

Distance

(cm) Source

Effective Dose

Equivalent

(mrem/yr)b

Home-

 maker

In house 5,214 760

(no beta)

6 glasses in

 cabinet

0.04

In kitchen 986 460

(no beta)

6 glasses in

cabinet

0.02

Washing

 dishes

260

6.1

91

Contact

(beta only)

6 glasses

6 glasses

0.9

0.02

Dining 365 40 1 glass 0.7

180 1 glass 0.04

91 2 glasses 0.4

0.61 Contact

(beta only)

1 glass 0.002

Total 2

Others In house 4,780 760

(no beta)

6 glasses in

 cabinet

0.04

In kitchen 620 460

(no beta)

6 glasses in

 cabinet

0.02

Dining 365

40

1 glass 0.7

180 1 glass 0.04

91 2 glasses 0.4

 0.61 Contact

(beta only)

1 glass 0.002

Total 1

a

 Based an 10% uranium by weight.

b

 1 mrem/yr = 0.01 mSv/yr.
3–229

Table 3.13.6 Summary of Potential Radiation Doses From Glassware 

Containing Uranium

Exposure Scenario

Individual Annual Effective

Dose Equivalent Rate

(mrem)b

Total Collective Effective

Dose Equivalenta

(person-rem)b

Distribution 4 50

Routine usec

 As drinking glasses

 On display

2

0.2d

 60

 10,000e

Disposal

 Incinerators 0.8 20

Accidents or misuse 4 NAf

a

 Based on an assumed average 5% uranium by weight. Refer to text for time period of

collective dose calculations.

b

 1 mrem = 0.01 mSv; 1 person rem = 0.01 person-Sv.

c

 Glassware is assumed to be used or on display for an average of 20 years.

d

 On display in a home at exemption of 10% uranium by weight.

e

 On display in public places at average of 5% uranium by weight.

f

 Not applicable.

13 Telegram to Yu Sung, Director, Coordination Council for North American Affairs, Washington,

D.C., from Chen-Wa Cheng, Secretary General, Atomic Energy Commission Taiwan, 1983.

3–231

3.14 Glass Enamel and Glass Enamel Frit Containing Source Material

3.14.1 Introduction

In 10 CFR 40.13 (c)(2)(iv), any person who receives, possesses, uses, or transfers glass

enamel and glass enamel frit is exempted from licensing requirements for source material,

provided that the enamel and enamel frit do not contain more than 10% by weight of source

material. In addition, this exemption applies only to glass enamel and glass enamel frit

imported or ordered for importation into the United States, or initially distributed by

manufacturers in the United States, prior to July 25, 1983. 

The above exemption was first proposed on July 3, 1964 (29 FR 8431), issued as a final rule on

November 17, 1964 (29 FR 15363), and finally suspended on July 25, 1983 (48 FR 33697). 

Elimination of the exemption for products distributed after that time was proposed on April 30,

1984 (NRC, 49 FR 18308), and issued as a final rule on September 11, 1984 (49 FR 35611). 

Thus, the importation or manufacture of glass enamel or glass enamel frit containing uranium is

no longer permitted in the United States.

Glass enamel and glass enamel frit may contain added uranium, but not thorium. The only

known published analyses of radiological impacts on the public from the use of glass enamel

and glass enamel frit containing uranium are contained in the Federal Register notices cited

above, which first established the exemption and then suspended and eliminated it.

3.14.2 Description of Items

Enamel is glass, and like other types of glass, it is colorless and transparent in its pure and

simple form (Chu and Chu, 1975). To obtain a rainbow of colors, various metal oxides are

added to the glass enamel. To make the colors opaque, other materials such as tin oxide and

kaolin (clay) also are used. Uranium oxide can be added in combination with other materials to

produce colors ranging from off-white or ivory through yellow to golden yellow or amber

(Conrad, 1973). The best estimates of the amount of uranium used in the glass enamel varies

from about 3% by weight for an ivory color to about 9% by weight for a golden yellow color.13

Enamels are made in large blocks of glass that are then crushed and sifted before being sold

as a powder to those engaged in the art of enameling, or the enamel may be sold in the form of

a glass frit to enamelers who grind it into a powder themselves (Chu and Chu, 1975). If an

enamel paste made by mixing powdered enamel and water is packed into very small cells

formed by metal wires or ribbons, and fired at temperatures of up to 850C, the material will

liquefy to fill the cells, then cool to form beautiful, multicolored cloisonne items. Most cloisonne

items imported into the United States are jewelry in the form of pendants, belt buckles, rings,

and earrings, some of which are enameled on both the front and back surfaces.

There is no indication in the literature that thorium was ever used in the making of these various

products.
3–232

3.14.3 Summary of Previous Analyses and Assessments

Radiation doses from routine use of glass enamel and glass enamel frit containing uranium

were estimated by the Atomic Energy Commission (AEC) (29 FR 8431) in establishing the

exemption and by the Nuclear Regulatory Commission (NRC) (48 FR 33697) in suspending the

exemption. External exposure was considered to be the only important exposure mode,

because proper manufacture and use of glass enamel and glass enamel frit ensure that

significant internal exposures are unlikely. The external dose estimates are summarized as

follows.

The AEC stated that the annual dose equivalent to the whole body from photon irradiation of

individuals who would use or be near glass enamel and glass enamel frit containing 10% by

weight of uranium would be a small fraction of the limit of 5millisieverts (mSv) (500 mrem) for

members of the public, the limit that existed at the time the exemption was established. 

However, a quantitative estimate of the dose was not given.

The NRC evaluated doses to the skin from beta particles for individuals wearing cloisonne

jewelry containing up to 7% by weight of uranium in the glass enamel. The maximum dose

equivalent rate to a small area of skin was determined from measurements to be approximately

0.07 mSv/h (7 mrem/h). The annual dose equivalent to the small area of skin was estimated to

be as high as 0.04 Sv (4 rem) if a piece of cloisonne jewelry was worn in contact with the skin

for 520 h/yr (10 h/wk for 52 wk/yr). When not in contact with the skin, the dose to a small area

of skin from wearing the jewelry for 520 hours was estimated to be less than 0.25 mSv

(<25 mrem).

While the use of some jewelry containing uranium did not constitute an immediate or significant

health hazard, the NRC staff concluded that use of the jewelry constituted an unnecessary

exposure to radiation. This view was based on the principle that there should not be any

exposure to manmade radiation without the expectation of some greater benefit resulting from

such exposure. In particular, the use of radioactive material in adornments was of marginal

benefit. Alternatives to the use of uranium were also available that involved no radiation dose

to members of the public (30 FR 3462).

3.14.4 Present Exemption Analysis

Since this exemption was suspended on July 25, 1983 (48 FR 33697) and eliminated on

September 11, 1984 (49 FR 35611), no present exempt analyses were done for distribution and

transport, routine use, and accidents and misuse.
3–233

3.15 Photographic Film, Negatives, and Prints Containing 

Uranium or Thorium

3.15.1 Introduction

In 10 CFR 40.13(c)(3), persons who receive, possess, use, or transfer photographic film,

negatives, and prints containing uranium or thorium are exempted from licensing requirements

for source material, and there is no limit on the amount of uranium or thorium that can be used

in these products. This exemption was established on March 20, 1947 (12 FR 1855), and it has

remained essentially unchanged since that time.

The Federal Register notice cited above provided no information on radiological impacts on the

public from use and disposal of photographic film, negatives, and prints containing uranium or

thorium. Information published by the Atomic Energy Commission in 1960 (25 FR 8619)

indicated that the exemption would not result in an unreasonable hazard to life or property, but

no indication was given that a dose analysis had been performed to support the exemption.

This exemption does not apply to individuals who use uranium or thorium in making or

processing film, negatives, or prints. For example, the use of uranium nitrate as a toner for

photographic prints would be covered under the general license in 10 CFR 40.22 or under a

specific license. However, the receipt, possession, use, or transfer of such photographs is

covered under the exemption granted in 10 CFR 40.13(c)(3).

3.15.2 Description of Item

There is no indication in the literature that thorium has ever been used in the making or

processing of these products.

According to Eder (1945), the light sensitivity of uranium salts was discovered by 

A. F. Geder in 1804, and uranium salts were first used in photography by J. C. Burnett in 1857. 

Burnett, an Englishman, invented a process for printing photographs on paper that was

impregnated with uranium nitrate. Between 1858 and 1860, Niepce de Saint-Victor, a

Frenchman, elaborated on the photographic process of printing on paper by means of uranium,

having the work of Burnett, in 1857, at his disposal (Eder, 1945). It is not clear from the

literature how long this process was used or how extensively it was applied in printing

photographs on paper.

According to Buckley et al. (NUREG/CR–1775), uranium nitrate has been used as a toner for

photographic prints. The uranium nitrate was used in a process as a constituent of a mordant

that was poured or brushed over the surface of black and white prints. The mordant was then

combined with dyes to form an insoluble compound that gave black and white prints the

appearance of being color prints. The toner itself gave the prints a sepia coloration. The

method was employed by both amateur and professional photographers prior to the

development of colored film, but it was not used extensively because it was a difficult, involved

process. Old prints from this process still exist, but there is no evidence that the process is now

used.
3–234

3.15.3 Summary of Previous Analyses and Assessments

There are no known previous analyses of radiological impacts on members of the public

associated with this exemption. Furthermore, the use of uranium or thorium in photography is

not mentioned in either of the reports on consumer products by the National Council on

Radiation Protection and Measurements (NCRP) (NCRP 56, NCRP 95).

3.15.4 Present Exemption Analysis

Although there is no evidence that the process is now being used, old prints from this process

are still in the public possession. Therefore, it is reasonable to determine representative

collective doses. In estimating individual and collective doses to members of the public under

this exemption, the following assumptions are made. First, without basis, only 10,000

photographs exist which were created with this process. Second, it is assumed without basis

that each of the photographs typically contain 1 gram of natural uranium evenly distributed over

a surface measuring 20.3 by 25.4 cm. Third, the photographs are assumed to have an average

lifetime of 50 years and all are at least 30 years.

3.15.4.1 Distribution and Transport

It is assumed that the photographs are shipped from one family member to another every 20

years. It is assumed that photographs are shipped by a parcel delivery system, and a local

parcel delivery driver in a large van picks up the photographs in small lots and takes them to a

local terminal for shipment to other local terminations for delivery to customers. A daily

shipment is assumed to consist of two photographs (i.e., the annual distribution of 500

photographs divided by 250 work days per year). It is further assumed that semi-trucks are

used to transport the photographs between local terminals, and that the photographs pass

through an average of four regional terminals before reaching their final destination. The

radiation dose to workers at both local and regional terminals are assumed to be the same as

those estimated in Appendix A.3 for workers at a large warehouse.

Based on the above assumptions and the generic methodology of Appendix A.3, the individual

receiving the largest dose is the local driver, who is assumed to pick up two photographs only

once in a year. The annual effective dose equivalent (EDE) to this individual is estimated to be

less than 1×105

 mSv (<0.001 mrem). Individual doses to other drivers, terminal workers, and

members of the public along the truck routes would be less. The annual collective EDE to all

parcel-delivery drivers, terminal workers, and members of the public along all truck routes is

estimated to be less than1×105

 person-Sv (<0.001 person-rem).

3.15.4.2 Routine Use

Two different modes of exposure may occur during routine use of the photographs. One mode

is exposure to the whole body to photons from the natural uranium on the photographs. The

other mode is exposure to the skin on the hands to beta particles while handling loose

photographs or photographs stored in albums. However, the beta-particle dose to the skin

should be insignificant for the following reasons: (1) the low surface density of the natural

uranium on the photographs, (2) the fact that one would tend to handle a loose photograph by

the edges or that one would not tend to place their hands directly on top of a photograph in an
3–235

album, and (3) the shielding of stem cells of skin by the thick pads of dead skin on hands,

particularly the fingertips (International Commission on Radiological Protection (ICRP) 59).

To assess this exemption, however, it is assumed that the users are primarily individuals who

want a family photograph, and it is displayed in a prominent place in the home using a picture

frame. For example, the picture frame is placed on a stand next to a chair. If an individual

sitting in the chair was located about 0.5 meter from the photograph, the EDE rate to the whole

body of the individual would be about 0.2 nSv/h (20 nrem/h).

The above dose rate was calculated using MicroShield (Computer Codes, Grove Engineering,

1996). The calculation assumed that (1) the picture frame provided no shielding, (2) the

photograph contained 1 g of natural uranium uniformly distributed over a surface of 20.3 by

25.4 cm, and (3) the natural uranium contained 99.2745, 0.72, and 0.0055% by weight of 238U,

235U, and 234U, respectively (Parrington et al., 1996). The calculation also included photons

from the radioactive progeny of these uranium radionuclides for a decay period of 20 years.

If the individual sits in the chair for an average of 4 h/day (1460 h/yr) while reading or watching

television, the EDE to the individual would be approximately 3×104

 mSv (0.03 mrem). 

Supposing that two individuals are exposed in such a manner to each of the 9,000

photographs, reserving 1,000 for museum use described later, then the collective EDE could be

0.005 person-Sv/yr (0.5 person-rem/yr) or 0.3 person-Sv (30 person-rem) over the assumed

50-year lifetime of the photographs.

To further assess this exemption, it is assumed that 1,000 photographs of this type may be on

display in museums across the United States. In the current assessment it is assumed that a

photograph is in a frame and mounted on a wall. The museum visitor stands 0.5 m from the

display for about 10 minutes. Based on these conditions, the estimated annual EDE from

viewing a photograph would be less than 1×105

 mSv (<0.001 mrem). It is unknown how many

visitors would view an exhibit of this kind. If it is assumed that there are one million people

annually viewing these photographs in museums, the annual collective dose equivalents would

be 3×105

 person-Sv (0.003 person-rem), and over the 50-year lifetime the total collective dose

would be about 0.002 person-Sv (0.2 person-rem).

During routine use, handling, and storage of the photographs, it is assumed that inhalation or

ingestion of uranium in the toner does not occur.

3.15.4.3 Disposal

To estimate potential individual and collective doses to the public from disposal in landfills and

incinerators, the generic disposal methodology in Appendix A.2 was used. It is assumed that

200 photographs are disposed annually containing a total of 200 g of uranium, and it is

assumed that the natural uranium in the toner is not readily inhaled or ingested during collection

and disposal at a landfill. Thus, a reduction by a factor of 10 is assumed in the following doseto-source ratios for inhalation and ingestion in Appendix A.2: (1) waste collectors at both

landfills and incinerators, (2) workers at landfills, and (3) off-site members of the public exposed

to airborne releases during landfill operations.

For disposal at landfills, the annual individual EDE would be about less than 1×105

 mSv

(<0.001 mrem) to waste collectors. The annual individual doses to workers at landfills, off-site
3–236

members of the public, and future on-site residents would be substantially less. The total

collective EDE for a period of 1000 years was found to be about less than 1×105 person-Sv

(<0.001 person-rem).

For disposal by incineration, the annual EDE would be less than 1×105

 mSv (<0.001 mrem) to

waste collectors. The annual individual dose to workers at incinerators and off-site members of

the public are substantially less. The annual total collective EDE is less than 1×105

 mSv

(<0.001 person-rem).

3.15.4.4 Accidents and Misuse

In the case of misuse, a hypothetical scenario is a young child who chews on a photograph and

ingests approximately 10% of the natural uranium on the photograph, or 0.1 g. The amount of

activity in 0.1 g of natural uranium is about 2.6 kilobecquerels (kBq) (0.07 microcurie (Ci)) and

the committed EDE to a 1-year-old child is about three times that to an adult (ICRP 69). Thus,

the EDE to the child based on the dose conversion factors for an adult in Table 3.1.6 of

Section 3.1 is estimated to be approximately 0.6 mSv (60 mrem).

In the case of accidents, a likely scenario is a residential fire that involves the 1 g of natural

uranium on a photograph. Based on a release fraction of 0.1% for the natural uranium on the

photograph and the generic accident methodology of Appendix A.1, the estimates of individual

dose from the residential fire are summarized as follows:

For a person trying to escape from the fire or a neighborhood hero trying to rescue a

person from the fire, the individual dose from a single photograph containing 1 g of

natural uranium could be 4×104

 mSv (0.04 mrem).

For a firefighter wearing a respirator, the individual dose from a single photograph

containing 1 g of natural uranium could be less than 1×105

 mSv (<0.001 mrem). The

individual dose could also be less than 1×105

 mSv (<0.001 mrem) to a worker who is

involved in the cleanup following the fire and does not wear a respirator.

3.15.5 Summary

Table 3.15.1 presents the results of this assessment of potential radiological impacts on the

public from distribution, use, and disposal of photographs containing uranium. These results

are based on a total of 10,000 photographs having an average lifetime of 50 years and

containing 1 g each of natural uranium.

For routine use including distribution and disposal, the most highly exposed individual was

estimated to receive an annual individual dose of 3×104

 mSv (0.03 mrem). The total collective

EDE to the public was estimated to be 0.3 person-Sv (30 person-rem), due almost entirely to

exposure to individuals to a photograph over a 50-year useful life. For misuse and accidents, it

was estimated that the maximum hypothetical EDE could be as much as 6 mSv (60 mrem) for a

child who chews on a photograph and ingests 10% of the natural uranium.
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Table 3.15.1 Potential Radiation Doses From Photographs Containing 

Natural Uranium

Exposure Pathway

Individual Annual Effective

Dose Equivalent

 (mrem)b

Collective Effective

Dose Equivalenta

(person-rem)b

Distribution and transport <0.001c <0.001

Routine use 0.03d 30

Disposal

 Landfills

 Incinerators

<0.001e

<0.001f

<0.001

<0.001

Accidents and misuse

 Ingestion by small child

 Fire

60g

0.04h

a

 Collective doses are based on the existence of 10,000 photographs having an average

lifetime of 50 years and containing 1 g each of natural uranium. Refer to the text discussion for

the time period of collective dose calculations.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 Dose estimate applies to local parcel-delivery driver; dose estimates considerably less for

terminal workers, long-haul semi-truck drivers, and members of public along truck routes are

(see Section 3.15.4.1).

d

 Dose estimate applies to exposure to one photograph during routine home use for 1 year (see

Section 3.15.4.2).

e

 Dose estimate applies to waste collectors at landfills; dose estimates are significantly less for

workers at landfills, off-site members of the public, and future on-site residents at landfills (see

Section 3.15.4.3).

f

 Dose estimate applies to waste collectors at incinerators; dose estimates are significantly less

for workers at incinerators and off-site members of public (see Section 3.15.4.3).

g

 Dose estimate applies to 1-year-old child who chews on photograph and ingests 0.1 g, or

10%, of natural uranium on photograph (see Section 3.15.4.4).

h

 Dose estimate applies to person escaping from a residential fire or neighborhood hero

attempting to rescue person from a residential fire; dose estimates are less for firefighters at

residential fires or workers involved in the cleanup following residential fires (see

Section 3.15.4.4).

3–239

3.16 Finished Tungsten- or Magnesium-Thorium Alloy Products

or Parts

3.16.1 Introduction

In 10 CFR 40.13(c)(4), persons who receive, possess, use, or transfer any finished product or

part fabricated of, or containing, tungsten- or magnesium-thorium alloys are exempted from

licensing requirements for source material, provided the thorium content of the alloy does not

exceed 4% by weight. The exemption does not authorize the chemical, physical, or

metallurgical treatment or processing of any such finished product or part. An exemption for

thoriated tungsten containing not more than 3% by weight of thorium, and without any other

conditions on treatment or processing of the material, was first established on March 15, 1949

(14 FR 1156). The exemption in its present form was proposed on September 7, 1960

(25 FR 8619), and issued as a final rule on January 14, 1961 (26 FR 284).

The Federal Register notices cited above contain little information on the radiological impacts

on the public from use of tungsten- or magnesium-thorium alloy products or parts containing not

more than 4% by weight of thorium. The notice of proposed rulemaking from 1960 states only

that the exemption would not result in an unreasonable hazard to life or property, but additional

information on radiation doses associated with the exemption was not provided. There have

not been any subsequent published dose assessments for this exemption.

3.16.2 Description of Items

This exemption applies to any finished products or parts containing tungsten- or

magnesium-thorium alloys. However, as described below, available information indicates only

limited uses of the alloys subject to this exemption at the present time.

First, information obtained from two manufacturers suggests that essentially all finished

products or parts subject to this exemption that are manufactured and used domestically

contain only magnesium-thorium alloys, but tungsten-thorium alloys have been used rarely, if at

all (Phone call, G. Crawford, Wellman Dynamics Corporation, Creston, IA, January 1995; phone

call, W. Girtz, Hitchcock Industries, Inc., Minneapolis, MN, January 1995). Although a

published report (Hedrick, 1994) indicates that thoriated tungsten has been used in some

products (e.g., welding electrodes and elements used in negative poles of magnetron tubes for

microwave ovens and radar systems), all such products that have been identified are subject to

other exemptions established by the Nuclear Regulatory Commission. These other exemptions

are considered elsewhere in this report.

Second, information obtained from the two manufacturers contacted above, indicates that the

magnesium-thorium alloys subject to this exemption are used almost exclusively in aircraft

parts, particularly parts for aircraft engines. The use of magnesium in aircraft parts is desirable

because of its light weight compared with other suitable metals. The addition of thorium to

magnesium also results in several other desirable properties, including increased hardness,

increased strength, and excellent creep resistance at elevated temperatures (Hedrick, 1985).

Product information obtained from the manufacturers mentioned previously indicates that

magnesium-thorium alloys have been used primarily in castings, particularly in parts for jet
3–240

engines for smaller, executive-sized aircraft, military helicopters, and army tanks. These

castings range in diameter from about 15 to 100 cm and in height from about 10 to 50 cm. 

They weigh about 2 to 75 kg and contain about 1.4 to 2.2% thorium by weight, with the average

thorium content being about 1.7%. Thus, the amount of thorium in the castings typically is

somewhat less than half of the maximum of 4% by weight allowed under this exemption.

Third, consistent with the general decline in uses of thorium by domestic manufacturers over

the past decade (Hedrick, 1985; Hedrick, 1991; Hedrick, 1994), the use of magnesium-thorium

alloys in aircraft parts appears to have declined. From 1973 to 1983, the amount of thorium

oxide used in aerospace alloys was nearly constant, averaging about 4000 to 5000 kg/yr

(Hedrick, 1985). By 1991, however, all metallurgical applications (i.e., including applications

other than in aerospace alloys) consumed only about 500 kg/yr (Hedrick, 1991), and all such

uses were reduced to only about 100 kg/yr by 1993 (Hedrick, 1994). 

Finally, only two domestic manufacturers of magnesium-thorium alloy products or parts

mentioned previously have been identified. Furthermore, these manufacturers have indicated

that they may cease production of these alloys, due in part to decreased demand and the

development of new thorium-free magnesium alloys with similar properties.

3.16.3 Summary of Previous Analyses and Assessments

As noted in Section 3.16.1, there are no published analyses of radiological impacts on the

public associated with this exemption. The statement in the 1960 notice of proposed

rulemaking cited above that says there would not be an unreasonable hazard to life or property

can be interpreted as indicating that doses estimated by the Atomic Energy Commission in

support of granting the exemption did not exceed a small fraction of the existing dose criterion. 

However, the magnitude of the dose estimates cannot be determined on the basis of the

published information.

3.16.4 Present Exemption Analysis

This section presents estimates of dose from routine use of the exempted products or parts

containing magnesium-thorium or tungsten-thorium alloys. These alloys are assumed to be

used in aircraft engine parts because, as noted in Section 3.16.2, this use apparently has been

by far the most common. However, dose estimates for other possible uses should not differ

significantly. Doses from accidents and misuse also are considered. The use of 100 kg/yr is

assumed to be split as 90% consumed in magnesium and 10% consumed in tungsten-thorium

alloys.

For routine uses of the exempted products or parts, external exposure should be the only

pathway of concern. Thorium would not normally be releasable from the alloys and, thus,

inhalation and ingestion exposures should not normally occur.

The dose assessment for routine use is based on (1) measurements of external dose rates

near finished parts containing a magnesium-thorium alloy, (2) information on changes in dose

rates over time due to the buildup of the short-lived, photon-emitting thorium decay products

given in Sections 3.1.3 to 3.1.5, (3) calculations of the dependence of dose rate on the distance

from a source, and (4) consideration of differences in the dose rate for magnesium-and
3–241

tungsten-thorium alloy sources. The dose assessment also assumes, based on information

discussed in Section 3.16.2, that the products or parts subject to this exemption have only

specialized uses, specifically as parts in aircraft. Therefore, exposures to members of the

public, other than workers in the transport and aircraft industries, are assumed to be

insignificant.

This assessment also considers potential doses to firefighters from accidental releases during a

warehouse fire and potential doses to workers and members of the public resulting from

inadvertent introduction of thorium into the scrap metal stream at a steel manufacturing plant. 

In both of these scenarios, exposures would occur primarily by inhalation.

3.16.4.1 Estimated Dose Rates Near Finished Parts

This section describes the estimates of external dose rates near finished parts containing

magnesium- or tungsten-thorium alloys used in estimating dose for routine exposures. These

are based on estimated dose rates near finished magnesium-thorium alloy parts containing

1.7% by weight of thorium.

A potentially important consideration in estimating external dose is the time after chemical

separation of the thorium at which the exposures occur. Information on the dose rate as a

function of time after chemical separation is given in Sections 3.1.3 to 3.1.5. If chemical

separation results in equal initial activity concentrations of 232Th and 228Th, then, for the first few

years after separation, the decrease in dose rate due to decay of the 228Th would be somewhat

compensated by the buildup of activity of the short-lived 232Th decay products. Thus, over the

first 5 years, the dose rate would decrease slightly for about the first 2 years and then increase

slightly, but the variation in the dose rate over this time would not be large. At times beyond

5 years, the dose rate is determined entirely by the buildup of the 232Th decay products. When

the decay products have reached activity equilibrium with the initial activity of the 232Th, which

essentially occurs within about 20 years, the dose rate would be higher than the average dose

rate during the first 5 years after separation by a factor of about 2.5.

The size of routine parts and the size of the largest part reported were considered. MicroShield

(Computer Codes, Grove Engineering, 1996) was used to model the dose rate as a function of

distance from a source that was assumed to be a right-circular cylindrical shell with a height of

15 cm, diameter of 13 cm, and wall thickness of 1.3 cm. The assumed geometry and

dimensions of the source represents a magnesium-thorium alloy casting which would weigh

about 1.6 kg. Table 3.16.1 shows the results of the calculation for the maximum of 4% by

weight allowed under this exemption. MicroShield also was used to model the dose rate as a

function of distance from a source that was assumed to be a thin cylinder with a diameter of

100 cm, and a height of 5.4 cm. The assumed geometry and dimensions of the larger source

represents a magnesium-thorium alloy casting which would weigh about 75 kg.

The information described in the table on estimated dose rates near magnesium-thorium or

tungsten-thorium alloy products or parts, including the dose rate as a function of time after

chemical separation of the thorium and the dose rate as a function of distance from a source, is

used in the dose assessments for distribution and transport and routine use described in the

following two sections.
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3.16.4.2 Distribution and Transport

The generic methodology in Appendix A.3 indicates that, during distribution and transport of

finished products or parts containing magnesium- or tungsten-thorium alloys, the highest

individual and collective doses would be received by truck drivers. Because the products or

parts would not normally be placed in storage for any appreciable length of time before or after

transport to the point of use, doses to distribution workers should be substantially less than

those for truck drivers. In addition, because the products or parts would not normally be sold in

retail stores, doses received by other members of the public during distribution and transport

also would be far less than those received by truck drivers.

Potential doses to individual truck drivers from external exposure to finished magnesiumthorium products or parts—containing the average amount of thorium of 1.7% by weight— were

estimated based on the following assumptions: (1) each shipment would consist of 100 cartons

(200 units) stacked in a cubical array, (2) the average distance between a carton in the first row

and a truck driver would be about 2 meters, and (3) the exposure time for a truck driver would

be about 100 h/yr. The estimated dose rate near a shipping carton presumably applies to

thorium that has recently been chemically separated (2 weeks), and this dose rate should be

appropriate for truck drivers because manufacturers should use relatively fresh thorium and

ship the parts soon after they are produced. The assumed exposure time of 100 h/yr is based

on the rather limited distribution of castings discussed in Section 3.16.2, in which case it would

be unreasonable to assume that an individual truck driver would be exposed for a large fraction

of the working year. With shielding considerations, the effective dose equivalent (EDE) rate to

the driver would be 4×104

 millisievert (mSv)/h (0.04 mrem/h).

Based on these assumptions, the estimate of the EDE to an individual truck driver during

transport of magnesium-thorium alloy castings would be about 0.04 mSv (4 mrem). For

finished tungsten-thorium shipments consisting of 10 tungsten-thorium castings, the estimated

dose would be about 0.01mSv (1 mrem) per shipment. 

The collective dose to truck drivers can be estimated based on the total number of castings

transported per year. As noted in Section 3.16.2, only about 100 kg/yr of thorium are used in all

metallurgical applications at the present time. If 90% of this amount is assumed to be used in

magnesium-thorium alloy castings subject to this exemption and if a typical casting is assumed

to weigh 1.6 kg, as indicated in Section 3.16.4.1, the number of castings containing 1.7% by

weight of thorium transported per year would be 3400. Therefore, based on the dose estimate

for a truck driver during transport of 200 magnesium-thorium alloy castings (100 cartons

containing 2 castings each) given above, the estimated collective EDE to truck drivers would be

about 17 times the estimated individual dose, or about 7×104

 person-Sv (0.07 person-rem).

For tungsten-thorium castings, the number of castings containing 1.7% by weight of thorium

transported per year could be 15. Therefore, based on the dose estimate for a truck driver

during transport of 10 tungsten-thorium alloy castings given above, the collective EDE dose to

truck drivers would be about 1.5 times the estimated individual dose, or about 2×105

 person-Sv

(0.002 person-rem). To estimate the dose rates for the maximum weight percent allowed (4%)

multiply by 2.35.
3–243

3.16.4.3 Routine Use

As noted in Section 3.16.2, finished products or parts subject to this exemption are used

primarily in aircraft engines. Therefore, in this assessment, the individuals receiving the highest

exposures are assumed to be maintenance workers on the engines. Members of the public,

including flight crews and flight attendants, also could be exposed. However, doses to these

individuals should be considerably less than doses to maintenance workers, because of the

much larger source-to-receptor distances and the increased shielding provided by the structure

of aircraft.

Potential doses to maintenance workers on aircraft engines from external exposure to

magnesium-thorium products or parts containing 1.7% of thorium by weight were estimated

based on the following assumptions: (1) exposure is from multiple sources consisting of a large

casting (75 kg containing 1.7% thorium) and ten smaller castings (1.6 kg each and containing

1.7% thorium), (2) the average distance between castings and a maintenance worker would be

about 0.5 m, and (3) the exposure time for a maintenance worker would be about 40 h/yr for the

multiple sources and additional 1000 h/yr for ten smaller castings. 

Calculated dose rates from magnesium-thorium alloy castings are based on 10-year-old

thorium and for the larger casting is about 0.002 mSv/h (0.2 mrem/h). As shown in

Table 3.16.1, the estimated dose rate at 50 cm for a single tungsten-thorium alloy casting

containing 1.7% thorium by weight would be about 4×105

 mSv/h (0.004 mrem/h).

Based on these assumptions, the estimated average annual EDE to an individual maintenance

worker would be 0.5 mSv (50 mrem) for magnesium-thorium alloy castings. As shown in

Table 3.16.1, this EDE would be higher by a factor of 2.35 if the castings were assumed to be

at the maximum allowable content of 4% by weight.

The individual doses calculated above could underestimate actual doses if several of the large

castings were used in a single aircraft engine. However, the estimated doses also should be

somewhat conservative because they do not take into account the shielding that would be

provided by the presence of other metal parts in an aircraft engine. In addition, the assumption

of exposure for an additional 1000 hours near ten smaller castings should be conservative.

Potential doses to maintenance workers working with parts containing tungsten-thorium was

similarly performed but considered exposure to only one smaller casting containing

1.7% thorium. Again, the available information indicates that tungsten-thorium alloy products or

parts subject to this exemption have not been used to a significant extent. Extrapolated from

Table 3.16.1, the estimated dose rate at 50 cm for a single tungsten-thorium alloy casting

containing 1.7% thorium by weight would be about 2×104

 mSv/h (0.02 mrem/h). Allowing a

single worker to be exposed to 1 tungsten-thorium casting for 1000 hours, the EDE would be

0.2 mSv (20 mrem). As shown in Table 3.16.1, this EDE would also be higher by a factor of

2.35 if the castings were assumed to be at the maximum allowable content of 4% by weight.

The estimate of collective dose is the summation of individual doses over the life of the product

for as many products as are assumed to be distributed in one year. The collective dose to

aircraft engine maintenance workers can be estimated as follows. First, it is assumed that

90 kg/yr of thorium is distributed under this exemption (see Section 3.16.4.2). Second, it is
3–244

assumed that the useful lifetime of the castings is ten years. Based on these assumptions and

the individual dose estimates for the exposure given above, the estimated collective EDE to

maintenance workers would be about 1 person-Sv/yr (100 person-rem/yr). A similar collective

EDE result would be obtained for tungsten-thorium alloy castings.

3.16.4.4 Disposal

The condition for this exemption as noted in Section 3.16.1 is that chemical, physical, or

metallurgical treatment or processing of the products or parts are not authorized. However,

since they are exempt from licensing, the possibility exists that these products or parts could be

disposed of in landfills. To provide a bounding evaluation, it is hypothetically assumed that

100% of the thorium annually used is disposed of as ordinary waste. To estimate potential

doses from disposal of the products or parts as ordinary waste in landfills, the generic disposal

methodology in Appendix A.2 is used. It is further assumed that this disposal is limited to 100

landfills instead of the default 3,500 of Appendix A. The annual EDE to a waste hauler would

be 0.007 mSv (0.7 mrem) and the estimate of collective dose from landfill disposal would be

0.7 person-Sv (70 person-rem).

3.16.4.5 Accidents and Misuse

This assessment also considered two scenarios for accidents and misuse of alloys subject to

this exemption. The first scenario involves a fire in a warehouse, and the second involves

inadvertent introduction of thorium into the generic scrap metal stream at a steel manufacturing

plant.

These scenarios represent a hypothetical maximum for the total amount of thorium to be

involved in a fire or introduced into steel. An average quantity was not determined and is also

speculative.

3.16.4.5.1 Warehouse Fire

In the first scenario, a fire is assumed to occur in a warehouse containing 170 parts which is

10% of the assumed annual production, each of which weighs an average of 1.6 kg and

contains 1.7% by weight of thorium. Thus, the warehouse is assumed to contain 2.7 kg of

thorium. Based on the generic methodology in Appendix A.1, the EDE to a firefighter using

respiratory protection during the fire would be 0.001 mSv (0.1 mrem), and the EDE to an

individual who is not using respiratory protection during cleanup after the fire would be

0.01 mSv (1 mrem). 

3.16.4.5.2 Introduction Into Steel

In the second scenario, the thorium in finished products or parts subject to this exemption is

assumed to be inadvertently introduced into the scrap metal stream at a steel manufacturing

plant. The dose assessment for this scenario is based on the generic methodology for

recycling in Appendix A.2. Potential individual doses to workers at a smelter and members of

the public residing near a smelter are considered.
3–245

In this assessment, 1 kg of thorium is assumed to be introduced into steel per year. This

should be a conservative estimate of the potential for misuse as it represents 1% of the

assumed annual production.

Based on the assumed amount of thorium recycled per year and the generic methodology for

recycling in Appendix A.2, the estimates of individual dose are summarized as follows:

For individual workers at a smelter, the annual EDE would be 5×105

 mSv

(0.005 mrem).

For off-site members of the public, the individual annual EDE would be less than

1×105

 mSv (<0.001 mrem).

3.16.5 Summary

Persons receiving any finished product or part fabricated of, or containing, tungsten-thorium or

magnesium-thorium alloys are exempt from licensing requirements for source material,

provided the thorium content of the alloy does not exceed 4% by weight. This assessment has

considered doses to the public from use and disposal of finished tungsten-thorium or

magnesium-thorium alloy products or parts containing the average amount of thorium of 1.7%

by weight which is representative of the material in actual use. The results of the assessment

are presented in Table 3.16.2 and may be summarized as follows:

Estimated doses to individuals during transport and routine use of exempted products or

parts containing magnesium-thorium alloy are a few mrem/yr. 

Estimated collective doses during routine use of exempted products or parts containing

magnesium-thorium alloy, based on an assumption of 100 kg/yr of thorium distributed

under this exemption and a useful lifetime of 10 years for the finished products or parts,

are 1 person-Sv/yr (100 person-rem/yr). This estimate probably is conservative,

because the current distribution of thorium under this exemption is likely to be less than

100 kg/yr. The collective dose during distribution and transport would be about

7×104 person-Sv (0.07 person-rem).

Estimated collective doses to workers at landfills from disposal of exempted parts or

products containing magnesium-thorium alloy are about 0.7 person-Sv (70 person-rem). 

Many of the dose estimates in Table 3.16.2 are based on assumptions about the total amount

of thorium involved for the exposure scenarios. Based on available information about the

current distribution of thorium in the exempted parts, these assumptions should be

conservative.

This assessment also considered doses resulting from a fire in a warehouse and the

inadvertent introduction of exempted products or parts into the scrap metal stream at a steel

manufacturing plant. Potential doses from a warehouse fire, either during the fire or cleanup

after the fire, appear to be low, due to the small amount of thorium that normally would be

stored in a warehouse and the assumptions that a firefighter would use respiratory protection

and that only a small fraction of the thorium involved in a fire would be available for inhalation or
3–246

ingestion during or after a fire. Similarly, potential individual doses due to the inadvertent

introduction of thorium into steel appear to be low.
3–247

Table 3.16.1 Comparison of Materials, Ages, and Dose Rates With Distances

Material Age Shield Distance from Carton

Dose Ratea

(mrem/h)b

Mg-4% Th 2 wk Box material 5 cm 0.23

W-4% Th 2 wk Box material 5 cm 0.89

Mg-4% Th 10 yr Box material 5 cm 0.28

W-4% Th 10 yr Box material 5 cm 1.0

Mg-4% Th 2 wk Box material 200 cm 0.00075

W-4% Th 2 wk Box material 200 cm 0.0026

Mg-4% Th 10 yr Air 50 cm 0.01

W-4% Th 10 yr Box material 50 cm 0.041

a

 To obtain dose rates for 1.7% by weight (the average thorium content), divide by 2.35.

b

 1 mrem = 0.01 mSv
3–248

Table 3.16.2 Summary of Potential Radiation Doses From Use of Finished 

Tungsten- or Magnesium-Thorium Alloy Products or Parts a

Exposure Scenario

Annual Individual Effective

Dose Equivalent

(mrem)c

Annual Collective Effective

Dose Equivalentb

(person-rem)c

Distribution and transportd

 Magnesium-thorium

 Tungsten-thorium

4e

1e

0.07

 0.002

Routine usef

 Magnesium-thorium

 Tungsten-thorium

 50

 20

100g

included above

Disposal in landfills 0.7h 70h

Accidents or misuse

 Warehouse firei

 Processing in scrapk

 1j

 0.005l

a

 Dose estimates are based on assumption that all finished products or parts contain the

amount of thorium of 1.7% by weight.

b

 Dose estimates are based on assumption that 100 kg of thorium is distributed per year under

this exemption. Based on data on current distribution of thorium in aerospace alloys,

assumption probably is conservative, especially for tungsten-thorium alloys.

c

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

d

 Dose estimates apply to truck drivers; doses to other members of the public during distribution

and transport would be substantially less (see Section 3.16.4.2).

e Dose estimate is based on assumption that 200 finished products or parts are transported per

shipment for magnesium- thorium alloy and 10 for tungsten-thorium alloy. (see

Sections 3.16.4.1 and 3.16.4.2).

f

 Dose estimates apply to maintenance workers on aircraft engines; doses to other members of

the public during routine use would be considerably less.

g

 Collective dose to maintenance workers on aircraft engines from 1 year’s distribution,

assuming 10-year useful lifetime for finished products or parts; no other significant uses of

exempted products or parts have been identified.

h

 The primary exposure pathway is direct exposure.

i

 Dose estimates are based on the assumption that 2.7 kg of thorium is contained in the

warehouse.

j

 Dose estimate is total dose during cleanup after fire without use of respiratory protection;

estimated dose to firefighter using respiratory protection is somewhat less.

k

 Dose estimates are based on the assumption that 1 kg/yr of thorium is inadvertently

introduced into scrap metal stream.

l

 Dose estimate applies to workers at smelter; estimate of individual dose for off-site members

of the public is considerably less.
3–249

3.17 Uranium in Counterweights

3.17.1 Introduction

In 10 CFR 40.13(c)(5), counterweights containing uranium installed in aircraft, rockets,

projectiles, and missiles or stored or handled in connection with installation or removal of such

counterweights are exempted from licensing requirements for source material. There is no limit

on the amount of uranium that can be contained in each counterweight. The exemption for

aircraft counterweights installed in aircraft was first established on July 8, 1960 (25 FR 6427). 

A proposal to extend the exemption for aircraft counterweights to include storage, installation,

removal, and incidental handling was published on August 9, 1961 (26 FR 7143), and a final

rule was issued on November 22, 1961 (26 FR 10929). The exemption was extended to

include uranium counterweights used in rockets, projectiles, and missiles, as well as in aircraft,

on December 29, 1962 (27 FR 12914). The present exemption, issued on September 9, 1969

(34 FR 14067), does not authorize the chemical, physical, or metallurgical treatment or

processing of any such counterweights other than repair or restoration of any plating or other

covering.

Published information on radiological impacts on the public from use of uranium counterweights

in aircraft, rockets, projectiles, and missiles includes the first Federal Register notice from 1961

(26 FR 7143). The present exemption was based on the belief (1) that manufacturing

techniques provide adequate protection against oxidation of uranium and (2) authorized

activities do not involve exposure hazards significantly different from those involved in handling

an undamaged counterweight. However, a recent report to the U.S. Air Force by a contractor

(Starmet, 1998) provides insight into contamination levels of corroded depleted uranium (DU). 

There also are events reported to the Nuclear Regulatory Commission (NRC) involving potential

contamination of maintenance workers while removing or cleaning corroded DU counterweights

(Nuclear Materials Events Database (NMED) ((NRC, Databases, NMED, Search))). 

Additionally, two unpublished occupational dosimetry studies were conducted by one of the

aircraft manufacturers, and there is exposure rate measurement data associated with varying

types of counterweights. For this assessment, annual individual and collective effective dose

equivalents (EDEs) were calculated for routine exposure to a flight crew and passengers and

for truck drivers, as well as for individuals potentially impacted by counterweight salvage,

storage, and disposal. EDEs were also calculated for selected accident scenarios.

3.17.2 Description of Items

The exemption does not specify the type of uranium that can be used; however, DU appears to

be the only type of uranium that has been used in counterweights. Counterweights or ballasts

made of DU are used to balance hinge points and control surfaces (rudders, stabilizers,

ailerons, and elevators) of aircraft. They are used to aid in hydraulic adjustments during flight. 

DU has also been used for military ballast and counterweight applications.

DU is used to take advantage of its high density (uranium is 1.6 times as dense as lead), since

counterweights are subject to space limitations. DU differs from natural uranium in that it

contains less 235U and 234U. The DU specific activity (SpA) used by Nuclear Metals, Inc.

(currently known as Starmet CMI) (1993), in its product transferring guidelines is

1.3×104

 becquerel (Bq)/g (3.6×107 curie (Ci)/g), which equates to a 235U content of about
3–250

0.20% and a 234U content of about 0.0005%. The external exposure rate from DU materials

depends on the physical form and containment but does not depend significantly on its

chemical form (Lemons, 1990).

3.17.2.1 Aircraft

DU counterweights have been used primarily in wide-body aircraft on rudders, outboard

ailerons (wing assembly), and outboard elevators (tail assembly). Counterweights come in a

variety of weights and shapes, and numerous weights and shapes are used in some aircraft. 

The DU counterweight can range in weight from 0.23 to 77 kg (Phone call, R. Toole, Carolina

Metals, Inc. (a subsidiary of Nuclear Metals, Inc.), Barnwell, SC, September 1994). Information

related to the use of DU counterweights in commercial aircraft are summarized in Table 3.17.1.

Some aircraft used for military and cargo also use DU counterweights; these include the 

Lockheed C–130 and C-141, Jetstar, and S–3A. The Boeing Company produced helicopters

utilizing DU as a rotor tip weight prior to 1979. These weights consisted of small 0.22-kg

triangular weights. One to three weights were installed per blade. Virtually all of the Boeing

helicopters manufactured prior to 1979 have had their blades replaced with composite blades

that do not contain DU weights.

3.17.2.2 Missiles and Rockets

DU counterweights may be used as anti-flutter devices in missile test applications. However, in

discussions with various organizations, it was unknown whether DU counterweights were used

for this purpose in any tactical missiles. DU counterbalances have been or are used in the

Trident missile program (Phone call, R. Nickell, EG&G, Kennedy Space Center, Cape

Canaveral, FL, September 1994).

DU counterweights are not used in the space shuttle systems or in the payloads, e.g., satellite

systems. However, DU counterweights were used during the Apollo space program (Phone

call, R. Nickell, EG&G, Kennedy Space Center, Cape Canaveral, FL, September 1994).

3.17.2.3 Production Information

The National Lead Company was the primary U.S. manufacturer of DU counterweights;

however, this company is no longer in business and stopped manufacturing new

counterweights in 1986 (Phone call, D. Barbour, Project Manager, Depleted Uranium Programs,

Philotechnics, Oak Ridge TN, October 1999). A company called Eldorado Resources Ltd.

(currently known as Cameco Corporation), located in Canada, is the only known North

American manufacturer of DU counterweights for nonmilitary applications (Phone call,

E. Lanchester, Quality Assurance Department, Nuclear Metals, Inc., Concord, MA, September

1994). Starmet CMI (formerly Nuclear Metals, Inc.) is the only refurbishing facility in the United

States (Phone call, E. Lanchester, Quality Assurance Department, Nuclear Metals, Inc.,

Concord, MA, September 1994).

In many cases, tungsten alloy counterweights have replaced DU counterweights in aircraft. 

Since 1981, The Boeing Company has provided customers with tungsten replacement

counterweights, and tungsten counterweights have been installed in new Boeing 747 aircraft. 

Tungsten equivalents have been sent as spares since 1981 (Gallacher, 1994). In 1988,
3–251

McDonnell-Douglas discontinued using DU counterweights and began using tungsten. 

Tungsten counterweights are used on the MD–11, MD–80, and MD–90 (Ford, 1994).

It is unknown how many DU counterweights are currently installed in aircraft. It is estimated

that approximately 15,000 weights may be associated with the Boeing 747 fleet (based on 550,

Boeing 747 aircraft produced between 1968 and 1981 and spare parts) (Gallagher, 1994). 

However, the number of aircraft that contain DU counterweights is decreasing. Rather than

refurbishing the DU (during maintenance operations), tungsten counterweights are used as a

replacement (Phone call, D. Barbour, Project Manager, Depleted Uranium Programs,

Philotechnics, Oak Ridge, TN, October 1999).

There are two changes in the pattern of distribution and use which implies that the demand for

DU counterweights has essentially disappeared: 

• During the past decade, replacing DU with tungsten as the counterweight material is

reported to be increasing. The decrease in use of DU is reflected by the activity in the

industry. Starmet is the only Federal Aviation Administration (FAA) approved facility in

the United States licensed to repair DU aircraft counterweights. Starmet refurbishes

counterweights in about 10 to 14 C-141 aircraft per year, about 100 weights per plane

ranging from 0.1 to 70 kg. About 50 counterweights are refurbished per month by

Starmet for their remaining customers. On rare occasions, Starmet is requested to

manufacture a new counterweight (Phone call, Don King, Starmet CMI, Concord, MA,

April 2001).

• As the wide-body planes in the operational fleet are being retired from service, over 100

planes in 1999, the counterweights are being removed and stored. Quantities of several

tons are being held by operators, parts suppliers, and tear-down facilities (Barbour,

2000a).

3.17.3 Summary of Previous Analyses and Assessments

The Atomic Energy Commission (AEC) provided information on radiological impacts on

members of the public from use of uranium counterweights in aircraft, rockets, projectiles, and

missiles in the first Federal Register notice from 1961 (26 FR 7143). Doses from external

exposure were estimated by the AEC using measured exposure rates at the surface of a

counterweight and film badge data and handling time studies for workers. Measured dose

rates in air from beta and photon radiation were about 1300 Gy/h (130 mrad)/h), of which the

photon component was only about 30 Gy/h (3 mrad/h). On the basis of its analysis, the AEC

concluded the following:

The annual dose equivalent from external exposure to the hands probably would not

exceed 0.05 sievert (Sv) (5 rem), which is less than 10% of the dose limit for the hands

and forearms of radiation workers of 0.75 Sv (75 rem) that had been established in

10 CFR 20 (based on the 1961 version).

The annual dose equivalent from external exposure to the whole body is unlikely to

exceed a small fraction of the dose limit for members of the public of 5 mSv (500 mrem)

that had been established in 10 CFR 20 (based on the 1961 version).
3–252

The National Lead Company measured beta- and gamma-dose rates at various distances from

(1) a bare and nickel-cadmium plated 645-cm2 flat surface of DU and (2) a bare DU and a

nickel-cadmium typical” counterweight (Michael, 1965). The Boeing Company conducted two

dosimetric studies of exposures to workers involved with DU counterweight installation into

aircraft (Gallacher, 1994). The Boeing Company also measured exposure rates attributed to a

15-kg DU counterweight. Results of these measurements and studies are summarized in the

following subsections.

3.17.3.1 National Lead Study

Two sets of dose rate measurements were made by the National Lead Company (Michel,

1965). The first set of measurements, as shown in Table 3.17.2, were considered to be

representative of dose rates for DU counterweights in the 32 to 454 kg range. These dose

rates were measured at various distances from a 645-cm2

 flat surface. The second set of

measurements, also shown in Table 3.17.2, provided dose rates representative of a typical”

2.5×103

 cm nickel-cadmium plated counterweight.

3.17.3.2 Boeing Studies

The following subsections describe the two dosimetric studies The Boeing Company conducted

of exposures to aircraft workers involved with DU counterweight installation (Gallacher, 1994). 

The periods of the studies were December 1968 to February 1970 and September 1977 to April

1978.

3.17.3.2.1 Counterweight Installation

The first Boeing dosimetry study involved four employees responsible for the installation of DU

counterweights. There were 27 separate monthly readings. The second Boeing study involved

37 employees, including personnel responsible for installation, storage, and transport of DU

counterweights. Under routine handling operations, the DU counterweights, which arrive in

crates, are taken to an interim storage facility. The crated DU counterweights remain in storage

until requested for installation. Depending on scheduled operations, there may be an inventory

of such crated DU counterweights. Once requested, the DU counterweights are transported to

the installation site. At the installation site, the DU counterweights are removed from the crates

and physically riveted into the appropriate location by the installation personnel (Phone call,

E. Edwards, The Boeing Company, Seattle, WA, August 1996). Associated with these job

activities were 296 monthly dose reports. Table 3.17.3 gives the measured doses to workers

for each of the handling activities (Gallacher, 1994).

According to Gallacher (1994), both studies showed all worker whole-body doses were less

than the detection limit of the devices used. The doses were less than 2.6% of the exposure

limits for occupationally exposed individuals (i.e., 0.05 Sv/yr (5 rem/yr)). Because of the

limitation of the technology used, The Boeing Company was unable to prove from these studies

that doses to workers were less than the 1 mSv (100 mrem) annual EDE limit for individual

members of the public. In addition, the second study demonstrated that all worker extremity

doses were typically less than 1.88 mSv (<188 mrem) per quarter (less than 1% of the

extremity dose limits) for occupationally exposed employees (Gallacher, 1994).
3–253

3.17.3.2.2 Flight Operations

The Boeing Company also measured dose rates associated with a 15-kg counterweight that

resulted in slightly lower readings than the National Lead study. It was determined that a 2-mm

aluminum shield (roughly equivalent to aircraft skin) yields a fourfold reduction in gamma-dose

rates. If a source-to-receptor distance between the flight staff and the DU counterweights was

8 meters, and the exposure duration was 2000 hours, then the estimated annual dose for the

flight staff was conservatively estimated to be 0.016 mSv (1.6 mrem) (Gallacher, 1994).

3.17.3.2.3 Transport

Dose rates at the surface of packages in transport were reported to be between 0.002 and

0.07 mSv/h (0.2 and 7 mrem/h) and at 1 meter from the package at background to be

0.005 mSv/h (0.5 mrem/h) (Gallacher, 1994). Nominal values for spare parts shipments were

0.03 mSv/h (3 mrem/h) at the package surface and 0.003 mSv/h (0.3 mrem/h) at 1 meter

(Gallacher, 1994). There was no information related to the number of counterweights or

quantity of DU contained in these packages.

3.17.3.3 U.S. Air Force Studies

The U.S. Air Force is an NRC specific licensee and monitors all activities with radioactive

material including material authorized for exempt distribution. The following summarizes some

of the monitoring information and describes activities which may parallel the commercial

industry. 

3.17.3.3.1 Starmet Report 

A recent report for the U.S. Air Force by a contractor (Starmet, 1998) provides insight into

contamination levels of excessively corroded DU counterweights. The reported contaminated

levels averaged greater than 200,000 dpm/100 cm2

. 

3.17.3.3.2 Event Notifications 

There are three events reported to the NRC by the U.S. Air Force regarding potential

contamination of maintenance workers while removing or cleaning corroded DU counterweights

(NMED Item Numbers 940856, 970387, and 990519 (NRC, Databases, NMED, Search)). 

These and other events reported to the NRC are often speculative regarding dose assessment;

however, there has never been a U.S. Air Force event involving DU counterweights where there

was any dose above (or even approaching) the occupational regulatory limit nor did any event

have any significant activity detected in a urinalysis. The report and notifications describe

drilling out or pressing out sheared bolts/screws, increasing bolt hole locations, using a hammer

and chisel to remove the DU, and cutting wing parts. The following is provided regarding event

notifications. 

• The NMED relates potential contamination of personnel due to cutting wing parts away

from DU counterweights. This is an operation common to commercial salvage activities

(Item Number 940856 (NRC, Databases, NMED, Search)). 
3–254

• A licensee reported that four individuals were potentially exposed to DU when they

attempted to use chemical cleaner to degrease a painted counterweight from which

some paint was flaking. No airborne contamination was detected. Licensee

calculations determined that none of the workers would have received an uptake in

excess of 1 ALI for 238U due to this event (Event Notification 970387).

• A licensee reported a possible overexposure of an employee who inhaled DU dust while

inappropriately working on counterweights. This event was later determined not to be

reportable. Licensee personnel were performing maintenance on a C-141 cargo aircraft

aileron using unauthorized activities that produced dust and debris. Surveys revealed

that contamination levels in the room where the maintenance was being performed were

above background. Bioassay results showed that the workers received committed

EDEs of less than 10 Sv (<1 mrem) (Event Notification 990519). 

3.17.4 Present Exemption Analysis

The current exemption was designed in part on the NRC belief in 1969 that ..(1) experience to

date with thousands of counterweights in use over the past several years indicates that current

manufacturing techniques provide adequate protection against oxidation of uranium, and

(2) activities which would involve processing of uranium are expressly prohibited, except for

processes which do not involve exposure hazards significantly different from those involved in

handling an undamaged counterweight. The Commission considers that the provisions in the

amendments adequately control the low radiation exposures that may result from discarded

counterweights.” (34 CFR 14067). Since the pattern of distribution and use of DU

counterweights has changed over the past 30 years, this analysis examines transportation and

distribution, routine use (flight operations), maintenance (with corroded counterweights or those

with missing plating), storage, disposal, and accidents or misuse.

The individual and collective EDEs for workers involved with the installation and removal of

counterweights in aircraft and for a truck driver, aircraft crew, attendants, and passengers are

estimated using MicroShield (Computer Codes, Grove Engineering, 1996).

When using MicroShield (Computer Codes, Grove Engineering, 1996) to calculate dose rates,

the DU counter weights were modeled with a coating that consisted of nickel and cadmium with

plating thicknesses of 2.5×103

 cm each, which reflects the minimum plating thickness

requirements of the industry.

3.17.4.1 Transport and Distribution

Shipments of counterweights and spare parts are transported by ground freight or air cargo. In

air shipments, DU counterweights are shipped as Cargo Aircraft Only” (Gallacher, 1994). If it

is assumed that two 15-kg counterweights are shipped as spare parts, and the distance

between the freight truck driver and the counterweight package is 1.4 meters (Etnier and

O’Donnell, 1979), the estimated EDE rate would be 4×105

 mSv/h (0.004 mrem/h), accounting

for counterweight coatings and 0.5 cm of metal shielding from the truck structure. Since these

counterweights are typically refurbished in Barnwell, SC, the maximum exposure duration from

transport of DU counterweights between the west coast of the United States and Barnwell is

assumed to be about 90 hours (round trip). Based on these assumptions, the resulting EDE is
3–255

estimated to be 0.004 mSv (0.4 mrem). Assuming 600 counterweights are refurbished per

year, representing the upper bound reported for 1994 (Phone call, E. Lanchester, Quality

Assurance Department, Nuclear Metals Inc, Concord, MA, 1994.) then the collective EDE to

truck drivers is estimated to be 0.001 person-Sv (0.1 person-rem).

3.17.4.2 Routine Use

3.17.4.2.1 Aircraft Counterweight Maintenance

Based on The Boeing Company experience from modification work conducted on the

Boeing 747 fleet in 1976, about 10% of then existing aircraft with DU counterweights still

installed may exhibit corrosion. The Boeing Company’s survey of this degradation indicated

that the typical damage area ranged from 1 to 50% of the exposed surface (Gallacher, 1994). 

Gallacher (1994) also reported contamination levels up to 1500 dpm per 100 cm2

; however,

contamination levels as high as 250,000 dpm per 100 cm2

 were identified in the U.S. Air Force

report (Starmet, 1998). 

No current data is available on the percentage of counterweights exhibiting corrosion. It is

assumed that only 5% of the counterweights are encountered per year at the 250,000 dpm per

100 cm2

 level (Starmet, 1998) and the remainder are contaminated at 1500 dpm per 100 cm2

(Gallacher, 1994). For this evaluation, it is assumed that a worker performs 60 maintenance 

activities each year on counterweights; two of the maintenance activities involve weights with

contamination levels of 250,000 dpm per 100 cm2. and the remainder with levels of 1500 dpm

per 100 cm2

. A resuspension factor of 1×10–5 m–1 from Appendix A.1 is assumed. For a two

hour exposure time to remove a counterweight, the estimated annual EDE from inhalation is

0.8 mSv (80 mrem). If it is further assumed that contamination of the hands is limited to 10% of

the maximum level and that a person may ingest all the contamination from 10 cm2

 of skin over

a 24-hour period (International Atomic Energy Agency (IAEA) SS No. 7), the estimated EDE is

0.01 mSv (1 mrem).

Both Boeing studies, as discussed in Section 3.17.3.2, showed that all occupational whole-body

doses were below the detection limit for the dosimeters. Based on calculations using

MicroShield, the dose rate to a worker exposed to a 15-kg counterweight at a distance of 30 cm

plus two additional counterweights at a distance of 1 meter would be 4×104

 mSv/h

(0.04 mrem/h). Assuming a maintenance worker is exposed for 4 hours per removal/installation

and performs 60 such activities each year, the committed EDE would be 0.1 mSv (10 mrem). 

For this scenario, 0.9 mSv (90 mrem) represents the maximum dose equivalent to a worker

during normal maintenance. Assuming that half of the 430 aircraft listed in Table 3.17.1 still

contain DU counterweights and that 5% of them require maintenance each year, the annual

collective EDE is estimated to be 0.01 person-Sv (1 person-rem).

These estimated doses do not include a dispersibility factor for any cutting or grinding. The

potential intake might be increased by as much as a factor of 10 if maintenance includes these

activities (NUREG–1400). 
3–256

3.17.4.2.2 Flight Operations

3.17.4.2.2.1 Individual Dose Estimates

To estimate annual EDEs to flight personnel and passengers at typical source-to-receptor

distances, it is assumed there is about 300 kg of DU located in the outboard elevator (tail

assembly) and 350 kg of DU located in each aircraft wing. Shielding provided by the plane skin

thickness is two layers of aluminum approximately 0.2 cm each and one 0.5 cm layer of plastic. 

The resulting annual individual EDEs for flight crew, attendants, and passengers are

summarized in Table 3.17.4. This table also describes assumed distances from the DU and the

geometry used in calculations.

3.17.4.2.2.2 Collective Doses

Using the individual dose estimates shown in Table 3.17.4, the collective EDEs to the flight

crew, flight attendants, and passenger population are summarized in Tables 3.17.5 and 3.17.6,

respectively. It is assumed the aircraft crew and attendants spend 1000 h/yr in flight on these

aircraft. For flight attendants, the exposure duration is estimated to be half the time in flight. It

is assumed a member of the public flies at least 1000 miles one way per year, and this flight is

estimated to take at least 3 hours one way or 6 hours round trip (Phone call, S. Russell, Air

Transport Association (ATA), Washington, DC, August 1996). Though the annual EDE is low,

about 1×10-4 mSv (0.01 mrem), the annual collective EDE of about 3 person-Sv (300 personrem) results because of the large number of people flying. However, these individual and

collective EDEs need to be put in perspective relative to the annual EDE from cosmic radiation. 

For example, an annual mean EDE to flight crews from cosmic radiation is approximately

5 mSv (500 mrem) (Paretzke and Heinrich, 1993; Friedberg, 1993), which is 250 times greater

than the estimated annual EDE to flight attendants due to exposure to DU counterweights.

3.17.4.2.3 Storage

Barbour (2000b) reported that there are 1,743 DU counterweights of several sizes and shapes

for sale by various companies. He cautions that this number does not reflect the total in

stockpile as the demand is now very low and they are probably not being advertised. Assuming

an average weight of 15 kg per counterweight for the aircraft described in Table 3.17.1, the

reported stockpile would be enough to refit more than 25% of the 215 aircraft that still may

contain DU counterweights.

Barbour (2000b) also reported that at least 18 companies listed the counterweights for sale. If

the stockpile is located in not more than 18 locations, not more than 100 counterweights would

be stored in each location. Based on these assumptions, an assumed average distance of

10 m from the worker to the storage location, a calculated dose rate of 6×105

 mSv/h

(6×103

 mrem/h) and an assumed exposure time of 500 hours per year, a warehouse worker

could receive an annual external dose of 3×104

 mSv (0.03 mrem). The estimated collective

dose to warehouse workers is 2×105

 person-Sv (0.002 person-rem). This estimate is probably

conservative as it does not account for stacking and selfshielding of counterweights. 
3–257

3.17.4.3 Disposal

Barbour reports that there is a changing pattern of distribution and use of counterweights

including the growing activity of salvage of overaged aircraft (Barbour, 2000a). There are

several instances in the NRC’s NMED (NRC, Databases, NMED, Search) involving the

activation of scrap yard portal monitors by DU counterweights. As all landfills are not equipped

with radiation detectors and some DU counterweights might not be detected, the possibility

exists that some of these counterweights are disposed of in landfills.

Since DU counterweights are produced using DU metal with high impurity values and are plated

with nickel, cadmium, and at times with chromium, direct remelting of the weights into shielding

products has not been successful (Andersen, 1996). One refurbisher, Starmet CMI (formerly

Nuclear Metals, Inc.), routinely disposes of the weights at licensed facilities (Andersen, 1996).

It is further assumed that DU counterweights are not sent to municipal incinerators, since DU

pyrophoric properties would result in potentially hazardous operations.

3.17.4.3.1 Landfill Disposal

Since it is possible that DU counterweights could be disposed in a landfill, it is hypothetically

assumed that 10% of the annually refurbished counterweights, or 60, are inadvertently

disposed as industrial waste. This assumption is also representative of the average amount of

DU on one plane as listed in Table 3.17.1.

Further, it is assumed that an individual waste hauler would pick-up and deliver for disposal no

more than 6 counterweights with 2 counterweights per haul. For an exposure time of 4 hours

per haul (see Appendix A.2) and a dose rate of 2×105

 mSv/yr (0.002 mrem/yr) (see

Section 3.17.4.1), the EDE to a waste hauler would be 2×104

 mSv/yr (0.02 mrem/yr.

For the disposal of 60 counterweights, the annual collective dose would be 2×106

 person-Sv

(2×104

 person-rem).

Another likely exposure pathway is off-site ingestion of contaminated groundwater. For the offsite receptors that ingest groundwater near landfills, it is assumed that 60 15-kg counterweights

are disposed annually for 30 years in landfills. Based on these assumptions, the annual

individual and collective EDE to an off-site receptor is less than 1×105

 mSv (<0.001 mrem) and

0.006 person-Sv (0.6 person-rem), respectively.

3.17.4.3.2 Recycle Operations

The generic disposal methodology was used to estimate the annual individual EDE to slag

workers and off-site receptors due to airborne releases. As reported above, there are 19 cases

from NRC’s NMED (NRC, Databases, NMED, Search) involving the activation of scrap yard

portal monitors by DU counterweights. That counterweights would be directed to scrap yards is

a reasonable expectation because they are associated with relatively high value aluminum

scrap,...” (Barbour, 2001). An NRC Daily Event, Number 37781 (NRC, Databases, Daily Event)

for February 27, 2001, documents a recent instance of disposal involving 118,000 pounds of

aluminum. For this assessment, it is assumed that 60 15-kg counterweights are recycled

annually at 60 smelters. The dose-to-source ratios (DSRs) from Table A.2.15 are multiplied by
3–258

the ratio 100/60 to reflect the limited number of items relative to the number of assumed

smelters (100). The annual EDE to a slag worker is estimated to be 0.02 mSv (2 mrem). The

estimated annual individual EDE to an off-site individual from airborne release is 2×105

 mSv

(0.002 mrem).

3.17.4.4 Accidents and Misuse

Three primary accident scenarios have been considered pertaining to DU counterweights:

(1) aircraft accident involving fire, (2) storage facility fire, and (3) loss and misuse of material. 

Impacts resulting from these scenarios are listed below.

3.17.4.4.1 Aircraft Accident Involving Fire

In the event of a fire involving DU counterweights the generic modeling of Appendix A.1 has

been used, where it has been assumed that the transportation accident would reasonably

approximate the exposures from an aircraft fire. The amount of material involved is 850 kg of

DU, which is that on a typical Boeing 747 aircraft. For the fireman, the EDE is estimated to be

0.3 mSv (30 mrem) and for the clean-up worker, the EDE is estimated to be 4 mSv (400 mrem).

Additionally, NRC’s NMED (NRC, Databases, NMED, Search) shows a report for one C-141

cargo plane that caught fire and burned on the tarmac. The coatings on the counterweights

were examined after the fire was extinguished and there was no breach of the protective

coating on any counterweight.

3.17.4.4.2 Storage Facility Accident Involving Fire

According to Gallacher (1994), a fire in a storage facility containing spare depleted

counterweights is considered to have an extremely low probability of occurrence. Only one

structural fire of this type had occurred in 26 years (Gallacher, 1994). However, if a fire

occurred in a storage facility (as per the accident methodology cited in Appendix A.1), that

contained 60 15-kg DU counterweights, the resulting EDE to firefighters is estimated to be

0.06 mSv (6 mrem). Since tungsten weights are now being used as spares, the likelihood of

DU being present, particularly in such an amount, is greatly reduced. During cleanup

operations, the EDE due to resuspension is estimated to be 0.4 mSv (40 mrem).

3.17.4.4.3 Loss and Misuse of Depleted Uranium Counterweights

Only two incidents have occurred at The Boeing Company since 1968 in which counterweights

have been lost. In the first incident, the counterweight was removed from a storeroom

(Gallacher, 1994). The counterweight was recovered and corrective action was taken to further

reduce the number of people with access to the area. In the second incident, three

counterweights were removed from aircraft as part of a replacement program prior to their loss. 

Corrective action was taken at the time of each event to improve access control and

accountability, with the result that no repetitions have been reported.

That DU counterweights are being identified at landfills is evidence that they are being lost. 

Due to the high density of DU as compared to lead, DU counterweights are reportedly used to

make bucking bars” to set rivets or trimming weights for racing car chassis. Assuming a 14-kg

DU counterweight is misused for undefined reasons, the dose rate would be about 0.5 Gy/h
3–259

(0.05 mrad/h) at a distance of 1 meter. Further assuming that an individual is exposed for

500 hours at a distance of 1 meter, an upper bound estimated annual dose is 0.2 mSv

(20 mrem). A larger potential dose is estimated if the DU counterweight is significantly

corroded or if the DU is sawed or cut creating airborne particulates. Using the same scenario

to obtain a DU counterweight from an airplane wing as described above, with a contamination

level of 250,000 dpm per 100 cm2

, the committed EDE to an individual from inhalation of

resuspended material during a half hour period is estimated to be about 0.8 mSv (80 mrem). 

The estimated dose includes a dispersibility factor of 10 for any cutting or grinding

(NUREG–1400). The 1 hour used here for the cutting and grinding versus the actual time

duration for the work is speculative. 

3.17.5 Summary

Table 3.17.8 summarizes the results of this current analysis of radiological impacts on the

workers and members of the public from distribution, use, misuse and disposal of DU

counterweights. Where possible actual measurements and dosimetry data were used to

estimate individual and collective doses. 

There is a change in the pattern of distribution and use which implies that the demand for DU

counterweights has essentially disappeared. As the wide-body planes in the operational fleet

are being retired from service, the counterweights are being removed and stored. The annual

dose equivalent to warehouse workers is 3×104

 mSv (0.03 mrem). The estimated collective

dose to warehouse workers is 2×105

 person-Sv (0.002 person-rem).

The annual EDE to a freight truck driver making one round trip from the west coast to South

Carolina was estimated to be 0.004 mSv (0.4 mrem). The annual dose equivalent to workers

responsible for installation, storage, and transport of DU counterweights was 0.09 mSv/yr

(90 mrem/yr). The estimated collective EDE to airline maintenance workers is 0.01 person-Sv

(1 person-rem).

Annual estimated individual EDEs to flight crews, attendants, and passengers were 0.001 mSv

(0.1 mrem), 0.01 mSv (1 mrem), and 1×104

 mSv (0.01 mrem), respectively. The annual

collective EDEs to flight crews and attendants was estimated to be 0.04 person-Sv

(4 person-rem). The annual collective EDE for domestic airline passengers is 3 person-Sv

(300 person-rem). If DU counterweights are located only in the tail section, the EDEs will

decrease (see Table 3.16.6). The doses to the flight personnel and passengers attributed to

the DU counterweights are less than 0.001 of their doses from cosmic radiation.

A misuse scenario was developed with hypothetical assumptions regarding the maximum

reported contamination levels and the separation of the DU counterweight from an aircraft wing

by sawing. The committed EDE to a salvage worker is estimated to be about 0.8 mSv

(80 mrem).
3–260

Table 3.17.1 Use of DU Counterweights in Domestic Aircraft

Aircraft

Type Manufacturer

Aircraft Number

Owned by

Domestic

Carriersa,b

Total Weight of

Counterweights

per Aircraft

(kg)

DC–10 McDonnell-

 Douglas

 168 ~ 1,000c

L–1011 Lockheed 60 ~ 680c

B–747 Boeing 202 ~ 850d

a

 Number of aircraft owned by domestic carriers that are members of the Air Transport

Association (ATA, 1999).

b

 A reasonable estimate is that 50% of these aircraft still contain DU counterweights. Phone

call J. Taylor, Starmet CMI, Barnwell, SC, August 1999.

c

 Phone call, E. Lanchester, Quality Assurance Department, Nuclear Metals, Inc., Concord, MA,

August 1996.

d

 Phone call, D. Barbour, Project Manager, Depleted Uranium Programs, Philotechnics, Oak

Ridge, TN, October 1999.
3–261

Table 3.17.2 Exposure Rates for DU Counterweights a

Distance Beta- and Gamma

Exposure Rate

(mR/h)c

Gamma

Exposure Rateb

(mR/h)c

Applicable to 32 to 454 kg U

 Surface

 15 cm

 31 cm

 

220

22

7

3.0 (at 1 cm)

0.9

0.5

Typical of nickel-cadmium

 plated (0.001 inch) 

 counterweight

 15 cm

 31 cm

 

3.5

0.66

 

0.06

0.02

a Michel, 1995.

b

 Plate thickness is 0.004 cm of cadmium.

c

 1 mR/h = 258 nC/kg-h.
3–262

Table 3.17.3 Installation of DU Counterweights in Aircraft—

Occupational Dosimetry Study a

Operation Dose Equivalent per Month

Installation and

 handling

Whole body: Average <0.1 mSv (10 mrem)

Maximum 0.2 mSv (20 mrem); 1 occurrence—due

 to installation and removal of parts

Extremity: All readings below detection 0.3 mSv (30 mrem)

Distribution Whole body: All readings below detection 0.1 mSv (10 mrem)

Extremity: All readings below detection 0.3 mSv (30 mrem)

Storage Whole body: All readings below detection 0.1 mSv (10 mrem)

Extremity: All readings below detection 0.3 mSv (30 mrem)

Transport Whole body: All readings below detection 0.1 mSv (10 mrem)

Extremity: 2 recordable exposures—

 1 mSv (100 mrem) (shipping clerk)

 0.5 mSv (50 mrem) (store clerk)

a

 Gallacher, 1994.
3–263

Table 3.17.4 Estimated Annual Effective Dose Equivalents—Wing and Tail Assembly

Exposed

Group

Counterweight

Location on

Aircrafta

Source-toReceptor

Distance

(m)

Dose Rate

(mrem/h)b

Exposure

Durationc

(h)

Annual

Individual

Effective

Dose

Equivalent

(mrem)b

Flight crew Wings

 Tail

32

60

1×104

2×105

1,000

1,000

0.1

0.02

Flight

 attendant

 Wings

 Tail

12

8

8×104

2×103

500

500

0.4

1

Passengersd Wings

 Tail

12

8

8×104

2×103

3

3

0.002

0.006

a

 For tail assembly, assume source with 300 kg of DU decayed for 25 years; dimensions are

hypothetical 4×20×200 cm. For wing assembly, assume 350 kg in each wing of DU decayed

for 25 years; dimensions are hypothetical 4×20×234 cm. The number of DU counterweights is

different in each plane and type. These dimensions were selected to represent the total mass

of DU that could be expected.

b

 1 mrem = 0.01 mSv.

c

 Flight crew and passengers are assumed to be stationary. For the flight attendant, the value

is the exposure duration from either the wings or the tail during the flight. The flight attendant is

actually in flight for a total of 1,000 hours.

d

 The average distance of a domestic flight is about 1,609 km one way, which equates to a

3-hour flight. It is assumed that on average a member of the U.S. population spends 6 hours

(round trip) in flight per year.
3–264

Table 3.17.5 Collective Effective Dose Equivalents for Aircraft Crew and Flight Attenda

Aircraft

Type

Aircraft

Numbera

Number of Crew

and Attendantsb

Annual Collective

Effective Dose

Equivalent, Aircraft

Crew

(person-rem)c

Annual Collective

Effective Dose

Equivalent, Flight

Attendants

(person-rem)c

A

E

DC–10 168 3 and 5 0.06 1 L–1011 60 3 and 6 0.02 0.5 B–747d 202 3 and 8 0.07

(0.01 tail only)

2

(2 tail only)

Total 430 1,290 and 2,816 0.2 

(0.01 tail only 747)

4

(2 tail only)

a ATA, 1999.b Phone call, W. Edmunds, Airline Pilots Association, Washington, DC, 1994.c 1 person-rem = 0.01 person-Sv.d The effective dose equivalent (EDE) for the aircraft crew and attendants is calculated assuming DU counterwe

wing and tail assembly. In the parentheses, the EDEs assume that the DU counterweights are located only in t
3–265

Table 3.17.6 Collective Effective Dose Equivalent Estimates for Aircraft Passengers

Aircraft

Type

Number

of

Aircrafta

Passenger

Number

per

Flighta,b

Number of

Nonconnecting

Departuresb

Annual

Passenger

Number

 Individual

Annual

Effective

Dose

Equivalent

(mrem)c

DC–10 168 194 7.5×104 1.5×107 0.01 

L–1011 60 208 4.1×104 8.5×106 0.01 B–747d 202 300 4.9×104 1.5×107 0.01 (0.006) 

Total 430 702 3.9×107 

a ATA, 1999; however, it is further assumed that 25% of these aircraft have replaced DU counterweights with tu

counterweights.

b Phone calls, S. Russell, Statistics, ATA, August/September 1996. The passenger number per flight accounts 

loading factor of 67%. The number of nonconnecting departures is estimated by multiplying the number of depof departures include flight connections).

c 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.d The annual collective effective dose equivalents (EDEs) for the passengers are calculated assuming DU coun

wing and tail assembly. In the parentheses, the EDEs assume that the DU counterweights are located only in ttail assembly.
3–266

Table 3.17.7 Summary of Exposure Rates Attributed to 33-lb Counterweights a

Source-to-Receptor 

Distance

 (m)

Beta and Gamma 

Exposure Rates

 (mR/h)b

Gamma Only

 Exposure Rates 

(mR/h)b

0.3 7.0 0.5

1 0.05 0.05

6 0.0013 0.0013

8 0.0008 0.0008

a

 Gallacher,1994. Based on the National Lead Company Measurements given in Table 3.17.2.

b

 1 mR/h = 258 nC/kg-h.
3–267

Table 3.17.8 Individual and Collective Effective Dose Equivalent Summary

Exposure Pathway

Annual Effective Dose

Equivalent to Individuals

(mrem)a

Collective Effective Dose

Equivalent to Population

(person-rem)a

Transport

 Driver 0.4 0.1

Routine use

 Installation and removal

 Flight operations

 -Crew

 -Attendant

 -Passenger

 Storage

 90

 0.1

 1

0.01

0.03

 1

 0.2 (0.01 tail only 747)

 4 (2 tail only 747) 

300 (200 tail only 747)

 0.002

Landfill disposal

 Waste hauler

 Off-site receptor

Recycle operations

 Slag worker

 Off-site receptor

 0.02

 <0.001

 2

 0.002

 2×104

 0.6

 NAb

NA

Accidents and misuse

 Aircraft incident involving fire

 Storage facility involving fire

 -Firefighter

 -Cleanup

 Loss/misuse

 30c

 6

 40

 80d

NA

NA

NA

NA

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Not applicable.

c

 Dose to firefighter. Dose to clean-up worker is estimated to be 4 mSv (400 mrem).

d

 Per NMED, Item Number 940856 (NRC, Databases, NMED, Search), relates potential

contamination of personnel due to cutting wing parts away from DU counterweights. This is an

operation common to commercial salvage activities.

3–269

3.18 Uranium Shielding in Shipping Containers

3.18.1 Introduction

In 10 CFR 40.13 (c)(6) it states that natural or depleted uranium used as shielding in any

shipping container is exempted from licensing requirements for source material, provided that

the uranium metal is encased in mild steel or an equally fire-resistant metal of thickness at least

one-eighth of an inch (3.2 mm). This exemption was proposed on August 9, 1961

(26 FR 7143), and issued as a final rule on November 22, 1961 (26 FR 10929). The Federal

Register notices cited above do not contain information directly addressing radiological impacts

on members of the public from use of uranium as shielding in shipping containers.

This exemption was intended to relieve byproduct material licensees from also obtaining a

source material license for containers used to ship the byproduct material. These licensees are

not relieved of the 10 CFR 20 requirements, but in practice, the control of exposures is

dependent on the proper use of the container as shielding. Though the uses of the containers

are exempt, the filling and emptying of the transported products are generally done at a

licensed facility (with the doses to workers controlled by occupational radiation protection

requirements). This exemption is unique in that the life after initial distribution involves

exposures associated with licensee facilities.

3.18.2 Description of Items

Depleted uranium (DU) is used as radiation shielding in shipping containers since it is readily

available and has a high density, good radiation absorption efficiency, mechanical strength, and

a high melting temperature (Derrington et al., 1994). Because of its high density,

uranium-shielded shipping containers can be smaller and lighter than containers using lead or

steel shielding with equivalent radiation absorption capabilities. DU alloys are currently used for

gamma-ray shielding in containers designed for the storage, transport, and disposal of

high-level radioactive wastes or spent nuclear fuel (SNF) as well as for the transport of other

gamma-ray sources, such as radiography sources.

There are several different designs of shielded containers for high-level radioactive waste and

SNF. Frequently these designs include large cylinders of DU for gamma radiation shielding

(Derrington et al., 1994). About 15 SNF casks currently use DU for shielding. Of these, there

are six or seven different designs (Phone call, L. Shappert, Transportation Technology Group,

Chemical Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN, August 1994). 

Some of the designs are oriented toward transporting large quantities of disposable radioactive

waste. Table 3.18.1 lists some of the shipping casks that use DU for shielding. Oak Ridge

National Laboratory (ORNL), a major producer of 192 Ir, owns 11 Department of Transportation

(DOT) Specification 20 WC–1 Type B casks, which are used to ship 192Ir to customers who

produce 192Ir radiography devices (Phone call, S. McGhee, Packaging Operations, Oak Ridge

National Laboratory, Oak Ridge, TN, July 1997). Approximately 60 kg of DU is used in each of

these casks. ORNL also uses a customer-owned cask for shipment of 192Ir that contains about

95 kg of DU.

The majority of shipping containers that use DU as shields are for the transport of radiography

sources, which include gamma-ray projectors and source changers. The actual number of DU
3–270

shipping containers currently used in the United States for radiographic exposure devices and

source changers is not available; however, one of the major suppliers provided data on the

number of their products manufactured or in use (Okvist, 1994). Listed in Tables 3.18.2 and

3.18.3 are examples of Nuclear Regulatory Commission (NRC)-licensed packages used for

radiographic exposure devices and source changers. The shielding used in the shipping

container” is the same as the shielding during use of radiography devices. However, this

exemption only covers use of the shipping container” as shielding during transport, not during

use of the device, particularly since these devices are used by NRC or Agreement State

licensees.

3.18.3 Summary of Previous Assessments

There are no known radiological dose assessments on the distribution, transport, and disposal

of shipping containers that use uranium as shielding. Regulations do exist for empty

radioactive materials packaging (49 CFR 173.428, which refers to 49 CFR 173.421). These

regulations state that the radiation level at any point on the external surface of the package

cannot exceed 0.005 millisievert (mSv)/h (0.5 mrem/h) (49 CFR 173.421 (a)(2)) and that the

packages are in unimpaired condition (40 CFR 173.428(b)) to be excepted from the shipping

paper and labeling requirements. If radiation levels exceed this rate, the DOT shipping and

labeling requirements must be met.

Dose rate measurements have been made at contact and at 1 meter from shipping casks used

to transfer 192Ir from production facilities to source fabrication facilities. Exposure rate

measurements also have been made at contact and at 1 meter from a bare DU metal sheet as

well as at varying thicknesses of stainless steel shielding over the DU metal sheet. Dosimetry

data were also obtained for workers handling (fabricating) radiographic containers storing DU. 

Results of these measurements and dosimetry data are summarized below.

ORNL owns 11 DOT Specification 20 WC–1 Type B casks which consist of a wooden overpack

with an inner package. The inner package or cask contains about 60 kg of DU. The measured

dose rates at the surface of the overpack with an empty cask ranged from 0.004 to

0.007 mSv/h (0.4 to 0.7 mrem/h), and at 1 meter, dose rates were less than or equal to

0.001 mSv/h (0.1 mrem/h) (Phone call, S. McGhee, Packaging Operations, Oak Ridge National

Laboratory, Oak Ridge, TN, July 1997). For the cask owned by a customer, a dose rate on the

surface of the overpack (which is a steel cage) of an empty cask was about 0.007 mSv

(0.7 mrem/h), and at 1 meter was about 0.002 mSv/h (0.2 mrem/h). Exposure rate

measurements were made at contact and at 1 meter from a bare depleted uranium sheet with

dimensions 71 cm × 210 cm × 0.6 cm (Manufacturing Sciences Corp., 1993). Additional

measurements were made with 0.14 cm, 0.26 cm, and 0.4 cm thicknesses of stainless steel

placed over the depleted uranium sheet. The results of these measurements are shown in

Table 3.18.4.

Information was obtained from the Amersham Corporation concerning potential exposures to

workers handling (fabricating) DU shielded radiographic device containers. Film badge reports

and DU logs were consulted to obtain these exposure results and are summarized in

Table 3.18.5. Based on this dosimetry information, workers handling these containers

(Model 660) either in manufacturing or handling of DU shielding for radiographic devices may

obtain an average dose rate of about 0.008 mSv/h (0.8 mrem/h). The measured dose rate at
3–271

1 meter from an empty Model 660 radiographic device was 0.001 mSv/h (0.1 mrem/h)

(Okvist, 1994).

3.18.4 Present Exemption Analysis

Shipping containers are used in the transport of nuclear material, therefore, transport and

possibly disposal are the primary applications of this exemption. This exemption does not

include the use of the container as shielding during device use, but does include the shielding in

these devices when used for shipment. To evaluate the situation where a large number of DU

shielded containers are used in the United States, the present exemption analysis focuses on

the distribution and transport of 192Ir for industrial uses. This includes (1) transport of 192Ir from

the production facility in DU shielded casks to the source fabrication facility and (2) transport of

radiographic exposure devices and source changers from the source fabrication facility to the

industrial user. The following subsections summarize the potential doses to workers and

members of the public from transporting, distributing, and disposing of these containers.

3.18.4.1 Distribution and Transport

To provide an example of the potential doses associated with the distribution and transport of

shipping containers that contain DU, the shipment of 192Ir for industrial uses was evaluated. 

There are two major transport phases: (1) shipment of 192Ir from the production facility to the

source fabrication facility and (2) shipment of the radiographic exposure devices and source

changers from the source fabrication facility to the industrial user (e.g., nondestructive testing

and well logging). The mode of shipment of 192 Ir from the production facility to the source

fabrication facility is at the customer’s discretion (Phone call, S. McGhee, Packaging

Operations, Oak Ridge National Laboratory, Oak Ridge, TN, July 1997). Based on past

activities, about 80% of these shipments are by air and the remainder (20%) are transported by

motor freight (Phone call, M. Ferren, Isotope Distribution, Chemical Technology Division, Oak

Ridge National Laboratory, Oak Ridge, TN, July 1997). Shipment from the source fabrication

facility to the industrial user is primarily (about 90%) by overnight air cargo. 

The measured dose rates, as described in Section 3.18.3, for shipping casks used by ORNL

are not inconsistent with calculations performed using MicroShield (Computer Codes, Grove

Engineering, 1996). However, dosimetry data, as shown in Table 3.18.5, for a radiographic

device container are higher than those calculated with MicroShield (Computer Codes, Grove

Engineering, 1996). Because the difference could not be explained with the data provided,

estimated dose rates calculated with MicroShield were used as a basis to estimate doses

associated with the handling and transport of radiographic devices from the source fabrication

facility to the industrial user. Source-to-receptor distances and selected exposure durations

used in the generic distribution methodology (see Appendix A.3) are used in the present

analysis. When user information was available, this was used in the current analysis. 

Table 3.18.6 summarize the annual individual effective dose equivalents (EDEs) attributed to

shipping containers that contain DU that are used for the distribution and transport of 192Ir for

industrial uses.

To estimate the annual collective EDEs, industry sources were contacted to suggest the

number of annual shipments. About 60 to 120 cask shipments of 192Ir are made annually from

production facilities to source fabrication facilities (Phone call, S. McGhee, Packaging
3–272

Operations, Oak Ridge National Laboratory, Oak Ridge, TN, July 1997; phone call, M. Ferren,

Isotope Distribution, Chemical Technology Division, Oak Ridge National Laboratory, Oak Ridge,

TN, July 1997). According to other industry sources, there may be 2000 shipments or more

each year of 192Ir radiographic devices and source changers. These shipment numbers are

representative of a major percentage of the shipments made in the United States. Based on

this information, the estimated collective annual EDEs from the distribution and transport of

shipping containers shielded with DU are also summarized in Table 3.18.6. 

3.18.4.2 Disposal

Based on available information, DU shipping containers are not expected to be disposed in

municipal sanitary landfills. Typically DU is removed from the shipping container and sent to

Nuclear Metals, Inc., Concord, MA, (currently known as Starmet CMI) for reuse and recycling

(Okvist, 1994). In some cases, source changers may be used as source storage and therefore

could be reused numerous times. However, since these containers are exempt from licensing,

it is possible that a fraction of these empty radiographic devices and source changers could be

discarded in municipal landfills. It is unknown how many of these devices could be discarded. 

Using the generic disposal methodology (see Appendix A.2), it is assumed 3500 devices each

containing 16 kg of DU were discarded in 1 year to municipal landfills (one device disposed

annually in a single landfill). The estimated annual individual and collective EDEs attributed to

disposal are summarized in Table 3.18.7. 

3.18.4.3 Accidents

The accident scenario considered for the present analysis for DU shielded containers is a

transportation accident in which a fire ensues. The generic accident methodology, as described

in Appendix A.1, is used in the present analysis. Two separate transportation accident

scenarios are evaluated: (1) an accident involving a shipping cask (60 kg DU) transporting 192Ir

from the production facility to the source fabrication facility and (2) an accident involving a

shipment of a radiographic device (16 kg of DU). In both cases, it is assumed that the shipping

package has been breached and the DU is exposed and has caught fire. For the transportation

accident involving a shipping cask that contains 60 kg of DU, the estimated individual EDEs are

0.02 mSv (2 mrem) to a firefighter and 0.3 mSv (30 mrem) to a worker involved in cleanup

following the fire. For the transportation accident involving a radiographic device in a shipping

cask that contains 16 kg of DU, the estimated individual EDEs are 0.006 mSv (0.6 mrem) to a

firefighter and 0.07 mSv (7 mrem) to a worker involved in the cleanup following the fire. The

estimated individual EDEs to a firefighter and a cleanup worker from the DU may be minor

compared to the EDEs from the transported materials (e.g., 192Ir radiographic exposure

devices).

3.18.5 Summary

Table 3.18.7 summarizes the individual and collective EDEs to workers and members of the

public from the transport and disposal from DU shielded shipping containers. The primary

exposure pathway is direct irradiation. Actual measurement and dosimetry data were used,

when applicable, to estimate individual and collective doses. To consider the broad use of DU

shielded containers, the distribution and transport of 192Ir in casks and radiography devices were

evaluated. These containers are not typically disposed in municipal landfills, they are more
3–273

likely to be reused as source storage or recycled. However, if these containers were disposed,

the condition where one container was disposed annually in each municipal landfill was

evaluated. Based on this analysis, the following general conclusions about radiological impacts

on the public associated with the exemption can be obtained:

 Air transport is the primary mode of transport for DU shielded shipping containers. 

Usually these shipments are made by overnight air carriers, thereby reducing potential

exposures to members of the public. The air-freight delivery truck drivers and the

loaders are the workers most likely to receive the higher annual doses as compared to

other transportation workers.

 The potential exposures resulting from the DU in the shipping container are negligible

compared to the potential exposures from the byproduct material or accelerator

produced material they are designed to shield.
3–274

Table 3.18.1 Shipping Cask Characteristics a

Cask No. Applicant

Depleted

Uranium

Thickness

(inches)b

Dimensions

(inches)b

Cask

Capacity

(pounds)b

FSV–1 Public Service of Colorado 2.25 46,025

IF–300 Pacific Nuclear Systems, 

Inc.

 4 Diameter 64

Length 210

140,000

NLI–1/2 Nuclear Assurance Corp. 2.75 Outside

 diameter 47

Length 95

 49,250

NLI–10/24 Nuclear Assurance Corp. Railcar

 Outside

 diameter 96

 Length 204

194,000

NLI–6502 Nuclear Assurance Corp. DU shielding

 angles inner

 region max.

 5.75 inches

 thick

Diameter 33

Length 130

GE–100 General Electric Co. DU liners may

 be inserted,

 not integral

a

 NUREG–0383.

b

 1 inch = 2.54 cm; 1 pound = 0.454 kg.
3–275

Table 3.18.2 Radiographic Exposure Devices a

Model No.

DU Shield

Weight

(pounds)b

DU Shield

Dimensions;

Radius, Length

(inches)b

Maximum Container

Activity

(curie (Ci))b

Number

in Usec

192Ir devices

460

520 (AL)

660

683

702

865

900

920

22

35

28

40

28

31

1.38, 7.44

1.64, 8.5

120

Varies

120

 10,000

240

120

240

 0

4,138

 28

 36

 49

60Co Devices

664

676

680

370

284

3.7, 12.63

3.9, 13.40

330

110

 33

 181 

192Ir/60Co devices

684

741

150

200 3.25, 11.44

11 (60Co); 240 (192Ir) 

33 (60Co); 240 (192Ir)

 104

 74

a

 NUREG–0383.

b

 1 inch = 2.54 cm; 1 pound = 0.454 kg; 1 Ci = 3.7×1010 Bq. c

 Okvist, 1994a.
3–276

Table 3.18.3 Source Changers a

Model No.

DU Shield 

Weight

(pounds)b

Maximum Container

Activity

(Ci)b Number in Usec

C–8 60Co: 200 1

AL 500 SU 39 192Ir: 120 132

650 40 192Ir: 240 333

C–1 192Ir: 240 0

771 213 60Co: 110 5

820 120 192Ir: 1,000

 (<240 Ci/source)

 8

770 355 60Co: 550 1

855 125 192Ir: 1,000

 (<240 Ci/source)

 3

864 192Ir: 360 9

850 48 192Ir: 240 16

C–10 65 192Ir 66

U–110 60 192Ir 29

a

 NUREG–0383.

b

 1 pound = 0.454 kg; 1 Ci = 3.7×1010 Bq. c

 Okvist, 1994a.
3–277

Table 3.18.4 Exposure Rate Measurement of Depleted Uranium Sheet with Shielding

Exposure Rate Measurements a

Material

Beta/gamma

(mR/h)b

Gamma

(mR/h)b

Beta/gamma

(mR/h)b

Gamma

(mR/h)b

AT CONTACT AT 1 METER

Bare Depleted

Uranium

40 6 20 1

Depleted

Uranium with

0.14 cm thick

stainless steel

shield

5 4.5 1 0.8

Depleted

Uranium with

0.26 cm thick

stainless steel

shield

4 4 0.8 0.6

Depleted

Uranium with

0.4 cm thick

stainless steel

shield

3 3 0.65 0.6

a

 Measurements taken by Manufacturing Sciences Corp., 1993.

b

 1 milliroentgen (mR)/h = 258 nanocoulomb (nC)/h.
3–278

Table 3.18.5 Dosimetry Data From Assembly of Radiographic Device Container a

Condition of Exposure

Averaged Dose Equivalent

(mrem/h)b

Average body dose rate working with DU 0.8

Average extremity dose rate working with DU 2

Model 660 device dose rate at surface 2

Model 660 device dose rate at 1 meter 0.1

a

 Okvist, 1994.

b

 1mrem = 0.01 mSv.
3–279

Table 3.18.6 Annual Effective Dose Equivalents From the Distribution

and Transport of Shipping Packages Containing Depleted Uranium

Exposure Pathway

Annual Individual Effective

Dose Equivalent

(mrem)a

Annual Collective Effective

Dose Equivalent

(person-rem)a

Production facility to

 source fabrication

 facility

 Motor freightb

 Loader

 Driver

 Air transportc

 Driver

 - to airport from

 production facility

 - from airport to

 industrial user

 Loader

 Aircraft crew

 

0.1

0.3

0.2

0.06

0.2

6×103

5×104

6×104

5×104

9×104

 4×105

Source fabrication

 facility to industrial

 user

 

 Motor freightd

 Loader

 Driver

 Air transporte

 Driver

 - pickup from

 fabrication facility to

 airport

 - drop off from airport

 to industrial user

 Loader

 Aircraft crew 

0.5

0.6

2.3

0.4

5

0.06

0.002

0.006

0.01

0.1

0.001

See following page for footnotes.
3–280

Footnotes to Table 3.18.5

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Twenty percent of cask shipments (20 out of 100 shipments) are assumed to be transported

by motor freight. The source-to-receptor distance for the loader is 1 meter and for the driver,

3 meters. The amount of DU in the shipping cask is about 60 kg. It is assumed that each of

five loaders handles four shipments per year and is exposed for 2 hours per shipment. It is

assumed that it takes 3 days (24 hours) for one driver to transport one shipment to the source

fabrication facility, and that this driver makes 10 trips in 1 year. The estimate of the collective

effective dose equivalent (EDE) assumes a total of 20 shipments per year.

c

 For the production facility to source fabrication facility air transport scenario, it is assumed that

there are 80 shipments per year and one aircraft crew transports 20 casks per year. Two air

freight pickup drivers share a route in which each transports 40 casks per year. It is assumed

that each of five aircraft loaders handles 16 casks per year. The amount of DU in a single cask

is assumed to be 60 kg. The exposure duration for the drivers is 2 hours per shipment, 1 hour

for the loader per cask, and 3 hours per flight for the flight crew. For the collective EDE

estimate, it is assumed there are two aircraft flight members and a total of 80 shipments. 

d

 For the source fabrication facility to industrial user motor freight transport scenario, it is

assumed there are 200 shipments per year. The source-to-receptor distance for the loader is

1 meter and for the driver, 3 meters. The exposure duration for the driver is 24 hours per

shipment; for the loader, it is 1 hour per shipment. It is assumed that a driver makes 20 trips

per year and a loader handles 40 devices per year. The annual collective EDE calculation

assumes 200 shipments annually.

e For the source fabrication facility to industrial user air freight transport scenario, it is assumed

there are 1,800 shipments annually. Ten planes are used annually to transport the devices,

equating to 180 devices being transported per plane per year. Five air freight drivers take the

device shipments to the airport and 30 drivers make drop-off shipments to the industrial users. 

The source-to-receptor distances are 1 meter for the loader, 2 meters for the air freight drivers,

and 11 meters for the flight crew. The annual collective EDE estimates assumed 1,800

shipments per year by air and two aircraft crew members per shipment.
3–281

Table 3.18.7 Effective Dose Equivalent Estimates for the Disposal of Shipping

Packages Containing Depleted Uranium a

30-Year Disposal

Receptor

Annual 

Individual

Effective Dose

Equivalent

(mrem)b

Annual

Collective

Effective Dose

Equivalent

(person-rem)b

Individual

Effective Dose

Equivalent

(mrem)b

Collective

Effective

Dose

Equivalentb

(person-rem)b

Workerc

 Collector

 Landfill operator

0.5

0.04

2

0.7

On-sited 0.1 200

Off-sitee

 Groundwater 5×105 0.4

a

 See Section 3.18.4.2 for the description of the source term used in this analysis. Refer to

Appendix A.2 for further information on the generic disposal methodology.

b

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

c

 The primary exposure pathway is direct irradiation; it is assumed there is no inhalation or soil

exposure pathway. For the workers’ collective effective dose equivalents, one collector and five

landfill operators are assumed to be associated with approximately 3,500 active municipal

landfills.

d

 On-site receptors: assume 10 on-site receptors exposed to 30 years of waste and residing 

on-site 30 years after landfill closure. The total on-site population is assumed to be about 

35,000.

e

 Off-site receptor: there is no airborne exposure pathway. Groundwater is the primary

exposure pathway for this exposure scenario.
3–282

Table 3.18.8 Individual and Collective Effective Dose Equivalents Summary 

for Shipping Packages Containing Depleted Uranium

Exposure Pathway

Annual Individual

Effective Dose Equivalent

(mrem)a

Annual Collective Effective

Dose Equivalent

(person-rem)a

Distribution and Transportb

 

Production facility to source

 fabrication facility

 

 Motor freight

 - Loader

 - Driver

 Air transport

 - Delivery truck drivers

 Sending

 Receipt

 - Loader

 - Aircraft crew

 Source fabrication facility

 to industrial user

 Motor freight

 - Loader

 - Driver

 Air transport

 - Delivery truck 

 Pickup

 Dropoff

 - Loader

 - Aircraft crew

0.1

0.3

0.2

0.06

0.2

0.006

0.5

0.6

2

0.4

5

0.06

 

5×104

6×104

5×104

9×104

4×105

0.002

0.006

0.01

0.1

0.001

Disposalc

 Workers

 - Collector

 - Landfill operator

 On-site receptors

 Off-site receptors

 - Groundwater

0.5

0.04

0.1

5×105

 

2

0.7

200

0.4

See end of table for footnotes.
3–283

Table 3.18.8 Individual and Collective Effective Dose Equivalents Summary 

for Shipping Packages Containing Depleted Uranium (continued)

Exposure Pathway

Annual Individual

Effective Dose Equivalent

(mrem)a

Annual Collective

Effective Dose Equivalent

(person-rem)a

Transportation accidentd

 Cask transport

 - Firefighter

 - Cleanup worker

 Radiographic devices

 - Firefighter

 - Cleanup worker

2

30

0.6

7

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv. Values are rounded to one significant

figure.

b

 Refer to Section 3.18.4.1 for additional information on the distribution and transport of

shipping containers.

c

 Refer to Section 3.18.4.2 for additional information on the disposal of shipping containers.

d

 Refer to Section 3.18.4.3 for additional information on transportation accidents involving

shipping casks and radiographic devices.

3–285

3.19 Thorium in Finished Optical Lenses

3.19.1 Introduction

In 10 CFR 40.13(c)(7), persons who receive, possess, use, or transfer finished optical lenses

containing thorium are exempted from licensing requirements for source material, provided

each lens does not contain more than 30% by weight of thorium. However, the exemption does

not authorize the shaping, grinding, or polishing of such lenses or any manufacturing processes

other than the assembly of such lenses into optical systems and devices without any alteration

of the lens. In addition, the exemption does not apply to the receipt, possession, use, or

transfer of thorium in contact lenses, spectacles, or eyepieces in binoculars or other optical

instruments. This exemption was proposed on May 8, 1963 (28 FR 4621), and issued as a final

rule on August 7, 1963 (28 FR 8021).

The Federal Register notices cited above do not contain quantitative information on radiation

doses to members of the public resulting from use of the exempted optical lenses. The first

notice states that, on the basis of measured external radiation levels, the use of finished optical

lenses as specified by the exemption would not result in an unreasonable hazard to life or

property. However, data on external radiation levels and estimates of dose were not published. 

The second notice does not provide any additional information.

Quantitative information on doses that might be received from use of finished optical lenses

containing thorium, based on data reported in the literature, has been summarized by the

National Council on Radiation Protection and Measurements (NCRP 95). The available

information is discussed in Section 3.19.3.

3.19.2 Description of Items

The exempted items consist of optical glass to which thorium has been added to provide

improved optical properties. Reported uses of thoriated optical glass under this exemption

include lenses for television cameras and 35-mm photographic cameras (NCRP 95). However,

the number of thoriated glass lenses in use in the United States is unknown. Furthermore,

particularly in the case of lenses for 35-mm photographic cameras, certain manufacturers

apparently have used thoriated glass extensively but others have not (Taylor et al., 1983).

Several manufacturers produced thorium-coated lenses up until the late 1980s for military

optical systems, cameras, microfilm readers, and aerial cameras. Lenses containing as much

as 28% thorium oxide are known to have been produced. Safety and environmental concerns

plus the availability of better optical manufacturing methods led to the decline in the use of

thorium with production virtually stopping in the 1980s. Some of the known, popular lenses and

cameras containing thorium that were produced include (Frame and Kolb, 2000):

Canon FL 58mm f1.2

 GAF Anscomatic 38mm F2.8 (Anscomatic 726 camera)

Kodak Ektanar 38mm f2.8 Instamatic 804 camera)

Kodak Ektanon 46mm f3.5 (Signet 40 camera)

Kodak Ektanon 50mm f3.9 (Kodak Bantam RF camera)

SMC Takumar 50mm f1.4 (Asahi Optical Co.)
3–286

Supre Takumar 35mm f2.0, 50mm f1.5, and 55mm f2 (Asahi Optical Co.)

Super Takumar 6X7 105mm f2.4 (Asahi Optical Co.)

Super-Multi-Coated Macro-Takumar (Asahi Optical Co.)

Yashinon-DS 50mm F1.7 (Yashica).

Thoriated optical glass also appears to have been used in ways not authorized under this

exemption. In particular, as summarized in NCRP 95, cases have been reported of eyepieces

in optical instruments containing elevated levels of thorium (i.e., much greater than 0.05% by

weight), even though such uses of thoriated glass are not exempted. The extent of

unauthorized uses of thoriated glass lenses in the United States is unknown.

3.19.3 Summary of Previous Analyses and Assessments

This section discusses estimates of doses from routine use of finished optical glass containing

thorium, as summarized in NCRP 95. In presenting these results, a distinction is made

between uses of thoriated optical glass that are authorized under this exemption

(Section 3.19.3.1) and uses that are not authorized at the present time (Section 3.19.3.2). 

Doses have previously been estimated only for routine uses of thoriated optical glass, but not

for distribution and transport, disposal, or accidents or misuse.

3.19.3.1 Authorized Uses of Thoriated Optical Glass

As indicated in Section 3.19.2, reported uses of finished optical glass containing thorium

allowed under this exemption include lenses for television cameras and 35-mm photographic

cameras (NCRP 95). This section discusses previously published estimates of dose resulting

from routine use of thoriated optical glass in these lenses. External exposure to photons has

been assumed to be the only pathway of concern. External exposure to alpha and beta

particles should not be important because of the short range of these particles in materials that

would be present between the lenses and any radiosensitive tissues of exposed individuals

during normal use. Inhalation or ingestion exposure generally has not been considered,

because radioactive material would not be removed from the glass during normal use and the

exemption does not authorize further shaping, grinding, or polishing of lenses.

Published data on photon dose rates near television cameras and 35-mm photographic

cameras containing thoriated optical glass lenses are summarized in Table 3.19.1 and

described in the following paragraphs.

The measurements of Lewinsky (1985) apply to a lens system in a television camera with five

lenses, three of which contained elevated levels of thorium (9%, 13%, and 10% by weight). 

Absorbed dose rates were measured at various locations around the camera. The

measurements were made in Denmark, and it is unknown if the camera studied (or other

television cameras with lenses containing elevated levels of thorium) is used in the United

States.

The measurements of Taylor et al. (1983) and Waligórski et al. (1985) apply to a photographic

lens mounted on a 35-mm camera body. The two sets of measurements are similar and were

obtained for a lens containing the highest activity of thorium among the different lenses tested. 

The highest activity in a lens reported by Taylor et al. (1983) was 13 kilobecquerel (kBq)
3–287

(0.36 microcurie (Ci)), which corresponds to a mass of 3 g. The mass of the lens was not

reported, but for the 35-mm lens on which the measurements were made, the thorium content

could be on the order of 10% by weight. From the measured dose rate at the back of the

camera, Taylor et al. (1983) also estimated that the absorbed dose to an exposed individual’s

abdomen from 6 hours of camera use during a single day would be about 5 microgray (Gy)

(0.5 mrad).

Taylor et al. (1983) performed measurements on eleven brands of photographic camera lenses

sold in the United States, but only two brands contained elevated levels of thorium. The lenses

studied by Waligórski et al. (1985) were manufactured in the former German Democratic

Republic and have not been sold in the United States. Therefore, it appears unlikely that

thoriated glass lenses for photographic cameras are widely distributed in this country.

3.19.3.2 Unauthorized Uses of Thoriated Optical Glass

As indicated in Section 3.19.2, there have been reports of eyepieces in optical instruments

containing elevated levels of thorium well above 0.05% by weight. The eye pieces reportedly

lacked any labeling or other specifications indicating that thoriated glass had been used, even

though the addition of thorium to finished optical glass that normally would be used near the

eye is not authorized under this exemption. This section discusses published estimates of

doses resulting from unauthorized uses of thoriated optical glass in eyepieces of optical

instruments. External exposure again was the only exposure pathway considered. However,

because of the close proximity of eyepieces to the eye during normal use, external exposure to

the eye to alpha particles has been the primary concern, with external exposure to beta

particles and photons regarded as less important.

Doses to the eye from exposure to alpha particles have been calculated by McMillan and Horne

(1973) and by Casarett et al. (Atomic Energy Commission (AEC), 1974). Doses were

calculated for the germinal cell layer of the cornea, which was assumed to be located at a depth

of 50 m below the surface of the eye. The lens of the eye, which is of concern in routine

radiation protection of workers in regard to preventing cataract formation, lies at a depth of

about 3 mm below the surface of the eye (International Commission on Radiological Protection

(ICRP) 26). Therefore, it could not be irradiated by alpha particles produced in the decay of

thorium and its decay products, due to their short range in tissue.

The calculations by Casarett et al. (AEC, 1974) were based on the following assumptions:

(1) the glass in the eyepiece contains 16% by weight of thorium, which represents the

maximum level observed in any eyepiece; (2) all decay products of thorium are present and in

equilibrium; (3) the air gap between the lens and outer surface of the eye during normal use is

0.1 cm; and (4) the instrument is used for 20 h/wk by a professional. The calculated absorbed

dose to the germinal cell layer of the cornea for these assumptions was 0.44 Gy/yr (44 rad/yr).

The calculations by McMillan and Horne (1973) for alpha irradiation of the germinal cell layer of

the cornea agree with those of Casarett et al. (AEC, 1974) within 20 to 40%. In addition, for an

eyepiece containing 18% by weight of thorium, the measured absorbed dose rate at the surface

of the eyepiece from beta particles and photons was 0.01 mGy/h (1 mrad/h). Based on the

calculations for alpha particles and the measurements for beta particles and photons, McMillan

and Horne estimated that the dose at the surface of the eye from alpha particles may be 50 to
3–288

1000 times greater than the dose from beta particles and photons and, thus, should be the

most important.

3.19.4 Present Exemption Analysis

This section presents estimates of dose from routine use, as well as distribution and transport

and disposal, of finished optical glass containing amounts of thorium allowed under this

exemption. Doses from accidents or misuse also are considered. As in Section 3.19.3, results

for uses of finished optical glass containing thorium that are authorized under this exemption

are discussed separately from results for unauthorized uses that apparently have occurred.

For uses of thoriated optical glass that are authorized under this exemption, the lens would be

enclosed by other materials or there would be additional material between the lens and the

tissues of an exposed individual, and only photon exposures would be of concern. However, in

unauthorized uses of thoriated optical glass, such as eyepieces of optical instruments,

exposure to the eye to alpha and beta particles also would be of concern. This concern is

because of the proximity of the glass to the eye and lack of sufficient absorbing material

between the source and tissues of the eye.

Because thorium in optical glass is dispersed throughout the volume of glass and, as specified

in the exemption, the glass would not be further shaped, ground, or polished, ingestion or

inhalation of glass containing thorium and its decay products would not normally occur. 

Therefore, internal exposure is not considered in this assessment, except in the case of

accidents or misuse discussed in Section 3.19.4.4. Inhalation exposures to 220Rn and its shortlived decay products are considered in Section 3.19.4.1.3.

3.19.4.1 Authorized Routine Uses of Thoriated Optical Glass

This section presents estimates of dose resulting from routine use of thoriated optical glass in

lenses for television cameras and 35-mm photographic cameras. As noted previously, these

are the only reported uses of thoriated optical glass authorized under this exemption.

3.19.4.1.1 Routine Use of Television Cameras

Estimates of effective dose equivalents (EDEs) during routine use of television cameras with

thoriated lenses can be obtained from the measurements of absorbed dose rates by Lewinsky

(1985), which are summarized in Table 3.19.1 and discussed in Section 3.19.3.1. For the highenergy photons emitted in the decay of 232Th and its decay products, measured absorbed

doses near a camera provide a reasonable approximation to the EDE, particularly when an

exposed individual is not located immediately adjacent to the camera lenses, as would normally

be the case.

Operators of television cameras are assumed to be the individuals who would receive the

highest doses, because they would be located near cameras for the longest period of time

during routine use. Furthermore, camera operators normally spend most of their working time

directly behind a camera, rather than at other locations near a camera. If a camera operator

would spend about half of the normal work time during a year (i.e., about 1000 hours) behind a

camera, which should be a reasonable upper bounding exposure time, then, based on the
3–289

absorbed dose rate of 2×104

 mGy (0.02 mrad/h) at this location measured by Lewinsky (1985),

the EDE would be 0.2 millisievert (mSv)/yr (20 mrem/yr).

Although measured dose rates near television cameras are higher at some locations (e.g., at

the top and bottom of the camera and at the side of the lens housing) than at the back of the

camera, the time an individual would spend at these other locations should be much less than

the time an operator spends behind a camera. Therefore, the annual doses in these cases

should be less than the estimate for camera operators obtained above.

The dose estimate of 0.2 mSv/yr (20 mrem/yr) obtained above applies to a lens system in which

three of the five lenses contain elevated levels of thorium, and the average thorium content of

the three lenses is about 10% by weight, i.e., one-third of the limit on thorium content of 30% by

weight specified in the exemption. It is unknown if this combination of lenses with elevated and

normal levels of thorium is typical. Also, it is not known if the amounts of thorium in the lenses

with elevated levels are typical. However, if lens systems containing greater amounts of

thorium were used, then the estimated dose would increase accordingly. Thus, for the camera

lens system considered in this analysis, the dose to a camera operator corresponding to the

maximum thorium content of 30% by weight allowed under this exemption could be 0.6 mSv/yr

(60 mrem/yr).

The extent to which lenses with elevated levels of thorium are used in television cameras is

unknown. In this assessment, the hypothetical collective dose is estimated assuming 1000

such television cameras in use. Each camera is assumed to irradiate a single individual located

behind the camera for 1000 h/yr, as in the assessment of dose to a camera operator described

above. These assumptions should provide a reasonable upper bound to the collective dose

considering the number of individuals that would be present in a television studio or other

location where a camera was used. However, the likelihood is that most of the exposed

individuals would, on average, be located considerably farther from the camera than an

operator. Since the dose rate decreases rapidly with distance from the camera (see

Table 3.19.1), the actual collective dose should be significantly less. Similarly, the thorium

content of television camera lenses used today may be zero, and the individual and collective

EDEs also may be zero.

Based on the assumptions described above, the collective EDE for 1000 television cameras

containing three lenses with an average thorium content of 10% by weight would be

0.2 person-Sv/yr (20 person-rem/yr), and assuming a 20 useful life, the collective EDE would be

4 person-Sv (400 person-rem). 

3.19.4.1.2 Routine Use of Photographic Cameras

Estimates of dose rates from 35-mm photographic cameras with lenses containing elevated

levels of thorium can be obtained from the measurements of absorbed dose rates by Taylor et

al. (1983), which are summarized in Table 3.19.1 and discussed in Section 3.19.3.1. These

data are supported by the measurements of Lewinsky (1985) and of Waligórski et al. (1985),

which also are summarized in Table 3.19.1.

Taylor et al. (1983) measured the absorbed dose rate at the back of a camera and the thorium

content of the lens. The thorium in the lens was estimated to be 13 kBq (0.36 Ci). Using the
3–290

methodology described in Appendix A.4 for sources close to the body, the dose rate at 10 cm

depth in the body was determined to be 1×104 mSv/h (0.01 mrem/h). 

A serious outdoor photographer is assumed to spend 30 days/yr in the field (average

photographers-10 days/yr) and to carry a camera next to the body for 6 hours per day during

that time. This exposure time should be conservative for most photographers. Based on the

assumed exposure time and the absorbed dose rate, the annual EDE would be 0.02 mSv

(2 mrem). For an average photographer the EDE would be 0.007 mSv (0.7 mrem)

The dose estimate obtained above results from an unknown amount, in weight percent, of

thorium in a lens. If photographic lenses are optically similar to the lenses for television

cameras discussed previously, then a lens may contain about 10% by weight of thorium,

i.e., one-third of the limit on thorium content of 30% by weight allowed under this exemption,

which agrees with the estimate obtained in Section 3.19.3.1. Thus, the annual EDE to an

individual photographer corresponding to the maximum allowable thorium content in a lens

would be about 0.06 mSv (6 mrem).

The number of photographic camera lenses with elevated levels of thorium currently in use is

unknown. In this assessment, the hypothetical collective dose is estimated assuming 1 million

such lenses in use. Based on these assumptions, the collective annual EDE for the average

photographers per million lenses would be 7 person-Sv/yr (700 person-rem/yr). For an

assumed 20-year useful life, the cumulative collective EDE would be 140 person-Sv

(14,000 person-rem). If this estimate is assumed to apply to lenses containing 10% by weight

of thorium, as described above, the collective annual dose per million lenses corresponding to

the maximum thorium content of 30% allowed under this exemption would be a factor of three

higher.

3.19.4.1.3 Exposures to Radon During Routine Use

As noted previously, potential inhalation exposures to 220Rn and its short-lived decay products

generally have not been considered in assessing doses during routine use of thoriated optical

glass lenses. However, the following considerations indicate that doses from 220Rn would not

be important.

Howard et al. (1995) have reported measurements of 220Rn emanation from a variety of rock

specimens. The rock type that most closely resembles glass lenses is obsidian. The

measurements on obsidian indicated that the emanation rate of 220Rn is less than 0.002% of the

production rate in the rock. Therefore, for a lens containing 0.26 MBq (7 Ci) of 232Th, which is

the highest activity for a television camera lens measured by Lewinsky (1985), the maximum

production rate of 220Rn would be 2.6×105

 atoms/s and the emanation rate of 220Rn from the

lens would be less than 5 atoms/s. If the released 220Rn is assumed to be removed from the air

only by radioactive decay, which is reasonable when the half-life of 56 seconds is much shorter

than typical air ventilation rates in rooms, the number of released atoms in the air at steady

state would be less than 420. If the release occurs to a small room with a volume of 30 m3

, the

resulting concentration of 220Rn in air would be less than 14 atoms/m3

, or an activity

concentration of less than 0.2 Bq/m3

 (<5 pCi/m3). Using the inhalation dose conversion factor

for 220Rn given in Table 3.1.7, the resulting EDE would be about 0.008 mSv (0.8 mrem) for

1000 h/yr exposure time. This estimate is conservative because it assumes that the radon

decay products would be in activity equilibrium with the parent radionuclide in air. Based on this
3–291

calculation, it is apparent that doses due to releases of 220Rn from a glass lens during normal

use can be neglected compared with doses from external exposure.

3.19.4.2 Distribution and Transport

Individual and collective doses during the distribution and transport of lenses containing thorium

were estimated using the methodology described in Appendix A.3. Separate estimates are

provided for lenses used in television and photographic cameras, because the amounts of

thorium used in each type of lens appear to be considerably different and the annual distribution

of photographic lenses presumably could be considerably higher than the annual distribution of

lenses for television cameras.

3.19.4.2.1 Lenses for Television Cameras

In estimating doses from the distribution and transport of lenses for television cameras, the

amount of thorium in each lens system, corresponding to an average amount of 10% by weight

allowed under this exemption, is assumed to be 100 g. This assumption is based on the data

reported by Lewinsky (1985) that the lens system of one camera contains 100 g of thorium

when the thorium content is about 10% by weight (see Section 3.19.3.1). Since the distribution

of lenses for television cameras presumably is rather limited, the assumed distribution is 1000

lens systems per year.

The distribution and transportation network for lenses for television cameras is unknown. In

this assessment, it is assumed that all lenses are shipped by small truck to United Parcel

Service (UPS) terminals over distances greater than 400 km, which should provide conservative

estimates of dose for this part of the network, and that the lenses then are shipped directly to

the user by small truck over distances between 32 and 400 km. It is further assumed that each

shipment contains 1% of the total annual distribution of lenses; i.e., each shipment contains

10 lens systems.

Based on the assumptions described above, the following results are obtained using the

generic methodology described in Appendix A.3:

 During distribution and transport of lens systems for television cameras, the annual EDE

to an individual truck driver could be as high as 0.01 mSv (1 mrem), but the dose to an

individual terminal worker would be nearly an order of magnitude less. 

• The collective annual EDE for 1000 lens systems distributed per year would be

0.001 person-Sv (0.1 person-rem), due primarily to exposure to terminal workers.

3.19.4.2.2 Lenses for Photographic Cameras

In estimating doses from the distribution and transport of lenses for photographic cameras, the

amount of thorium in each lens, corresponding to an average amount of 10% by weight allowed

under this exemption, is assumed to be 3 g. This assumption is based on the data reported by

Taylor et al. (1983) that one manufacturer distributes a lens that contains 3 g of thorium and an

estimate that this amount of thorium is 10% by weight (see Section 3.19.3.1). The assumed

distribution is 1 million lenses per year.
3–292

The distribution and transportation networks for lenses for photographic cameras are unknown. 

In this assessment, the following network is assumed. First, 80% of the lenses are assumed to

be shipped by large truck to UPS terminals over distances greater than 400 km, and 20% to

UPS terminals over distances between 32 and 400 km. Second, 50% of the lenses are

assumed to be sent from the terminals to large telephone- or mail-order firms, which are

represented by warehouses, by small truck over distances of 32 to 400 km, and 50% by small

truck to retail stores. Finally, each terminal and warehouse is assumed to receive 1% of the

total annual distribution of lenses, and each retail store is assumed to receive 0.1%.

Based on the assumptions described above, the following results are obtained using the

generic methodology in Appendix A.3:

 During distribution and transport of lenses for photographic cameras, the EDE to an

individual truck driver, terminal worker, worker in a telephone- or mail-order firm, or

worker in a retail store would be 0.03 to 0.07 mSv (3 to 7 mrem).

• The collective annual EDE for 1 million lenses distributed per year would be 1 person-Sv

(100 person-rem), almost entirely from exposures in retail stores.

3.19.4.3 Disposal

Optical lenses containing thorium eventually may be sent to landfills or incinerators for disposal. 

Doses from disposal in landfills or by incineration are estimated using the generic methodology

described in Appendix A.2. The dose estimates assume that 80% of the lenses are sent to

landfills and 20% to incinerators.

The number of lenses containing elevated levels of thorium that might be disposed during a

year is unknown. As in the analysis of doses from routine use discussed in Sections 3.19.4.1.1

and 3.19.4.1.2, doses from disposal are estimated for 1,000 television camera lens systems

and 1 million photographic camera lenses disposed per year. Furthermore, each television

camera lens system is assumed to contain 100 g of 232Th and each photographic camera lens

is assumed to contain 3 g of 232Th, these values again corresponding to the average amount of

thorium of 10% by weight. Therefore, the assumed annual disposals are 100 kg for television

camera lenses and 3000 kg for photographic camera lenses.

3.19.4.3.1 Disposal in Landfills

In the generic methodology for landfill disposal described in Appendix A.2, doses are estimated

for waste collectors, workers at landfills, off-site individuals who reside near landfills during

operations or after closure, and individuals who might reside on the landfill sites after closure. 

However, for disposal of glass lenses in landfills, only doses from external exposure are

considered, because the lenses either would be intact or broken into pieces too large to inhale

and ingestion exposure also would be very unlikely, and the thorium in the lenses would not be

leachable in water to any significant extent. Furthermore, as discussed in Section 3.19.4.1.3,

releases of 220Rn and subsequent inhalation exposures would be negligible. Therefore, only

external exposure to waste collectors, landfill workers, and future on-site residents are of

concern for this assessment.
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Based on the assumptions described above, the estimates of individual and collective doses

from landfill disposal of television camera lens systems and photographic camera lenses are

summarized as follows.

Disposal of 800 television camera lens systems per year in the landfill, and correcting for

disposal of the limited number of items (see Appendix A.2.3.1.4)–

 The annual EDE to individual waste collectors would be 1×103

 mSv (0.1 mrem). The

annual EDE to a future on-site resident would be 1×103

 mSv (0.1 mrem). The annual

EDE to individual workers at landfills would be about a factor of 10 less.

 The collective EDE from 1 year’s disposals would be about 0.3 person-Sv

(30 person-rem), due almost entirely to exposure to future on-site residents for

1000 years after facility closure. If exposure to future on-site residents are not taken

into account, the annual collective EDE from exposure to waste collectors and workers

at landfills would be 0.001 person-Sv (0.1 person-rem).

Disposal of 800,000 photographic camera lenses per year in the landfill–

 The annual EDE to individual waste collectors would be 0.007 mSv (0.7 mrem). The

annual EDE to a future on-site resident would be 0.008 mSv (0.8 mrem). The annual

EDE to individual workers at landfills would be about a factor of 10 less.

 The collective EDE from 1 year’s disposals would be about 9 person-Sv

(900 person-rem), due almost entirely to exposure to future on-site residents for

1000 years after facility closure. If exposure to future on-site residents are not taken

into account, the annual collective EDE from exposure to waste collectors and workers

at landfills would be 0.04 person-Sv (4 person-rem).

The dose estimates for landfill disposal given above would be quite conservative if the actual

disposal of lenses containing elevated levels of thorium is substantially less than the assumed

numbers of lenses disposed per year. In addition, lenses with elevated levels of thorium used

in these calculations appear to contain about one-third of the maximum amount allowed under

the exemption.

3.19.4.3.2 Disposal in Incinerators

In the generic methodology for disposal in incinerators described in Appendix A.2, doses are

estimated for waste collectors, workers at incinerators, and off-site individuals who reside near

the incinerators during operations. In estimating doses to waste collectors and workers at

incinerators, only external exposure to the thorium in optical lenses would be of concern, but

thorium is assumed to be released into the air during incineration.

Based on the assumptions described above and the generic disposal methodology in

Appendix A.2, the following estimates of individual and collective doses from disposal in

incinerators are obtained.
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Disposal of 200 television camera lens systems per year in incinerators–

 The annual EDE to individual waste collectors would be 0.001 mSv (0.1 mrem). The

annual EDE to individual workers at incinerators would be about three orders of

magnitude less, and the dose to individual off-site residents near incinerators would be

about six orders of magnitude less.

 The collective EDE from 1 year’s disposals would be about 2×104 person-Sv

(0.02 person-rem), due almost entirely to exposure to waste collectors.

Disposal of 200,000 photographic camera lenses per year–

 The annual EDE to individual waste collectors would be 0.04 mSv (4 mrem). The

annual EDE to individual workers at incinerators would be about three orders of

magnitude less, and the dose to individual off-site residents near incinerators would be

about six orders of magnitude less.

 The collective EDE from 1 year’s disposals would be about 0.006 person-Sv

(0.6 person-rem), due almost entirely to exposure to waste collectors.

The dose estimates for incineration given above would be quite conservative if the actual

incineration of lenses containing elevated levels of thorium is substantially less than the

assumed numbers of lenses incinerated per year. In addition, lenses with elevated levels of

thorium used in these calculations appear to contain about one-third of the maximum amount

allowed under the exemption.

3.19.4.4 Accidents or Misuse

Two scenarios for accidents or misuse of thoriated optical lenses could be considered:

(1) shattering of a lens resulting in potential ingestion or inhalation exposure and (2) a fire

resulting in potential inhalation exposure. However, based on the following considerations, it

does not appear credible to assume that either of these scenarios could result in a significant

internal exposure.

Lenses on television and photographic cameras are broken on occasion. However, most of the

broken glass would be contained by the housing for the lenses, and most of the glass would

consist of large particle sizes. Therefore, it is highly unlikely that shattering of a glass lens

could result in a significant airborne concentration of respirable particles, or that a significant

amount of shattered glass could be ingested. Furthermore, even if some broken glass were

inhaled or ingested, the dose to body tissues would be much less than for intakes of thorium

that is not incorporated into glass, because the thorium would not easily be leached from the

glass and absorbed into blood. Therefore, only the very small amount of thorium at the surface

of a glass lens could deliver a dose from alpha particles to either the lungs or the

gastrointestinal tract. Based on these considerations, it does not seem reasonable to assume

that shattering of a glass lens could result in significant inhalation or ingestion exposure.

A fire could occur during distribution or storage of thoriated optical lenses. However, the boiling

point of thorium is much higher than temperatures that would occur in an accident-related fire

(NUREG–0137), and the glass itself should begin to fuse only at temperatures above those
3–295

expected in a fire. Therefore, it is unlikely that a fire would result in a significant release of

thorium from glass lenses and that a firefighter or other individual could receive a significant

inhalation exposure at a distribution or storage location for the lenses.

3.19.5 Summary

This assessment has considered doses to members of the public from uses of thorium in

finished optical lenses that are authorized under this exemption. Doses from two particular

products were considered: lens systems in television cameras and lenses for 35-mm

photographic cameras. Results of the assessments for the television and photographic camera

lenses are presented in Table 3.19.2 and 3.19.3, respectively. These results may be

summarized as follows:

 Based on reported measurements of dose rates near lenses, doses to individuals during

routine use of television or photographic cameras could be substantial (i.e., several

hundreds of Sv/yr (tens of mrem/yr)) if the lenses contain the maximum amount of

thorium allowed under this exemption. However, based on the limited information

available, lenses containing thorium appear to contain about one-third of the maximum

allowable amount. Therefore, the dose estimates for individuals during routine use

probably should be reduced by at least a factor of 3 to represent doses for the expected

amounts of thorium in lenses.

 All estimates of individual and collective doses, except doses to individuals during

routine use of lenses, are based on assumptions about the number of lenses containing

thorium to which individuals and populations would be exposed per year—specifically,

1,000 television camera lens systems and 1 million 35-mm photographic camera lenses. 

If the actual number of such lenses distributed, used, or disposed per year were known,

these dose estimates could be adjusted accordingly. It seems likely, for example, that

far fewer than 1 million lenses for photographic cameras containing elevated levels of

thorium are distributed, used, and disposed per year. In addition, as noted above, these

dose estimates probably should be reduced by about a factor of 3 to represent doses for

the expected amounts of thorium in lenses.

 If the assumptions of 1,000 television camera lens systems and 1 million 35-mm

photographic camera lenses distributed, used, or disposed per year do not

underestimate the actual numbers of thoriated optical lenses, then individual doses

during routine use of lenses would be considerably higher than individual doses during

distribution or disposal, regardless of the actual thorium content of the lenses.

With the exception of the dose estimates for individuals during routine use, the estimates of

individual and collective doses obtained in this assessment probably provide upper bounding

representations of doses from actual uses of lenses containing elevated levels of thorium. This

conclusion is based primarily on the presumption that the inclusion of thorium in optical lenses

has declined over time and the number of thoriated optical lenses actually distributed, used, or

disposed per year should be considerably less than the quantities of 1,000 for television camera

lens systems and 1 million for photographic camera lenses assumed in this assessment. This

presumption is supported by noting that only a single instance of lenses in television cameras

has been reported and that most major brands of 35-mm camera lenses currently distributed in
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the United States apparently do not contain elevated levels of thorium. However, the actual use

of lenses containing elevated levels of thorium cannot be estimated based on the available

information, and more realistic estimates of dose for the situations other than exposure to

individuals during routine use were not assessed.
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Table 3.19.1 Summary of Measured Photon Dose Rates From Cameras Containing

Thoriated Optical Glass

Camera Type Location

Dose Rate

(mrad/h)a

Television camerab Surface of front lens

Side of lens housing

Bottom of camera

Top of camera

Back of camera

0.5 m from surface of front lens

1 m from surface of front lens

2 m from surface of front lens

3.0

0.8

0.4

0.15

0.02

0.04

0.02

0.0

35-mm photographic camerac Lens surface

Back of camera

0.48

0.08

35-mm photographic camerad Lens surface

Film plane of camera

9.3e

0.15e

a

 1 mrad/h = 10 Gy/h.

b

 Measurements of Lewinsky (1985).

c

 Measurements of Taylor et al. (1983).

d

 Measurements of Waligórski et al. (1985).

e

 Values are exposure rates in units of mR/h; absorbed dose rates in units of mrad/h are

obtained by multiplying exposure rate by factor of 0.877 (Cember, 1983).
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Table 3.19.2 Summary of Potential Radiation Doses From Use of Finished Optical

Lenses Containing Thorium in Television Cameras a

Exposure Scenario

Individual Annual Effective

Dose Equivalent

(mrem)c

Collective Effective

Dose Equivalentb

(person-rem)c

Distribution and transportb 1d 0.1

Routine usee 20 400

Disposalb

 Landfills

 Incinerators

0.03f

0.1h

 30g

 0.02

Accidents or misuse

 Shattering of lensi

 Firej

NAk NA

a

 Dose estimates are based on assumption that all lenses contain thorium of 10% by weight

allowed under exemption. Lens systems in television cameras are assumed to include three

lenses with elevated levels of thorium, and each lens system is assumed to contain 100 g of

thorium.

b

 Dose estimates are based on the assumption that 1,000 television camera lens systems

containing elevated levels of thorium are distributed, used, or disposed per year.

c

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

d

 Dose estimate applies to truck drivers; dose estimate is considerably less for terminal workers

(see Section 3.19.4.2.1).

e

 Dose estimates apply to operators of television cameras.

f

 Dose estimate applies to future on-site residents at landfills; dose estimates for workers at

landfills and waste collectors, are less, and doses to off-site members of the public are

assumed to be negligible, due to low dispersibility and leachability of thorium incorporated into

glass (see Section 3.19.4.3.1).

g

 Dose estimate applies to future on-site residents at landfills for 1,000 years after disposal due

to 1 year’s disposals; if dose to future on-site residents is not taken into account, estimated

annual collective dose to waste collectors and workers at landfills is 0.004 person-Sv

(0.4 person-rem).

h

 Dose estimate applies to waste collectors at incinerators; dose estimates are considerably

less for workers at incinerators and off-site members of the public (see Section 3.19.4.3.2).

i

 Shattering of glass lenses is assumed not to result in significant inhalation or ingestion

exposure, due to physical form of lenses and incorporation of thorium into glass (see

Section 3.19.4.4).

j

 Fire during distribution or storage of glass lenses is assumed not to result in significant

inhalation exposure, due to high boiling point of thorium and high fusion temperature of glass

into which thorium is incorporated (see Section 3.19.4.4).

k

 Not applicable.
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Table 3.19.3 Summary of Potential Radiation Doses From Use of Finished Optical

Lenses Containing Thorium in 35-mm Photographic Cameras a

Exposure Scenario

Annual Individual Effective

Dose Equivalent

(mrem)c

Annual Collective Effective

Dose Equivalentb

(person-rem)c

Distribution and transportb 7d 100

Routine usee 0.7 14,000

Disposalb

 Landfills

 Incinerators

0.8f

4h

900g

 0.6

Accidents or misuse

 Shattering of lensi

 Firej

NAk NA

a

 Dose estimates are based on assumption that all lenses contain an amount of thorium of 10%

by weight. Lenses in 35-mm photographic cameras are assumed to contain 3 g of thorium.

b

 Dose estimates are based on assumption that 1 million photographic camera lenses

containing elevated levels of thorium are distributed, used, or disposed per year.

c

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

d

 Dose estimate applies to truck drivers; dose estimates are somewhat less for terminal

workers, workers in large telephone- or mail-order firms, or workers in retail stores (see

Section 3.19.4.2.2).

e

 Dose estimates apply to users of photographic cameras.

f

 Dose estimate applies to future on-site residents at landfills; dose estimates for workers at

landfills and waste collectors are less, and doses to off-site members of the public are assumed

to be negligible, due to low dispersibility and leachability of thorium incorporated into glass (see

Section 3.19.4.3.1).

g

 Dose estimate applies to future on-site residents at landfills for 1,000 years after disposal due

to 1 year’s disposals; if dose to future on-site residents is not taken into account, estimated

annual collective dose to waste collectors and workers at landfills is 0.04 person-Sv (4

person-rem).

h

 Dose estimate applies to waste collectors at incinerators; dose estimates are considerably

less for workers at incinerators and off-site members of the public (see Section 3.19.4.3.2).

i

 Shattering of glass lenses is assumed not to result in significant inhalation or ingestion

exposure, due to physical form of lenses and incorporation of thorium into glass (see

Section 3.19.4.4).

j

 Fire during distribution or storage of glass lenses is assumed not to result in significant

inhalation exposure, due to high boiling point of thorium and high fusion temperature of glass

into which thorium is incorporated (see Section 3.19.4.4).

k

 Not applicable.

3–301

3.20 Aircraft Engine Parts Containing Nickel-Thoria Alloy

3.20.1 Introduction

In 10 CFR 40.13(c)(8), persons who receive, possess, use, or transfer finished aircraft engine

parts containing nickel-thoria alloy are exempted from licensing requirements for source

material, provided (1) the alloy does not contain more than 4% by weight of thorium and (2) the

thorium is dispersed in the alloy in the form of finely divided thorium dioxide. The exemption

does not authorize the manufacture of such aircraft engine parts, but the exemption permits

such activities as repair of aircraft engine parts and handling and processing of scrap. An

exemption for any finished product or part containing thorium-metal alloys with not more than

4% by weight of thorium was proposed on August 7, 1963 (28 FR 8043). The present

exemption, which applies only to aircraft engine parts containing nickel-thoria (i.e., nickelthorium dioxide) alloy, was issued as a final rule on November 18, 1967 (32 FR 15872).

The Federal Register notices cited above contain little information on radiological impacts on

the public from use of nickel-thoria alloy aircraft engine parts containing not more than 4% by

weight of thorium. The notice of proposed rulemaking states only that use of the thorium-metal

alloys would not result in an unreasonable hazard to life or property. The notice for the final

rule states that it is highly unlikely that the small number of workers carrying out operations

involving the exempted nickel-thoria alloys would receive radiation doses in excess of limits for

the public recommended by the International Commission on Radiological Protection (ICRP). 

At the time the exemption was established, these recommendations, which also had been

incorporated into the radiation protection standards for the public by the Atomic Energy

Commission (AEC, 25 FR 10914), included limits on dose equivalent of 5 millisieverts (mSv)

(0.5 rem) per year to the whole body from external exposure and 30 mSv (3 rem) per year to

the bone or 15 mSv (1.5 rem) per year to the lungs from ingestion or inhalation (ICRP 2,

ICRP 6). There have not been any other published sources of information on radiological

impacts on the public for this exemption.

3.20.2 Description of Items

The use of nickel-thoria alloy containing not more than 4% by weight of thorium is restricted to

aircraft engine parts. Such parts require high-strength materials that can withstand high

temperatures, and the addition of thorium to nickel alloys acts as a dispersion hardening agent

(Fraser et al., 1985).

At the present time, however, nickel-thoria alloys apparently are used in aircraft engine parts

only to a limited extent, if at all. This conclusion is based on the following information. First, as

described in Section 3.16.2, the total amount of thorium used in all aerospace alloys has

declined over the last decade from 4000 to 5000 kg per year to only about 100 kg per year

(Hedrick, 1985; Hedrick, 1991; Hedrick, 1994). Second, most of the thorium used in the

aerospace industry has been in the form of magnesium-thorium alloys, but only a small fraction

has been in the form of nickel-thoria alloys (Hedrick, 1991; Hedrick, 1994). Finally, information

obtained from government officials who monitor thorium use in manufacturing (Phone call,

J. B. Hedrick, Bureau of Mines, U.S. Department of the Interior, Washington, DC, 1996) and

from radiation safety personnel at aircraft companies (Phone call, R. Edwards, The Boeing Co.,
3–302

Seattle, WA, 1996) indicates that aircraft engine parts containing nickel-thoria alloys may not be

manufactured at all at the present time.

Efforts to obtain specific information on nickel-thoria alloy parts used in aircraft engines from

licensed manufacturers, including their size, shape, weight, and thorium content, were

unsuccessful. Therefore, information on magnesium-thorium alloy parts used in aircraft

engines, described in Sections 3.16.2 and 3.16.4.1, is assumed to be suitable for use in a dose

assessment for the exempted nickel-thoria alloy parts.

3.20.3 Summary of Previous Analyses and Assessments

As indicated in Section 3.20.1, little information has been published on radiological impacts on

the public associated with this exemption. The final rule establishing the exemption indicated

only that doses to workers carrying out operations involving nickel-thoria alloy parts containing

not more than 4% by weight of thorium should be substantially less than 5 mSv/yr

(<500 mrem/yr), but further information was not provided on the magnitude of any dose

estimates.

In the absence of published information on radiological impacts on the public associated with

this exemption, a hypothetical assessment presented in the following section was performed.

3.20.4 Present Exemption Analysis

This section presents estimates of dose from routine use of exempted aircraft engine parts

containing nickel-thoria alloy. Doses from accidents and misuse also are considered.

For routine uses of the exempted aircraft engine parts, external exposure should be the primary

pathway of concern. As described in Section 3.20.4.3.2, inhalation and ingestion exposures

should not normally occur during routine use.

As indicated in Section 3.20.1, no information on external dose rates near nickel-thoria alloy

parts is available. However, information on dose rates near magnesium-thorium alloy parts

discussed in Section 3.16.4.1 can be used in conjunction with calculations indicating the

difference in dose rates between nickel-thoria and magnesium-thorium alloy parts of the same

size, shape, and weight percent of thorium, as described in the following section, to obtain

estimates of external dose rates from nickel-thoria alloy parts. The typical amount of thorium

used in nickel-thorium alloy is unknown.

3.20.4.1 Estimated Dose Rates Near Finished Parts

In this analysis, estimated external dose rates near nickel-thoria alloy parts are based on the

measurements and calculations of dose rates near magnesium-thorium alloy parts discussed in

Section 3.16.4.1. Relevant information is summarized as follows.

First, based on a calculated exposure rate, the effective dose equivalent (EDE) rate at a

distance of 5 cm from a magnesium-thorium alloy casting containing 4% by weight of thorium,

which presumably had recently been chemically separated, was estimated to be 0.002 mSv/h
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(0.2 mrem/h). The dose estimate would be about a factor of 2.5 higher for thorium that had

been aged for about 20 years or longer.

Second, based on a measured exposure rate for a magnesium-thorium alloy ingot that was

considerably heavier than the casting on which the measurement described above was made,

the dose rate does not depend greatly on the mass of the source for the same weight percent

of thorium. Therefore, the difference in dose rate between a nickel-thoria alloy source and a

magnesium-thorium alloy source containing the same weight percent of thorium should depend

primarily on the difference in elemental composition for the two types of alloy.

Calculations performed with MicroShield (Computer Codes, Grove Engineering, 1996) were

used to investigate the effect on the dose rate of using the different metals (nickel or

magnesium) in the alloy. Calculations were performed for a solid spherical source, in which the

amount of self-shielding provided by nickel was significantly greater than the self-shielding

provided by magnesium, and for a cylindrical shell source, in which the self-shielding provided

by either metal would be insignificant. Calculations for a solid sphere indicated that the dose

rate from a nickel source should be about two times higher than from a magnesium source

containing the same weight percent of thorium. Thus, the decrease in dose rate due to the

increased atomic number for nickel compared with magnesium (i.e., the increase in

self-shielding provided by the source) is more than compensated by the higher density of nickel

(8.9 g/cm3

) compared with magnesium (1.7 g/cm3), which permits a considerably greater

amount of thorium per unit volume in the alloy for the same weight percent. The calculations

for a cylindrical shell indicated that the dose rate from a nickel source should be about three

times higher than from a magnesium source containing the same weight percent of thorium. 

This is the expected result, considering the ratio of the densities of the two metals, for a source

providing little self-shielding.

Thus, the calculations described above indicate that the dose rate from a nickel-thoria alloy

source should be about two to three times higher than the dose rate from a similar magnesiumthorium alloy source. In this hypothetical assessment, the difference in dose rates is

conservatively assumed to be a factor of 3 in all cases. This is based on the assumption that

nickel-thoria alloy parts used in aircraft engines should not be massive, in order to minimize the

weight of the aircraft, and, thus, should not provide a significant amount of self-shielding. 

However, the calculations also indicate that this assumption should not overestimate the dose

rate by more than a factor of 2 for any size and mass of the nickel-thoria alloy parts.

The information described above is used in the hypothetical dose assessments for distribution

and transport and routine use described in the following two sections.

3.20.4.2 Distribution and Transport

As in the assessment for magnesium-thorium alloy parts described in Section 3.16.4.2, the

highest doses from external exposure to nickel-thoria alloy parts during distribution and

transport should be received by truck drivers during transport. The estimated dose to an

individual truck driver was obtained from:

• the estimated EDE to an individual truck driver during transport of magnesium-thorium

alloy parts of 0.04 mSv/yr (4 mrem/yr) given in Section 3.16.4.2,
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• the conclusion described in the previous section that the dose rate from a single nickelthoria alloy part would be three times higher than from a single magnesium-thorium alloy

part, and

• the results of calculations performed using MicroShield (Computer Codes, Grove

Engineering, 1996) that showed, for the assumed stacking of 100 cartons of parts

during transport described in Section 3.16.4.2, the shielding provided by the nickelthoria alloy parts would decrease the dose rate near the stack by a factor of 2 compared

with the dose rate near a stack of magnesium-thorium alloy parts. Based on these

assumptions, the EDE to an individual truck driver would be about 0.1 mSv (10 mrem)

per year.

The dose estimate given above applies to nickel-thoria alloy containing 4% by weight of

thorium, which is the maximum amount allowed under this exemption. Thus, this estimate

would be conservative if the alloy contained a lesser weight percent of thorium, as is typically

the case for the magnesium-thorium alloy parts discussed in Section 3.16.2. In addition, the

assumption that a truck driver would be exposed to 100 cartons of parts for 100 hours per year

(see Section 3.16.4.2) probably is conservative, based on the information described in

Section 3.20.2, which indicates that there is little likelihood of production and distribution of the

exempted parts at the present time.

The collective dose to truck drivers can be estimated from the total amount of thorium in nickelthoria alloy parts transported per year. This amount is unknown but, as noted above, is likely to

be small. If it is assumed, that 5 kg of thorium is distributed per year (5% of all thorium

estimated to be distributed annually), the collective EDE to truck drivers would be a factor of

2 less than the estimate for magnesium-thorium alloy parts. This takes into account the selfshielding provided by stacking of the parts during transport as described above, or about

2×105

 person-Sv/yr (0.002 person-rem/yr). Again, this estimate probably is quite conservative,

because the assumed amount of thorium distributed per year under this exemption is a

considerable overestimate.

3.20.4.3 Routine Use

The following section considers doses from external exposure during routine use of aircraft

engine parts containing nickel-thoria alloy. The potential for internal exposure, which should be

relatively unimportant during routine use, is discussed in Section 3.20.4.3.2.

3.20.4.3.1 Doses From External Exposure

As in the assessment for magnesium-thorium alloy parts in Section 3.16.4.3, the highest doses

from external exposure to nickel-thoria alloy parts during routine use would be received by

maintenance workers on aircraft engines. The dose to an individual maintenance worker was

estimated from (1) the estimated EDE during maintenance on aircraft engines containing

magnesium-thorium alloy parts of 0.04 mSv/yr (4 mrem/yr) and (2) the conclusion described in

Section 3.20.4.1 that the dose rate from a single nickel-thoria alloy part would be three times

higher than from a single magnesium-thorium alloy part. This estimate is appropriate when the

parts are in the form of thin castings that provide little self-shielding.
3–305

Based on the assumptions described above, the EDE to an individual maintenance worker

would be about 0.1 mSv/yr (10 mrem/yr). Again, this dose estimate applies to nickel-thoria

alloy containing 4% by weight of thorium, which is the maximum amount allowed under the

exemption. As described in Section 3.16.4.3, this estimate could under predict actual doses if

several parts were used in a single aircraft engine. However, the assumptions that no shielding

would be provided by other metal parts in an engine and that exposure would occur for

1000 hours per year should be conservative.

The collective dose to aircraft engine maintenance workers depends on the annual distribution

of thorium under this exemption and the useful lifetime of the alloy parts. Neither of these

factors is known for this exemption. However, if it is assumed, as in the assessment for

magnesium- and tungsten-thorium alloy parts in Section 3.16.4.3, that the annual distribution of

thorium under this exemption is 5 kg (5% of all thorium estimated to be distributed annually)

and the useful lifetime of the parts is 10 years, and if it is further assumed that the parts are

mainly in the form of thin castings that provide little self-shielding, the estimated collective EDE

to aircraft engine maintenance workers from 1 year’s distribution would be about

0.05 person-Sv/yr (5 person-rem/yr). This estimate is conservative, again because the

assumed amount of thorium distributed appears to be a considerable overestimate, given that

nickel-thoria alloy may not be used in aircraft engine parts to any significant extent at the

present time.

3.20.4.3.2 Potential for Internal Exposure

During routine use of aircraft engine parts containing nickel-thoria alloy, inhalation and ingestion

exposures potentially could occur only during occasional maintenance activities involving

grinding or drilling of parts. Information on potential internal exposures obtained from a facility

authorized by the Federal Aviation Agency to repair aircraft engines (Phone call, T. Evans,

Chromalloy Corporation, Tallahassee, FL, 1996) is described in the following paragraphs.

Grinding is performed occasionally in the vicinity of nickel-thoria alloy parts. Specifically, for

heat shields consisting of nickel-thoria alloy that are held in place in aircraft engines by rivets

and cones made of other material, grinding of old rivets in housing for the heat shields may be

required to loosen the part. Such grinding could release respirable particles. However,

grinding occurs only on the rivets but not on the nickel-thoria heat shield itself. Therefore,

under normal grinding procedures, there should be no release of thorium.

On rare occasions, drilling into the nickel-thoria alloy part occurs to match the size of drill holes

in the alloy part and its housing, and nickel-thoria drill cuttings are produced. The drill cuttings

should be too large to be respirable, but they could be ingested. However, ingestion of drill

cuttings is not expected to occur. Furthermore, even if inadvertent ingestion of drill cuttings

would occur, the resulting dose should be substantially less than would be estimated using the

standard dose coefficient given in Table 3.1.7, due to the large size of the cuttings.

Thus, the potential for internal exposure to thorium contained in the exempted aircraft engine

parts does not appear to be an important concern during routine use. Some inadvertent

inhalation or ingestion of thorium could occur, but such exposures should be rare. Therefore,

doses from internal exposure during routine use are not estimated in this hypothetical

assessment because such calculations would not be meaningful, especially in comparison with

doses from external exposure.
3–306

3.20.4.4 Disposal

The final disposition of aircraft engine parts containing nickel-thoria alloy following their useful

lifetime is likely either to be recast into ingots or recycled through smelting with scrap metal. 

However, in this hypothetical assessment since these products or parts are exempt from

licensing, the possibility exists that a fraction of them could be disposed of in landfills. Disposal

by incineration presumably would not occur and is not considered.

Of the hypothetical 5 kg/yr of thorium distributed under this exemption, 5% or 0.25 kg is

assumed to be introduced into scrap metal and a like quantity also into landfills.

To estimate potential doses from disposal of the products or parts as ordinary waste in landfills,

the generic disposal methodology in Appendix A.2 is used. It is further assumed that this

disposal is limited to 100 landfills instead of the default 3,500 of Appendix A. The annual EDE

to a waste hauler would be 2×105

 mSv (0.002 mrem) and the estimate of collective dose from

landfill disposal would be less than 1×105

 person-Sv (<0.001 person-rem).

Estimates of individual doses from smelting of the exempt aircraft engine parts at a facility for

processing of scrap metal were obtained using the generic methodology for recycling in

Appendix A.2. Doses are estimated for workers at a smelter and members of the public

residing near a smelter. 

Based on the assumed amount of thorium recycled per year and the generic methodology for

recycling in Appendix A.2, the estimates of individual dose are summarized as follows:

 For individual workers at a smelter, the annual EDE would be 1×105

 mSv (0.001 mrem).

 For off-site members of the public, the individual annual EDE would be less than

1×105

 mSv (<0.001 mrem).

3.20.4.5 Accidents

An accident scenario involving a fire in a warehouse is considered in this hypothetical

assessment. The warehouse is assumed to contain 0.25 kg of thorium, an amount that is 5%

of the assumed annual distribution for this exemption and, thus, should be conservative. Based

on the generic methodology described in Appendix A.1, the EDE to a firefighter using

respiratory protection would be 7×105

 mSv (0.007 mrem), and the EDE to an individual who is

not using respiratory protection during cleanup after the fire would be 5×104

 mSv (0.05 mrem).

3.20.5 Summary

This hypothetical assessment has considered doses to the public from use and disposal of

finished aircraft engine parts containing nickel-thoria in which the thorium content is 4% by

weight, which is the maximum amount allowed under this exemption. Results of the

assessment are presented in Table 3.20.1 and may be summarized as follows:

 Estimated annual doses to individuals from external exposure during transport and

routine use of aircraft engine parts containing nickel-thoria alloy are about 0.04 mSv
3–307

(4 mrem) and 0.1 mSv (10 mrem) respectively. Doses from internal exposure during

transport and routine use should be unimportant.

 Estimated collective doses during routine use of aircraft engine parts containing nickelthoria alloy, based on an assumption of 5 kg of thorium distributed per year under this

exemption and a useful lifetime of 10 years for the parts, are 0.05 person-Sv

(5 person-rem/yr). The collective dose during distribution and transport should be much

less.

 Disposal of used aircraft engine parts into landfills was considered. For landfill disposal,

the estimated dose to individual workers is 2×105

 mSv/yr (0.002 mrem/yr).

 Disposal of used aircraft engine parts into scrap metal in smelters was considered. For

disposal by recycling, the estimated dose to individual workers is 1×105

 mSv/yr

(0.001 mrem/yr).

The hypothetical dose estimates in Table 3.20.1 are based on assumptions about the total

amount of thorium involved for the exposure scenarios. No information is available about the

current distribution of thorium in the exempt nickel-thoria alloy parts; hence, these assumptions

should be conservative. Indeed, the available information suggests that the exempt nickelthoria alloy parts may no longer be used in aircraft engines to any significant extent.

This hypothetical assessment also considered doses resulting from a fire in a warehouse. 

Potential doses from a warehouse fire, either during the fire or cleanup after the fire, appear to

be low. This is due to the small amount of thorium that normally would be stored in a

warehouse and the assumptions that a firefighter would use respiratory protection and that only

a small fraction of the thorium involved in a fire would be available for inhalation or ingestion

during or after a fire.
3–308

Table 3.20.1 Summary of Potential Radiation Doses From Use of Finished

Aircraft Engine Parts Containing Nickel-Thoria Alloy a

Exposure Scenario

Individual Annual

Effective

Dose Equivalent

(mrem)c

Annual Collective

Effective

Dose Equivalentb

(person-rem)c

Distribution and transportd 4e 0.002

Routine usef 10 5g

Disposal

 Landfills

 Processing in scrapi

0.002h

0.001j

<0.001

<0.001

Accidents or misuse

 Warehouse firek 0.05l

a

 Dose estimates are based on assumption that all finished parts contain the maximum amount

of thorium of 4% by weight allowed under exemption. Typical amount of thorium used in nickelthoria alloy is unknown.

b

 Dose estimates are based on assumption that 5 kg of thorium is distributed per year under

this exemption. Based on a lack of data on any distribution of thorium in aerospace alloys, and

considering that nickel-thoria alloy apparently is used in aircraft engine parts only to a very

limited extent, assumption probably is conservative.

c

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

d

 Dose estimates apply to truck drivers; doses to other members of the public during distribution

and transport are substantially less (see Sections 3.20.4.2 and 3.16.4.2).

e

 Dose estimate is based on assumption that 200 products or parts are transported per

shipment (see Sections 3.16.4.1 and 3.16.4.2).

f

 Dose estimates apply to maintenance workers on aircraft engines; doses are considerably less

to other members of the public during routine use.

g

 Collective dose to maintenance workers on aircraft engines from 1 year’s distribution,

assuming 10-year useful lifetime for finished parts.

h

 The primary exposure pathway is direct exposure.

i

 Dose estimates are based on the assumption that 0.25 kg/yr of thorium is introduced into

scrap metal stream.

j

 Dose estimate applies to workers at smelter; estimate of individual dose is considerably less

for off-site members of the public.

k

 Dose estimates are based on assumption that 0.25 kg of thorium is contained in warehouse.

l

 Dose estimate is total dose during cleanup after fire without use of respiratory protection;

estimated dose is somewhat less to firefighter using respiratory protection.
3–309

3.21 Uranium in Fire Detection Units

3.21.1 Introduction

In 10 CFR 40.13(d), detector heads containing uranium for use in fire detection units are

exempted from licensing requirements for source material, provided that each detector head

does not contain more than 185 becquerel (Bq) (0.005 microcurie (Ci)) of uranium. This

exemption was proposed on July 27, 1963 (28 FR 7677), and issued as a final rule on

December 27, 1963 (28 FR 14309).

The Federal Register notices cited above do not contain information on analyses of radiological

impacts on the public from use of the exempted fire detectors. The first notice states only that

the Atomic Energy Commission determined that it was desirable and consistent with radiation

safety to grant this exemption. This statement was based on their consideration of whether the

use of this product could be conducted without a reasonable hazard to life or property.

Since there are no known radiological assessments for uranium in fire detection units, annual

individual and collective effective dose equivalents (EDEs) were estimated using exposure

scenario assumptions derived for smoke detectors containing 241Am.

3.21.2 Description of Items

Fire detection units containing up to 185 Bq (0.005 Ci) of uranium do not appear to be used in

the United States at the present time. The original petition for an exemption was submitted by

Electro-Tronics, Inc., which went bankrupt in 1992. According to the company’s founder and

past president, only two or three prototypes were produced. Besides uranium as the ionizing

source, the fire detector incorporated a one-chamber design instead of the two-chamber design

common in the 1960s. Electro-Tronics worked with the University of Florida in the development

of this fire detector prototype. The uranium was placed on a 2-inch planchet and heated to

adhere the uranium to the disk. It appears that natural uranium was used. No other fire

detectors using uranium were produced by Electro-Tronics. (Private communication with

H. H. Morgan, Sr., founder and past president of Electro-Tronics, NC, July 1996). The Florida

Department of Health and Rehabilitative Services, Office of Radiation Control, did not have any

current or archival licensee records for Electro-Tronics. (Private communication with P. Vause,

Office of Radiation Control, Material Section, Florida Department of Health and Rehabilitative

Services, Tallahassee, FL, June 1995). No one contacted within the smoke or fire detector

industry knew of any uses of uranium in fire detection units (Private communication with J.

Johnson, retired president of Cerberus Pyrotronics, Hilton Head, SC, July 1996; private

communication with P. Patty, Associate Managing Engineer, Underwriters Laboratory, Inc.,

Northbrook, IL, May 1995; private communication with M. Bunker, Fire Protection Engineer,

National Fire Protection Association, Chicago, IL, May 1995; private communication with

I. Mande, Manager of Industry Affairs, G.S. Edwards, Norwalk, CT, May 1995; and private

communication with I. Ellner, Director of Engineering, Cerberus Pyrotronics, Cedar Knolls, NJ,

June 1995). Radium-226, 63Ni, and 241Am are the only radionuclides known to have been used

in fire and smoke detection equipment (Private communication with J. Johnson, retired

president of Cerberus Pyrotronics, Hilton Head, SC, July 1996; private communication with

P. Patty, Associate Managing Engineer, Underwriters Laboratory, Inc., Northbrook, IL, May

1995; private communication with M. Bunker, Fire Protection Engineer, National Fire Protection
3–310

Association, Chicago, IL, May 1995; private communication with I. Mande, Manager of Industry

Affairs, G.S. Edwards, Norwalk, CT, May 1995; and private communication with I. Ellner,

Director of Engineering, Cerberus Pyrotronics, Cedar Knolls, NJ, June 1995).

3.21.3 Summary of Previous Analyses and Assessments

In the notice (28 FR 7677) and final rule (28 FR 14309), there was no reference as to whether

any dose analysis was performed to support the exemption. In addition, there are no known

reports on production or use or any radiological dose analyses attributed to the use of uranium

in fire detection units.

3.21.4 Present Exemption Analysis

To perform the radiological assessment for this exemption, many exposure scenario

assumptions are based on smoke detectors containing 241Am (O’Donnell et al., 1981) (see

Section 2.15.4). Collective doses are not assessed at this product is not believed to be

currently manufactured or in wide-scale use. However, for purposes of estimating distribution,

transport, and disposal, an annual distribution and disposal of 10 million smoke detectors is

assumed.

3.21.4.1 Distribution and Transport

The individuals who might receive the highest doses during the distribution of smoke detectors

were found to be stock handlers working in warehouses. These individuals could receive an

annual EDE of 6×104

 mSv (0.06 mrem), as shown in Table 3.21.1. 

3.21.4.2 Routine Use

Although these detectors were not produced, the routine use assumptions for smoke detectors

containing 241Am are applied to this exemption. It is assumed that 2 detectors are installed in a

residence. Doses are calculated assuming 1 year of use and a one-time purchase and

installation. If these fire detectors have a 10-year useful life, this set of smoke detectors could

deliver the estimated doses for 10 years. The primary exposure pathway would be direct

irradiation. Using source-to-receptor distances and exposure durations similar to those in

O’Donnell et al. (1981), the estimated individual and collective EDEs for household use of fire

detectors are shown in Table 3.21.2. The maximum annual individual EDE attributed to smoke

detector home use is estimated to be less than 1×105

 mSv (<0.001 mrem).

3.21.4.3 Disposal

The generic disposal methodology, as described in Appendix A.2, is used to estimate individual 

EDEs due to disposal of fire detectors. In applying this methodology, however, it is assumed

that the source remains intact during waste collection, handling, and burial. Thus, the pathway

for inhalation is not operative in the case of off-site members of the public near landfills and the

pathways of ingestion and inhalation are not operative in the case of workers at landfills and

waste collectors at both landfills and incinerators. Doses are estimated for 10 million detectors

discarded annually, as well as for the total number that could be disposed in a landfill for 30

years. It is conservatively assumed that 1850 MBq (50 mCi) are disposed annually (10 million
3–311

detector units) with 185 Bq (0.005 Ci) of U-natural in each detector. Eighty percent of the

discarded detectors would be sent to 3500 municipal landfills, and 20% would be sent to 150

municipal waste incinerators. Individual EDEs that could be received from land disposal and

combustion of 10 million detectors disposed are summarized in Table 3.21.3.

Annual EDE to the waste collectors and landfill operators is estimated to be less than

1×105

 mSv (<0.001 mrem). The annual individual EDE for a maximally exposed off-site

receptor attributed to ingestion of groundwater is also less than 1×105

 mSv (<0.001 mrem). 

The annual individual EDE, to a future on-site receptor assuming 30 years of landfill operation

(and waste accumulation) and 30 years of postclosure controls is less than 1×105

 mSv

(<0.001 mrem).

For incineration of 20% of the smoke detectors, the annual individual EDE for waste collectors

and incinerator workers are estimated to be 4×105 mSv (0.004 mrem) and less than

1×105 mSv (<0.001 mrem), respectively. The estimated annual individual EDE to the maximally

exposed off-site individual due to municipal waste incinerator operations is less than

1×105

 mSv (<0.001mrem).

3.21.4.4 Accidents and Misuse

The accident exposure scenarios evaluated for fire detectors containing uranium include a

warehouse fire, a residential fire, and a transportation accident resulting in dispersal of

shipment contents. Using the generic accident methodology, described in Appendix A.1, it is

assumed that radiation exposure occurs to firefighters. Resulting EDEs are summarized in

Table 3.21.3. 

3.21.4.4.1 Warehouse Fire

In this fire scenario, it is assumed that 14,400 detectors (the contents of two shipments) are

stored in a small warehouse. It is assumed that firefighters are wearing supplied-air respirators

and are exposed for 30 minutes. Based on previous investigations, small chips of uranium can

be ignited under enhanced oxidation conditions; in air and carbon dioxide, respirable aerosol

fractions up to approximately 1% have been reported. Below this level, the respirable aerosol

fractions showed wide variation with temperature and other conditions. A release factor of 103

seemed to be reasonable for exempt items containing small amounts of uranium. Therefore, if

a fire resulted in a warehouse where 14,400 fire detectors were stored, the estimated EDE to a

firefighter is 1×105

 mSv (0.001 mrem). The estimated EDE to cleanup workers due to

particulate resuspension is 9×105

 mSv (0.009 mrem).

3.21.4.4.2 Residential Fire

In the residential fire scenario, doses are estimated for firefighters, a person escaping a fire or a

rescuer, and cleanup workers. Only two fire detection units are assumed to be in the

residence. The estimated EDE to a firefighter or a clean-up worker is estimated to be less than

1×105

 mSv (<0.001 mrem). The EDE to a person escaping a fire or to an individual trying to

rescue a person from the fire is also estimated to be less than 1×105

 mSv (<0.001 mrem). 

Unlike the firefighter, this individual does not have respiratory equipment.
3–312

3.21.4.4.3 Transportation Accident

For the transportation accident, it is assumed the accident occurs indoors, such as in a

cargo-handling bay or in a transportation vehicle. Here, as in the previous accident scenarios,

doses to a firefighter are estimated. It is assumed one shipment of detectors is involved in the

fire. The individual EDE to a firefighter is 4×105

 mSv (0.004 mrem). The estimated EDE to

cleanup workers for a transportation accident is 4×104

 mSv (0.04 mrem).

3.21.4.4.4 Misuse

In the case of misuse, the exposure to a teacher who removes a 185 Bq (0.005 Ci) uranium

source from a smoke detector for use in classroom demonstrations about radioactivity is

considered. To estimate the potential radiation dose to the teacher, it is assumed that the

teacher stores the uranium source in a convenient location in a classroom and is exposed at an

average distance of about 1 meter from the source for 1000 h/yr. The EDE rate at 1 meter

from a 185 Bq (0.005 Ci) uranium source without regard for shielding by other materials is

about 3.6×10-10 mSv/h (3.6×108

 mrem/h) and the annual EDE to the teacher would be less

then 1×10-5 mSv (<0.001mrem). Also, it is assumed that the teacher handles the 185 Bq

(0.005 Ci) uranium source at a nominal 1 cm distance (due to the size of the source and disk)

for 10 h/yr during classroom demonstrations. Using VARSKIN MOD2 (Computer Codes,

Durham, 1992), the dose equivalent to a small area of skin on the hand is about 0.1 mSv

(10 mrem). Assuming a 10 cm2

 exposed skin area out of a total skin area of 1.8×104 cm2

(ICRP 26) and a skin weighting factor of 0.01 (ICRP 60), the calculated EDE would be less than

1×105

 mSv (<0.001 mrem). For potential ingestion of material due to the handling of a 185 Bq

(0.005 Ci) uranium source, the generic accident methodology developed in Appendix A.2 is

applied for spills of radioactive material in the form of a powder. It is assumed, first, that 10% of

the material on the source is deposited on the body and, second, that 0.1% of this deposited

material is ingested before it is removed from the body by washing. The EDE would be less

than 1×105

 mSv (<0.001 mrem).

3.21.5 Summary

Apparently fire detection units containing 185 Bq (0.005 Ci) of uranium were never

manufactured for consumer use. Since the 1960s, fire and smoke detector manufacturers have

chosen to use radionuclides other than uranium. Hypothetical radiological impacts from routine

use, disposal, and accidents associated with storage of fire detection units containing uranium

are summarized in Table 3.21.3. Many of the assumptions used in this assessment were

based on smoke detectors containing 241Am. For routine operations, the annual individual and

collective EDEs appear to be very low.
3–313

Table 3.21.1 Dose Estimates for Routine Exposure From Distribution and Transport a

Activity

Exposure

Duration

(h)

Number of

Detectors

Handled

Distance

 (cm)

Individual Annual

Effective Dose 

Equivalent

 (mrem)b

Warehouse workers

 Storeroom clerk

 - Handle cargo

 - Near cargo

 - In building

 Forklift operator

Total

30

0.083

259

598

3

3,600

7,200

7,200

30

180

600

180

0.002

<0.001

0.005

0.05

0.06

Transport

 Large truck driver

 -Handle cargo

 - Driving load and

 unloading

 Total 

122

50

3

7,200

30

300

0.009

0.003

0.01

At merchandise facility

 Store clerk

 - Handle item

 - Near display

 Total

 Store customer

 - Handle item

 - Near display

 Total

30

1,000

1

3

3

144

3

144

30

300

30

300

0.002

<0.001

0.002

<0.001

<0.001

<0.001

a

 Assume 7,200 detectors per shipment (50 pallets with 48 cartons with 3 detectors per carton,

144 detectors/pallet). Each driver is assumed to make ten 400-km trips per year at

approximately 80 km/h.

b

 1 mrem = 0.01 mSv.
3–314

Table 3.21.2 Dose Estimates for Routine Exposure for Household Members

Activity

Exposure

Duration

(h/yr)

Number of

Detectors

Handled

Distance

(cm)

 Individual Annual

Effective

Dose Equivalent

(mrem)a

Homeowner

 Purchase, Install

 and Maintain

 Sleep

3.5

2,920

2

2

30-90

180 & 600

<0.001

<0.001

 Other

Total

1,460 2 600 <0.001

<0.001

Spouse

 Sleep

 Other

Total

2,920

2,920

2

2

180 & 600

180 & 600

<0.001

<0.001

<0.001

a

 1 mrem = 0.01 mSv.
3–315

Table 3.21.3 Summary of Effective Dose Equivalents for Uranium in Fire Detectors

Scenarios - Receptor

Individual Annual Effective

Dose Equivalent

 (mrem)a

ROUTINE USE

Household resident 

Warehouse worker

<0.001

0.06

DISPOSAL

Landfill

 Worker

 -Collector

 -Landfill operator

 On-site receptor

 Off-site receptor(s)

 -Groundwater releases

Incinerator

 Worker

 -Collector

 -Incinerator operator

 Off-site receptor

<0.001

<0.001

<0.001

<0.001

0.004

<0.001

<0.001

ACCIDENTS AND MISUSE

Warehouse fire

 Firefighter

 Cleanup worker

Residential fire

 Firefighter

 Resident escaping fire

 Cleanup worker

Transportation fire

 Firefighter

 Cleanup worker

Misuse

 Teacher

0.001 

0.009 

<0.001

<0.001

<0.001

0.004

0.04

<0.001

a

 1 mrem = 0.01 mSv.

4–1

4 ITEMS CONSIDERED FOR EXEMPT DISTRIBUTION

In addition to the dose assessments for exempted products or materials containing radioactive

material presented in Sections 2 and 3, dose assessments were performed in this study for

certain items containing byproduct material that are generally licensed at the present time. The

particular devices considered in this study are listed in the table of contents. Some of these

items may also be used under a specific license.

The generally licensed items considered in this study were identified by the Nuclear Regulatory

Commission (NRC) as potential candidates for exemption, primarily because their use is not

expected to result in significant radiological impacts on the public and, thus, the requirements of

a general license were viewed as excessively burdensome (NRC, Memoranda, Bernero, 1994). 

The expectation of insignificant radiological impacts results from the small amounts of

byproduct material contained in each device, requirements on the design and manufacturing of

the devices for the purpose of enhancing safety, and the small number of devices normally

distributed annually.

However, a systematic assessment of radiological impacts on the public associated with normal

use and with accidents and misuse is needed for each of the generally licensed items that are

potential candidates for exemption, in order to verify that the impacts indeed would be

insignificant. If, on the other hand, the assessments indicate the radiological impacts would not

be insignificant, the results of the assessments could be used by the NRC to establish

additional conditions for exempting the particular devices. Examples might include, limits on

the amounts of radioactive material in each device that are more stringent than the present

limits specified in the general licenses, requirements on the design and manufacturing of the

devices to enhance safety, and requirements for labels on the devices and instructions for

users.

The dose assessments for the generally and specifically licensed items containing byproduct

material do not raise any technical issues that have not been addressed in the assessments for

exempted products or materials containing byproduct material in Section 2. In particular, the

data on radiological properties of byproduct materials presented in Section 2.1 are used in all

dose assessments for the generally licensed items.

4–3

4.1 Static Eliminators and Ion Generators Containing Polonium-210

4.1.1 Introduction

Static eliminators and ion generators are devices that contain a radioactive source for

the purpose of reducing electric charge buildup on equipment and materials. The radiation

from the source produces ions in air, which neutralize the static charges in their vicinity. Most

of the static eliminators and ionization generators used in the United States are manufactured

domestically, but some are imported from England.

A general license is issued to anyone in 10 CFR 31.3 to transfer, receive, acquire,

own, possess, and use byproduct material incorporated in certain devices that have been

manufactured, tested, and labeled in accordance with the specifications contained in a specific

license granted to the manufacturer by the Commission. These devices are described below:

 Static elimination device. Devices designed for use as static eliminators that contain, as

a sealed source or sources, byproduct material consisting of a total of not more than

19 megabecquerel (MBq) (500 microcurie (Ci)) of 210Po per device.

 Ion generating tube. Devices designed for ionization of air that contain, as a sealed

source or sources, byproduct material consisting of a total of not more than 19 MBq

(500 Ci) of 210Po or 1.9 GBq (50 mCi) of tritium (3

H) per device.

In addition, certain static eliminators and ion generating tubes containing 3

H, 85Kr, or 210Po are

covered by the general license provided in 10 CFR 31.5 for use only by commercial and

industrial firms; research, educational, and medical institutions; individuals in the conduct of

their business; and Federal, State, and local government agencies. There are no source limits

on the amount of byproduct material used in the devices, but an applicant for a specific license

to manufacture or initially transfer devices for use under 10 CFR 31.5 must demonstrate that

the devices will meet certain requirements contained in 10 CFR 32.51. These requirements are

described below:

 The device can be safely operated by persons not having training in radiological

protection.

 Under ordinary conditions of handling, storage, and use of the device, the byproduct

material contained in the device will not be released or inadvertently removed from the

device, and it is unlikely that any person will receive in any 1-year period a total effective

dose equivalent (EDE) in excess of 5 millisieverts (mSv) (500 mrem), or the sum of the

deep-dose equivalent and committed dose equivalent to any individual organ or tissue

other than the lens of the eye in excess of 50 mSv (5 rem); an eye dose equivalent in

excess of 15 mSv (1.5 rem); or a shallow-dose equivalent in excess of 50 mSv (5 rem)

to the skin or to any extremity (i.e., hand, elbow, arm below the elbow, foot, knee, or leg

below the knee).

 Under accident conditions (such as fire and explosion) associated with handling, storage

and use of the device, it is unlikely that any person would receive an external dose or

internal dose commitment in excess of 0.15 Sv (15 rem) to the whole body, head and
4–4

trunk, active blood-forming organs, gonads, or lens of the eye; 2 Sv (200 rem) to the

hands and forearms, feet and ankles, or localized areas of skin averaged over areas no

larger than 1 cm2

; and 0.5 Sv (50 rem) to any other organs.

It has been suggested that some static eliminators or ion generators are candidates for

exemption from the general licensing requirements of either 10 CFR 31.3(a) and (d) or

10 CFR 31.5. This assessment evaluates the potential radiation doses that could result if the

receipt, possession, use, transfer, and disposal of certain of these products were exempt from

licensing.

4.1.2 Description of Items Considered for Exempt Distribution

Static electric charges may develop when different materials are in close contact. If the

materials are nonconducting materials, the charge will remain as a static charge, and the

presence of this static charge can lead to various problems. Commercial applications for static

eliminators include the following: (1) to reduce the risk of fire or explosion due to static charge

buildup and discharge in volatile and explosive environments, (2) to reduce the buildup of static

charges that can damage electronic circuits and hard drives during assembly and repair of

personal computers, (3) to reduce the buildup of dust on surfaces to be electroplated or

painted, and (4) to reduce the static cling of processed material on sheet-fed webs and rollers. 

As a consumer product, their use is generally limited to elimination of static charges on

photographic film and lenses and the static charges that can hinder the delicate operation of

balances of precision.

4.1.2.1 Consumer Products

The two widely distributed consumer products are a static eliminator in the form of a brush and

an air ionizer for a balance of precision. These products are generally licensed at present

under 10 CFR 31.3(a) and (d) for use by anyone, subject to the requirements of these general

licenses.

The brushes consist of a handle made of plastic and aluminum with soft (camel) hair bristles

and a removable cartridge at the end of the handle (NUREG/CR–1775). The removable

cartridge contains the 210Po source. Two sizes are manufactured with the approximate

dimensions of 12.7 cm in length, 1.6 to 1.9 cm in thickness, and either 2.5 or 7.6 cm in width. 

The smaller brush is said to nominally contain 7.4 MBq (200 Ci) of 210Po, while the larger

brush nominally contains 19 MBq (500 Ci) of 210Po. The 210Po source is located about 6 mm

behind an open grill in the cartridge, which fits into the handle of the brush (NUREG/CR–1775).

The air ionizer for a balance of precision is a small source located inside a housing assembly to

protect the surface of the source from damage. It is placed inside the balance chamber on the

balance floor in close proximity to the balance pan, and its purpose is to eliminate any static

electrical charges that may interfere with operation of the balance. A custom size standoff can

be purchased to properly position the source within the balance chamber (i.e., about 2.5 cm

from the balance pan). The source nominally contains less than 7.4 MBq (<200 Ci) of 210Po. 

Powder or dust accumulation inside the chamber can also be removed as necessary using a

small static eliminator brush.
4–5

Due to the short half-life of 138 days for 210Po, a static eliminator has an effective life of

approximately 1 year. Each static eliminator is dated to indicate when it should be replaced by

the user, and labels on these devices state that they should be returned to the manufacturer for

proper disposal. Should these products be covered by an exemption from licensing, the

number of sources disposed of as ordinary trash would probably go from a small fraction at

present (estimated to be approximately 10%) to essentially 100%.

4.1.2.2 Commercial Products

The two widely distributed commercial products are the linear air ionizer and the ion air gun or

blower. These products are generally licensed at present under 10 CFR 31.5 for use only by

commercial and industrial firms; research, educational, and medical institutions; individuals in

the conduct of their business; and Federal, State, and local government agencies.

The linear air ionizer is designed to be mounted 2.5 cm above processed material on sheet-fed

webs or rollers. These devices are used to eliminate static electrical charges, which cause

material jamming, feed problems, or reduced process speeds. They are available in lengths

from 10.2 cm to 224 cm and in source strengths from 0.37 GBq (10 mCi) to 7.4 GBq (200 mCi)

of 210Po. The sources are mounted inside a rugged protective housing with a typical width and

height of 2.5 to 5.1 cm. These devices are safe to use in volatile and explosive environments

because they do not require electricity or power supplies and have no moving parts. The

sources in these units can normally be replaced by the user.

The ion air gun and blower are designed to eliminate static electricity from large work areas. 

For example, compressed air passing through a cylindrical source of 210Po in the air gun is

ionized, allowing for quick removal of static charge on a surface being prepared for painting. 

Dust and dirt clinging to the surface is blown away by the air stream, and with static eliminated,

they are not attracted back to the surface. In an ion air blower, air is gently blown through

a cylindrical source of 210Po by a fan to cover an area such as a bench top and to eliminate

static charges that may damage equipment during assembly and repair. The cylindrical

sources used in these devices typically contain from as little as 0.37 GBq (10 mCi) to as much

as 1.9 GBq (50 mCi) of 210Po, depending on the intended application. The sources in these

units can normally be replaced by the user.

The user is notified prior to expiration and provided with the information necessary for renewal. 

If the customer agrees to a renewal of the lease, a new replacement source is sent to the

customer, and the customer returns the old source to the manufacturer for disposal. Should

these devices be covered by an exemption from licensing, the number of sources disposed of

as ordinary trash would probably go from a very small fraction at present (estimated to be less

than 1%) to essentially 100%.

4.1.3 Summary of Previous Analyses and Assessments

An assessment of the radiological impacts on the public from static eliminators distributed as

consumer products under 10 CFR 31.3 has been made by Buckley et al. (NUREG/CR–1775)

and the National Council on Radiation Protection and Measurements (NCRP 95). The NCRP

provided estimates of the EDE to individuals and the public based on the organ dose

calculations from the earlier work of Buckley et al. Results of the extensive organ dose
4–6

calculations of Buckley et al., are not addressed in this analysis, but an attempt has been made

to summarize the major features and conclusions of their assessment.

Buckley et al. based their assessment on the only line of static eliminator devices manufactured

and sold in the United States as of 1980. These devices, which utilized 210Po, were employed

mainly to eliminate static charges from high-fidelity phonograph records and from photographic

film and lenses. The 210Po was contained in ceramic microspheres with a range of diameters

from 20 to 60 m and a mean diameter of about 38 m. The physical size of the microspheres

was large enough to make inhalation of the material unlikely, and it was claimed they were

insoluble in body fluids, if ingested. The microspheres were heat treated to fix the polonium,

then coated with nickel before being resin bonded to an aluminum backing plate for use in the

static elimination devices. Experience indicated that the microspheres permitted some

migration of 210Po from the source to other surfaces, but the greatest hazard from routine usage

was inhalation of crushed microspheres from the 210Po source (Robertson and Randle, 1974).

In the United States, static eliminator brushes are usually quoted as having nominal source

activities of either 7.4 MBq (200 Ci) or 19 MBq (500 Ci) of 210Po at the time of manufacture

and distribution. Buckley et al. assumed, however, that the nominal activities did not

refer to the time of manufacture but to 1 year prior to the expiration date stamped on each static

elimination device. They noted that tests conducted in England on the 210Po in nominal

7- to 9-MBq (200- to 500-Ci) static eliminators showed activities of 10 ± 1 MBq (280 ± 30 Ci)

and 28 ± 3 MBq (750 ± 80 Ci), respectively, and the expiration date was 16 months after the

test date (Webb et al., 1975). Other measurements on similar units made at 11.5 months

before the expiration date showed 6.7 MBq (180 Ci) and 17 MBq (450 Ci) of 210Po,

respectively. Thus, Buckley et al. assumed that all sources were effective for 18 months after

manufacture, the amount of activity of the nominal 19-MBq (500-Ci) static eliminator at the

time of manufacture and distribution would be 46 MBq (1250 Ci), and the amount of activity at

the end of its effective lifetime of 18 months (or 1 year of usage by the customer) would be

3 MBq (80 Ci).

To assess the potential for internal exposure during routine usage, Buckley et al. relied on

results of wipe tests that had been conducted on individual static elimination devices to

ascertain the amount of activity available for inhalation or ingestion (Webb et al., 1975;

NUREG/CR–0070). The maximum activity to be removed by vigorous wipe tests of any one

source was about 0.02% of the total activity in the 210Po source. The maximum activity to be

removed by wet wipe tests over the outside of any one device was 0.000024% of the total

activity of the 210Po source. The latter value was used by Buckley et al. By assuming 10% of

this amount could be ingested and 0.1% inhaled, 50-year committed organ doses were

estimated for 1 year of routine usage. For disposal of the static eliminators after 1 year of use,

Buckley et al. found that the only significant organ doses to members of the public were those

resulting from inhalation and ingestion of 210Po in the stack effluent from an incinerator.

The NCRP has carried out a dose assessment for static eliminator brushes as a consumer

product in NCRP 95. Based on an estimated annual sale of 37,000 brushes, each containing a

nominal activity of 19 MBq (500 Ci) of 210Po and the extensive organ dose calculations of

Buckley et al., the NCRP calculated a collective EDE to the public of 0.13 person-Sv

(13 person-rem) when both routine use and disposal were considered. The NCRP also noted

that the most significant doses associated with the use of static eliminator brushes were based

on inadvertent burning of devices stored in a warehouse, for example, and they calculated that
4–7

the EDE to a firefighter at a warehouse fire could be as much as 0.32 Sv (32 rem). This

calculation was also based on the organ dose estimates of Buckley et al., who assumed

(1) the warehouse fire involved 1000 static eliminators with source activities of 46 MBq

(1250 Ci), (2) 10% of the 210Po in the warehouse was released as respirable particles, and

(3) the firefighter breathed the resulting concentrations in the warehouse over an 8-hour period

without any respiratory protection. The NCRP noted, however, that the respirators normally

worn by firefighters should provide adequate protection if considered in the calculation.

It is important to note here that both Buckley et al. and the NCRP considered the potential for

internal exposure to individuals and the public from disposal of static eliminator brushes at

landfills and incinerators, which implies that they believed that this product was exempt from

licensing. However, this was not the situation that existed at the time of their analyses or the

situation that exists at the time of the present exemption analyses for this report.

4.1.4 Present Exemption Analysis

This section provides an assessment of the radiation doses to individuals and the public from

routine use, transportation and distribution, disposal, and accidents.

A technology similar to that used in making 241Am sources for smoke detectors is currently

being used to make the 210Po sources. The 210Po sources made of ceramic microspheres are

no longer used in the manufacturing of static elimination devices in the United States. The

210Po sources made in the United States have a silver backing plate covered by a thin gold foil

and a second composite foil of gold and 210Po. These foils are locked together by a pressure

weld metallurgy process. The composite foil of gold and 210Po is then gold plated to provide an

encapsulated source that is insoluble and inert in most chemicals. The solid metal source is

mechanically fastened within a rigid housing and steps are taken to prevent disassembly of

the source housing. Vibration and impact normal to commercial applications will not adversely

affect source integrity. Hence, it is assumed that ingestion and inhalation of the 210Po from the

source will not occur normally, and the principal pathway for exposure during routine use will be

external irradiation of the whole body.

For the small static elimination devices currently distributed as consumer products under

10 CFR 31.3, an initial 210Po source activity of 19 MBq (500 Ci) is assumed in estimating the

maximum radiation dose to an individual user and an annual distribution of 30,000 units with a

total initial 210Po source activity of 0.56 TBq (15 Ci) in estimating the collective radiation dose to

members of the public. For the large static elimination devices currently distributed as

commercial products and used under 10 CFR 31.5, an initial 210Po source activity of 7.4 GBq

(200 mCi) is assumed in estimating the maximum radiation dose to an individual user and an

annual distribution of 10,000 units with an initial 210Po source activity of 1.9 GBq (50 mCi) per

unit or 19 TBq (500 Ci) total in estimating the collective radiation doses to the public. These

assumptions regarding the number of devices distributed annually and the total radioactivity

contained in these devices are based on information from an industry source (Phone call,

L. Keating, NRD, Inc., Grand Island, NY, October 1995).
4–8

4.1.4.1 Distribution and Transport

It is assumed that the devices are shipped primarily by a parcel-delivery system. A local

parcel-delivery driver in a large van is assumed to pick up the static eliminators in small lots and

take them to a regional terminal for shipment to other regional terminals for delivery to

customers. A daily shipment is assumed to consist of 120 consumer devices and 40

commercial devices (i.e., the annual distribution of 30,000 consumer products and 10,000

commercial products divided by 250 work day/yr). It is further assumed that semi-trucks are

used to transport the static eliminators between terminals and that the static eliminators pass

through an average of five regional terminals before reaching their final destination. The

distance between the regional terminals is assumed to be approximately 400 km.

Based on the above assumptions and the generic methodology of Appendix A.3, the individual

receiving the largest dose is the local driver, who is assumed to pick up an average of 120

small consumer devices containing a total of 2.2 GBq (60 mCi) of 210Po and 40 large

commercial devices containing a total of 74 GBq (2 Ci) of 210Po from the same manufacturer

each day (250 days/yr). This individual is estimated to receive an annual EDE 6×104

 mSv

(0.06 mrem) from the small consumer devices and 0.02 mSv (2 mrem) from the large

commercial devices. The individual doses are substantially less to other drivers, terminal

workers, and members of the public along the truck routes.

The annual collective EDEs to all parcel-delivery drivers, terminal workers, and members of the

public along all truck routes are estimated to be 6×106

 person-Sv (6×104 person-rem) in the

case of the small consumer products and 2×104

 person-Sv (2×102 person-rem) in the case of

the large commercial products.

4.1.4.2 Routine Use

Because static eliminators are made in a variety of designs and may be used in many different

exposure situations, the following representative scenarios were chosen to indicate potential

doses from routine use.

Scenario I. It is assumed that a small consumer device initially containing 19 MBq (500 Ci) of

210Po is on a workbench in the immediate environment of a person, who is exposed for

2000 h/yr at an average distance of 2 meters from the device. From data in Table 2.1.2 of

Section 2.1, the dose rate at 2 meters from this device would be 6.25×10-12 Sv/h

(6.25×1010 rem/h) initially or 3×10-12 Sv/h (3×1010 rem/h) when averaged over the period of

1 year to account for decay of the 210Po source. The annual dose equivalent to the individual

would be less than 1×105

 mSv (<0.001 mrem). For simplicity, it is assumed that this person is

typical, and no one else is exposed. Using a calculated dose of 6×106

 mSv (6×104 mrem)

instead of the less than value, the collective EDE from routine use for an annual distribution of

30,000 devices containing a total of 0.56 TBq (15 Ci) of 210Po would be approximately

2×104

 person-Sv (2×102 person-rem). The devices are assumed to have an effective lifetime

of 1 year.

Scenario II. It is assumed that a large commercial device initially containing 7.4 GBq (200 mCi)

of 210Po is used in the immediate environment of three different people exposed at an average

distance of 3 meters from the device for 2000 h/yr. From the data in Table 2.1.2 of Section 2.1,

the initial dose rate at 3 meters from this device would be 1.2×109

 Sv/h (1.2×107rem/h). When
4–9

the dose rate is averaged over the period of 1 year, as before, to account for radioactive decay

of the 210Po source, the annual EDE to an individual exposed at 3 meters is estimated to be

0.001 mSv (0.1 mrem). For an annual distribution of 10,000 devices containing a total of

19 TBq (500 Ci) of 210Po, the collective EDE from routine use would be approximately

3×102

 person-Sv (3 person-rem). The devices are assumed to have an effective lifetime of

1 year.

4.1.4.3 Disposal

To estimate potential individual and collective doses to the public from disposal of the 210Po in

static eliminators and ion generators, the generic disposal methodology in Appendix A.2 is

used. The large commercial devices contain 19 TBq (500 Ci) initially and 3 TBq (80 Ci) at the

assumed time of disposal 1 year later. The small commercial devices contain 0.56 TBq (15 Ci)

initially and 89 GBq (2.4 Ci) at the assumed time of disposal. The conservative assumption is

made that all static eliminators are disposed of at the end of 1 year.

During waste collection at both incinerators and landfills and disposal at landfills, it is assumed

that the sealed sources in these devices remain intact and the 210Po is not dispersed (see

Section 4.1.4). Based on this assumption and the short half-life of 210Po, the only significant

pathway of exposure is direct whole-body irradiation of waste collectors and landfill workers by

photons from the 210Po in the static eliminators and ion generators. All exposure pathways are

assumed, however, to be operative in the case of workers at incinerators and off-site members

of the public near the incinerator sites (see Appendix A.2).

The estimates of individual and population doses from landfill disposal of 80% of large

commercial devices containing 210Po are summarized as follows:

 For collectors at landfills, the individual annual EDE would be less than 1×105

 mSv

(<0.001 mrem), and the annual collective EDE could be 2×105

 person-Sv

(2×103

 person-rem).

 For workers at landfills, the individual annual EDE would be less than 1×105

 mSv

(<0.001 mrem), and the annual collective EDE could be 9×106

 person-Sv

(9×104

 person-rem).

The estimates of individual and population dose from incineration of 20% of large commercial

devices containing 210Po are summarized as follows:

 For collectors at incinerators, the individual annual EDE could be 3×105

 mSv

(0.003 mrem), and the annual collective EDE could be 4×106

 person-Sv

(4×104

 person-rem).

 For workers at incinerators, the individual annual EDE could be 1×104

 mSv

(0.01 mrem), and the annual collective EDE could be 4×105

 person-Sv

(4×103 person-rem).

 For off-site members of the public near incinerators, the individual annual EDE would be

less than 1×105

 mSv (<0.001 mrem), and the annual collective EDE could be

2×103

 person-Sv (0.2 person-rem).
4–10

To estimate potential individual and collective doses for landfill disposal and incineration of

small consumer devices containing 210Po, the above values for large commercial devices can

simply be multiplied by 0.03 (i.e., the ratio of 0.09 TBq (2.4 Ci) disposed of each year in small

consumer devices to the 3 TBq (80 Ci) disposed of each year in large consumer devices).

4.1.4.4 Accidents and Misuse

In the case of misuse, the scenario selected, consistent with that used for other exceptions,

involves a consumer product containing 19 MBq (500 mCi) of 210Po that is carried in a

individual’s pocket when not in use. It is assumed that the alpha particles from the 210Po source

are attenuated by the source housing of the handheld device so the dose is due entirely to the

0.8-MeV gamma rays from the 210Po source. Based on gamma-ray calculations with

MicroShield (Computer Codes, Grove Engineering, 1996), the dose equivalent to the skin was

estimated to be 2.6×104

 mSv/h (0.026 mrem/h) initially or 1.2×104 mSv/h (0.012 mrem/h)

when averaged over a period of 1 year to account for decay of the 210Po source. The EDE rate

to the whole body was estimated to be 1.7×106

 mSv/h (1.7×104 mrem/h) or 7.8×107 mSv/h

(7.8×105

 mrem/h) when averaged over the period of 1 year. In keeping with the modeling of

Appendix A.4, the dose equivalent rate to the skin was based on a calculation for a separation

distance of 1 cm between the skin and 210Po source of the handheld device. The EDE rate for

the total body was based on a calculation at a body depth of 10 cm, which is considered a

reasonable approximation for the average depth of the body organs relative to a small source

on the surface of the body. 

If the small handheld source is carried in a worker’s pocket for 2000 h/yr, then the annual EDE

to the whole body could be 0.002 mSv (0.2 mrem), and the dose equivalent to a small area of

skin could be 0.2 mSv (20 mrem). If a commercial product containing 200 mCi source of 210Po

was carried in a worker’s pocket for the same number of hours per year, the annual EDE to the

whole body could be 0.8 mSv (80 mrem), and the dose equivalent to a small area of skin could

be 0.08 Sv (8 rem). These dose estimates are considered to be conservative because it is

unlikely that anyone would carry a source in their pocket for as much as 2000 h/yr (40 h/wk ×

50 wk/yr).

In the case of accidents, the following are considered: (1) a residential fire involving a static

eliminator brush, (2) a transportation fire involving an average daily shipment of 120 small

consumer devices containing a total of 2.2 GBq (60 mCi) of 210Po and 40 large commercial

devices containing a total of 74 GBq (2 Ci) of 210Po (see Section 4.1.4.1), and (3) a

manufacturer’s warehouse fire containing a number of devices equal to five daily shipments. A

release fraction of 0.1% is assumed for static eliminators involved in a fire. Based on these

assumptions and the generic accident methodology in Appendix A.1, the estimates of individual

dose from fires involving static eliminators are summarized as follows:

 For a person trying to escape from a residential fire or a neighbor trying to rescue a

person from a residential fire, the individual EDE from a single static eliminator brush

containing 19 MBq (500 Ci) could be 0.02 mSv (2 mrem).

 For a firefighter wearing a respirator at a residential fire, the individual EDE from a

single static eliminator brush containing 19 MBq (500 Ci) could be 5×105

 mSv

(0.005 mrem). The individual EDE could be 3×104

 mSv (0.03 mrem) in the case of a

worker who is involved in cleanup following the fire and who does not wear a respirator.
4–11

 For a firefighter wearing a respirator at a transportation fire, the individual EDE from 120

small consumer devices containing 2.2 GBq (60 mCi) could be 0.005 mSv (0.5 mrem). 

The individual EDE from 40 large commercial devices containing 74 GBq (2 Ci) could be

0.2 mSv (20 mrem). The individual EDEs could be 0.06 mSv (6 mrem) from the small

consumer devices and 2 mSv (200 mrem) from the large commercial devices in the

case of a worker who is involved in cleanup following the fire and who does not wear a

respirator.

 For a firefighter wearing a respirator at a warehouse fire, the individual EDE from 600

small consumer devices containing 11 GBq (300 mCi) could be 0.005 mSv (0.5 mrem). 

The individual EDE from 200 large commercial devices containing 0.37 TBq (10 Ci)

could be 0.2 mSv (20 mrem). The individual EDEs could be 0.03 mSv (3 mrem) from

the small consumer devices and 1 mSv (100 mrem) from the large commercial devices

in the case of a worker who is involved in cleanup following the fire and who does not

wear a respirator.

4.1.5 Summary

Table 4.1.1 presents the results of the potential radiological impacts on the public from the

distribution, use, and disposal of static eliminators containing 210Po. These results are based on

the annual distribution of 30,000 small consumer devices initially containing a total of 0.56 TBq

(15 Ci) and 10,000 large commercial devices initially containing a total of 19 TBq (500 Ci) of

210Po. The effective lifetime of the devices is assumed to be 1 year.

The radiological assessments for static eliminators done here are based on a sealed source

technology similar to that used in making smoke detectors (see Section 4.1.4). For the current

situation, which does not authorize disposal of the 210Po in the static eliminators as ordinary

waste, a collective EDE to the public of less than 0.001 person-Sv (0.1 person-rem) is

estimated for the small consumer devices and approximately 0.2 person-Sv (20 person-rem) for

the large commercial devices (see Table 4.1.1).

For an exemption, which would allow for disposal of the 210Po in the static eliminators as

ordinary waste, a total collective effective dose is estimated that is only slightly greater than that

for the current situation (see Table 4.1.1). For both the current situation and the exemption

considered here, an annual individual EDE of 0.02 mSv (2 mrem) or less is estimated during

disposal, distribution and transport, or routine use (see Table 4.1.1 and Section 4.1.4). A

maximum individual dose equivalent of 2 mSv (200 mrem) or less is estimated from accidents

involving fire or misuse by an individual who carries a device in a pocket for a long period of

time (2000 h/yr).
4–12

Table 4.1.1 Potential Radiation Doses From Static Eliminators Containing 210Po

Exposure Pathway

Individual Annual

Effective Dose

 Equivalent

(mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

COMMERCIAL DEVICES

Distribution and transport 2c 2×102

Routine use 0.1d 3

Disposal as ordinary trash

Landfills

Incinerators

<0.001e

0.01f

3×103

0.2

Misuse and accidents

Carrying in pocket

Fire

 80g

 200h

CONSUMER DEVICES

Distribution and transport 0.06c 6×104

Routine use <0.001d 2×102

Disposal as ordinary trash

Landfills

Incinerators

 <0.001e

 <0.001f

9×105

6×103

Misuse and accidents

Carrying in pocket

Fire

 0.2g

 6h

See following page for footnotes.
4–13

Footnotes to Table 4.1.1

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Collective doses are based on an annual distribution of 10,000 commercial devices initially

containing an average of 1.9 GBq (50 mCi) per device and 30,000 consumer products initially

containing an average of 19 MBq (500 Ci) per device. An effective lifetime of 1 year is

assumed for these devices. 

c

 Dose estimates apply to local parcel-delivery driver; dose estimates are considerably less for

terminal workers, long-haul semi-truck drivers, and members of public along truck routes (see

Section 4.1.4.1).

d

 Dose estimate applies to users of a commercial device initially containing 7.4 GBq (200 mCi)

or user of a consumer device initially containing 19 MBq (500 Ci) of 210Po (see Section 4.1.4.2). e

 Dose estimate applies to waste collectors at landfills; dose estimates are significantly less for

workers at landfills, future on-site residents, and off-site residents from well water ingestion. 

Dose estimates for off-site residents from inhalation are zero (0) because the static eliminators

and ion generators containing 210Po are assumed to remain intact during landfill disposal (see

Section 4.1.4.3).

f

 Dose estimate applies to workers at incinerators; dose estimates are significantly less for

waste collectors at incinerators and off-site members of the public near incinerator sites (see

Section 4.1.4.3).

g

 Dose estimate applies to user of a commercial device initially containing 7.4 GBq (200 mCi) or

user of a consumer product such as a static eliminator brush initially containing 19 MBq

(500 Ci) of 210Po; estimated dose equivalents for a small area of skin on the whole body are

0.08 Sv (8 rem) for commercial products and 0.2 mSv (20 mrem) for consumer products (see

Section 4.1.4.4).

h

 Dose estimates apply to inhalation exposure to a worker during cleanup following a fire

involving multiple devices; dose estimates are significantly less for inhalation by firefighters, a

person escaping a residential fire, or a neighbor attempting to rescue a person from a

residential fire (see Section 4.1.4.4).

14 Carbon-14 also is used sparingly in beta backscatter devices to measure paint layers with

thickness of 10 mils (0.12 m) or less.

4–15

4.2 Beta Backscatter and Transmission Devices

4.2.1 Introduction

A general license is granted in 10 CFR 31.5 to acquire, receive, possess, use, or transfer

byproduct material contained in devices designed and manufactured for a number of specific

purposes, including detecting, measuring, gauging, or controlling thickness or density. Included

in the general license are requirements for labeling, leak testing, instructions for use, and

proper storage or disposition of the device. The licensee is also subject to terms and conditions

set forth in 10 CFR 31.2 dealing with general license requirements, transfer of byproduct

material, reporting and recordkeeping, and inspection. Leak testing is required except for

devices containing only krypton, devices containing only tritium or not more than

3.7 megabecquerel (MBq) (100 microcurie (Ci)) of a beta- and/or gamma-emitting material or

0.37 MBq (10 Ci) of an alpha-emitting material, and devices held in storage in the original

shipping container prior to initial installation.

Beta backscatter and transmission devices use beta particles from a variety of sources to

measure the thickness or density of thin films and thin coatings on other materials. The beta

transmission devices are known to use sources of 85Kr, 90Sr, and 147Pm, while the beta

backscatter devices are known to use sources of 90Sr, 106Ru, 109Cd, 147Pm, 204Tl, and 210mBi.14

There are no limits in 10 CFR 31.5 on the amount of material that can be used in the devices,

but an applicant for a specific license to manufacture or initially transfer such devices for use

under 10 CFR 31.5 must demonstrate that these devices will meet certain requirements

contained in 10 CFR 32.51. These requirements are discussed below:

 The device can be safely operated by persons without training in radiological protection.

 Under ordinary conditions of handling, storage, and use of the device, the byproduct

material contained in the device will not be released or inadvertently removed from the

device, and it is unlikely that any person will receive in a 1-year period a total effective

dose equivalent (EDE) in excess of 5 millisieverts (mSv) (500 mrem), or the sum of the

deep-dose equivalent and committed dose equivalent to any individual organ or tissue

other than the lens of the eye in excess of 50 mSv (5 rem); an eye dose equivalent in

excess of 15 mSv (1.5 rem); or a shallow-dose equivalent in excess of 50 mSv (5 rem)

to the skin or to any extremity (i.e., hand, elbow, arm below the elbow, foot, knee, or leg

below the knee).

 Under accident conditions (such as fire and explosion) associated with handling,

storage, and use of the device, it is unlikely that any person would receive an external

dose equivalent or committed internal dose equivalent in excess of 0.15 Sv (15 rem) to

the whole body, head and trunk, active blood-forming organs, gonads, or lens of the

eye; 2 Sv (200 rem) to the hands and forearms, feet and ankles, or localized areas of

skin averaged over areas no larger than 1 cm2

; and 0.50 Sv (50 rem) to any other

organs.
4–16

Some beta backscatter and transmission devices are potential candidates for exemption from

the licensing requirements of 10 CFR 31.5. This assessment evaluates the potential radiation

doses that could result if the receipt, use, and transfer of certain of these devices were exempt

from licensing. The assumed conditions for this possible exemption are the use of sources

containing not more than 10 times a quantity of a byproduct material as defined in

10 CFR 30.71, Schedule B.

4.2.2 Description of Items Considered for Exempt Distribution

A beta backscatter and transmission device consists of the following basic components: (1) a

sealed radioactive source with a thin window to supply the beta particles, (2) a beta-particle

detector to provide the measurement signals, and (3) associated electronics required for signal

storage and analysis. The source is mounted in a housing that has an opening (beam port) to

allow beta particles to reach the sample. A protective cover (shutter) can be placed over the

beam port to shield workers and the operator from the source when the device is not in

operation. Interchangeable sources may be provided to meet specific measurement needs,

and during normal operations, the sources are either contained in storage containers or source

housing. Special tools are provided for removal and installation of the interchangeable sources,

and these tools, along with a separate storage container for each source, are designed to

minimize operator exposure.

Beta transmission devices are installed or mounted in a fixed configuration with an open

sample-slot between source and detector for items to be examined (ANSI N538–1979). 

Available information on these devices indicate that single, fixed sources are used and that

there is no added exposure from interchangeable sources. Beam port covers provided by the

manufacturer serve as shielding to reduce beta exposure during transport and installation. 

Workers may be exposed during normal operation to radiation transmitted through the source

housing or to the beta-particle beam from the device. The activity of sources in these devices

is known to be 19 MBq (500 Ci) of 85Kr, 19 to 56 MBq (500 to 1500 Ci) of 90Sr, and 67 MBq

(1800 Ci) of 147Pm (see Table 4.2.1). It is estimated that 200 devices containing 85Kr sources

are in use, with an additional 20 devices shipped annually. It is also believed that about

500 devices containing 90Sr sources and an equal number containing 147Pm sources are in use,

with an additional 50 of each shipped annually. The number shipped annually is based on an

assumed effective lifetime of 10 years for these devices.

Beta backscatter devices are constructed of a single housing in which both the source and the

detector are mounted (ANSI N538–1979). The device is simply positioned against or near the

sample to be tested. This technique (and device) can be used to examine thin coatings on

substrates that are too thick for beta-particle transmission. The device can be installed in a

fixed position or used as a portable, handheld instrument with the housing shielding the

operator from the source. A cap or cover over the sample end can be used as shielding when

the devices are not being operated. The activity of sources in these devices is known to be

0.19 MBq (5 Ci) of 90Sr, 0.74 MBq (20 Ci) of 106Ru, 22 MBq (600 Ci) of 109Cd, 2.2 to 15 MBq

(60 to 400 Ci) of 147Pm, 1.1 to 1.9 MBq (30 to 50 Ci) of 204Tl, and 5.6 MBq (150 Ci) of 210mBi

(see Table 4.2.1). The most widely used sources in beta backscatter devices appear to be

106Ru and 210mBi. Approximately 1000 to 1500 devices containing each of these radionuclides

were distributed over the past 10 years. Thus, it is assumed that about 3000 beta backscatter
4–17

devices are in use and that an additional 300 are shipped annually based on an effective

lifetime of 10 years for these devices.

The housings around the sources provide good shielding from beta particles in directions away

from the beam port. However, the radionuclides commonly used in the beta backscatter and

transmission devices decay by modes other than, or in addition to, emission of beta particles,

resulting in external exposure to photons. For example, three (109Cd/109mAg, 204Tl, and

210mBi/206Tl) decay by modes that include electron capture or internal conversion, both of which

produce K- and L-shell X-rays, and three (85Kr, 106Ru/106Rh, and 210mBi/206Tl) decay by

beta-particle emission, followed by emission of a gamma ray. Only two commonly used

radionuclides (90Sr/90Y and 147Pm) are pure beta-particle emitters, and these have associated

bremsstrahlung produced within the source housings. The source housings are assumed to be

made of light materials such as plastic and aluminum and to provide little or no shielding of

workers or operators to the bremsstrahlung, X-rays, or gamma rays from the devices.

The dose rate in the beta-particle beam from a beta backscatter and transmission device

appears to be very sensitive with regard to thickness of the thin window on the source supplying

the beta particles and size of the aperture used to shape the beta-particle beam from the

device. Thus, available data from manufacturer’s measurements of dose rates to skin in the

beta-particle beam at a distance of about 10 cm from the unshielded sources of existing

devices have been summarized in Table 4.2.1.

4.2.3 Summary of Previous Analyses and Assessments

There are no known previously published analyses or assessments of the radiation doses to

personnel operating beta backscatter and transmission devices. However, each applicant for a

specific license to manufacture or initially transfer these devices for use under 10 CFR 31.5 is

required to submit information to the Nuclear Regulatory Commission (NRC) to show that their

product meets the dose criteria summarized in Section 4.2.1.

4.2.4 Present Exemption Analysis

If a comparison is made of the activities of the sources currently being used in beta backscatter

and transmission devices with values equal to 10 times the quantities for various byproduct

materials from Schedule B, the most likely candidates for exemption appear to be 85Kr, 106Ru,

and 204Tl. The activity equal to 10 times a quantity of 85Kr, 106Ru, and 204Tl is 37 MBq

(1000 Ci), 0.37 MBq (10 Ci), and 3.7 MBq (100 Ci), respectively (see 10 CFR 30.71,

Schedule B). The radioactive half-life of 85Kr, 106Ru, and 204Tl is 10.7 years, 368.2 days, and

3.8 years, respectively (see Section 2.1).

To investigate potential radiation doses to the public for this potential exemption, it is assumed

that the following number of devices could be shipped each year under an exemption: 20 beta

transmission devices containing 37-MBq (1000-Ci) sources of 85Kr, 100 beta backscatter

devices containing 0.37-MBq (10-Ci) sources of 106Ru, and 100 beta backscatter devices

containing 3.7-MBq (100-Ci) sources of 204Tl. The useful or effective lifetimes of these devices

are assumed to be 10 years.
4–18

If the sources in the beta backscatter and transmission devices are replaced after one half-life

of radioactive decay, the total number of sources shipped in these devices or as replacement

sources for these devices could be 20 sources of 85Kr, 300 sources of 204Tl, and 1000 sources

of 106Ru. These estimates are based on the number of devices currently being distributed for

use under the general license granted in 10 CFR 30.15, but the removal of some of the

requirements of a general license could ultimately increase both the number of devices and the

number of sources distributed annually.

4.2.4.1 Distribution and Transport

The potential radiation doses from sources shipped with beta backscatter and transmission

devices or as replacement sources for these devices are considered in this section, using the

generic distribution methodology in Appendix A.3. In applying this methodology, it is assumed

that the sources are shipped primarily by a parcel-delivery service, and that a driver in a large

van picks up the sources and takes them to a local terminal for shipment to other local

terminals for delivery to customers. A typical shipment from a manufacturer or supplier is

assumed to consist of either a single source of 85Kr containing 37 MBq (1000 Ci), four sources

of 106Ru containing 0.37 MBq (10 Ci) each, or a single source of 204Tl containing 3.7 MBq

(100 Ci). It is further assumed that (1) a single local parcel-delivery driver may pick up all of

the 85Kr, 106Ru, or 204Tl sources from a single supplier, (2) the sources are transported by semitruck between local terminals, and (3) the sources pass through an average of four regional

terminals before reaching their final destination. Radiation exposure to workers at both local

and regional terminals is considered to be similar to the radiation exposure to workers at a large

warehouse (see Appendix A.3).

Based on the above assumptions and the generic methodology of Appendix A.3, the annual

EDE could be 2×104

 mSv (0.02 mrem) to the local parcel-delivery driver who picks up the

20 sources of 85Kr, 1×104

 mSv (0.01 mrem) to the local parcel-delivery driver who picks up the

300 sources of 204Tl, and 0.004 mSv (0.4 mrem) to the local parcel-delivery driver who picks up

the 1000 sources of 106Ru. Individual doses would be less to other drivers, terminal workers,

and members of the public along the truck routes. The annual collective EDEs to all truck

drivers, terminal workers, and members of the public along the truck routes are estimated to be

3×106

 person-Sv (3×104 person-rem) for the 85Kr sources, 8×105 person-Sv

(8×103

 person-rem) for the 106Ru sources, and 2×106 person-Sv (2×104 person-rem) for the

204Tl sources. These dose estimates are very conservative because the generic distribution

methodology does not account for the shielding of sources shipped in the beta backscatter and

transmission devices or the shielding of the shipping containers for the replacement sources.

4.2.4.2 Routine Use

The potential doses from routine use of beta backscatter and transmission devices distributed

under the potential exemption are assumed to be the same as those from devices being

distributed under the current regulatory scheme. However, there could be differences

depending on the particular design and other requirements that might be imposed on the

manufacturers or suppliers of these devices.
4–19

4.2.4.2.1 Beta Transmission Devices

The beta transmission devices with 85Kr sources are small, low-maintenance devices designed

for use in gauging films with thicknesses of 1 cm or less. These devices are typically used on a

factory floor, being permanently and rigidly attached to the manufacturing equipment, and the

immediate vicinity of the source is occupied only occasionally by factory workers. The

performance of the devices is not affected by changes in humidity or atmospheric pressure, but

dust accumulation may degrade performance and require careful cleaning of the fragile thinwindow source and the beam path between the source and detector.

To estimate potential doses from beta-particle irradiation of the hands (fingers) during routine

use, the following factors were considered: (1) the initial installation of the source within the

fixed source housing at the point of use, and (2) the periodic cleaning of dust from the source

and the beam path between the source and detector within the device. From actual

measurements (see Table 4.2.1), it is estimated that the dose rate to skin at a distance of

10 cm from a shielded 37-MBq (1000-Ci) source of 85Kr before its installation in the fixed

source housing is 1.6 mSv/h (160 mrem/h), and the dose rate to skin in the beta-particle beam

at a distance of 10 cm from an unshielded 3.7-MBq (1000-Ci) source of 85Kr is 5.4 mSv/h

(540 mrem/h). For the initial installation of the source in the fixed source housing, it is

estimated that (1) the worker’s fingers are actually in contact with the shielded source for only

10 minutes, (2) the worker’s fingers are about 10 cm from the shielded source during this time,

(3) the dose rate to the skin is 1.6 mSv/h (160 mrem/h), and (4) the beta-particle dose to the

hands (fingers) is about 0.3 mSv (30 mrem). For the periodic cleaning of the source and the

beam path between the source and detector within the device, it is estimated that (1) the time

required is 5 minutes per month (1 h/yr), (2) the user’s fingers are in the beta-particle beam at a

distance of about 10 cm from the unshielded source during this time, (3) the beta-particle dose

to skin from the unshielded source is 5.4 mSv/h (540 mrem/h), and (4) the beta-particle dose to

the hands (fingers) is about 5 mSv/yr (500 mrem/yr). If the irradiated area of the hands

(fingers) is assumed to be 10 cm2

 or less, then the EDE from beta-particle irradiation of the

hands is small compared to the following estimates of the EDE from photon irradiation of the

whole body.

To estimate the potential doses from photon irradiation of the whole body during routine use, an

individual was considered to be exposed at an average distance of 2 meters from a 37-MBq

(1,000-Ci) source of 85Kr for 2000 h/yr. If the assumption is no shielding by the source holder

for bremsstrahlung, X-rays, or gamma rays from the device, the photon dose rate as calculated

using CONDOS (Computer Codes, O’Donnell et al., 1975) (see Appendix A.3) is about 5 nSv/h

(0.5 rem/h), and the annual EDE to the individual would be about 0.01 mSv (1 mrem). The

collective EDE over the first year of routine use would be about 4×10-4 person-Sv

(4×10-2 person-rem) if two workers are exposed for 2000 h/yr at an average distance of

2 meters from each of the 20 sources distributed annually. The total collective dose equivalent

over the expected 10-year lifetime of the sources and the devices containing these 85Kr sources

would be about 3×103

 person-Sv (0.3 person-rem).

4.2.4.2.2 Beta Backscatter Devices

The beta backscatter devices with 106Ru or 204Tl sources are used primarily to measure the

thicknesses of coatings on various substrates in either a factory setting or a research

laboratory. The devices can be mounted in a fixed position or used as a portable instrument. 
15 The dose rate per unit activity is assumed to be about the same for both 106Ru/106Rh and 204Tl

(see National Council on Radiation Protection and Measurements (NCRP) 112, Table 5.1), and

the dose rate to skin in the beta-particle beam at 10 cm from an unshielded 204Tl source per unit

activity is assumed to be about 14 mSv/h per MBq (0.5 mrem/h per Ci) (see Table 4.2.1). 

4–20

In use, most of the beta particles are stopped in the housing surrounding the source. Those

that escape through the housing with energies less than 70 keV will not penetrate to the

sensitive layer of the skin. However, potentially significant radiation doses to the hand can

occur from exposure to the beta-particle beam from the device.

To estimate potential doses from beta-particle irradiation of the hands (fingers) during routine

use, the following factors were considered: (1) the placement and removal of test samples over

the measuring opening (beam port) of the devices, and (2) the periodic installation of new 106Ru

and 204Tl sources over the 10-year effective lifetime of the devices. From actual measurements

(see Table 4.2.1), it is estimated that the dose rate to skin in the beta-particle beam is 50 Sv/h

(5 mrem/h) at 10 cm from an unshielded 0.37-MBq (10-Ci) source of 106Ru and 0.5 mSv/h

(50 mrem/h) at 10 cm from an unshielded 3.7-MBq (100-Ci) source of 204Tl.15 For the

placement and removal of test samples over the beam ports to these devices, the very

conservative assumption was made that only one worker (user) performs this operation using

tweezers or forceps, so that the distance of the user’s hands from an unshielded source is

about 10 cm, each operation takes about 10 seconds, and there are 100 operations per work

day and 250 work days per year. Thus, the estimated annual dose equivalent from betaparticle irradiation of the hands (fingers) could be about 4 mSv (400 mrem) from the unshielded

106Ru source and 40 mSv (4 rem) from the unshielded 204Tl source. For the periodic installation

of the new sources, it is estimated that (1) the time required is 10 minutes, (2) the installer’s

(user’s) hands will be about 10 cm from the unshielded sources during this time, and (3) the

beta-particle dose to the hands (fingers) from the 106Ru and 204Tl sources is about 0.1 mSv

(10 mrem) and 0.3 mSv (30 mrem), respectively. If the irradiated area of the hands (fingers) is

assumed to be 10 cm2

 or less, then the EDE to a user from beta-particle irradiation of the hands

is small compared to the following estimates of the EDE from photon irradiation of the whole

body.

To estimate potential doses from photon irradiation of the whole body during routine use, it is

assumed that a fixed or portable instrument containing either a 0.37-MBq (10-Ci) of 106Ru or a

3.7-MBq (100-Ci) source of 204Tl is located at an average distance of 30 cm from a user’s body

for about 500 h/yr during sample testing. If the assumption is no shielding of bremsstrahlung,

X-rays, and gamma rays from the device, the photon dose rates as calculated using CONDOS

(Computer Codes, O’Donnell et al., 1975) (see Appendix A.3) for the 106Ru and 204Tl sources

are about 100 nSv/h (10 rem/h) and 20 nSv/h (2 rem/h), respectively. If radioactive decay of

the sources over their first year of usage is considered, the annual EDE to users of the devices

containing the 106Ru and 204Tl sources are about 0.04 mSv (4 mrem) and 0.01 mSv (1 mrem),

respectively. For the 1000 sources of 106Ru distributed annually and used for about 1 year in

these devices, the total collective effective dose is estimated be 4×102

 person-Sv

(4 person-rem). For the 300 sources of 204Tl distributed annually and used for about 3 years in

these devices, the total collective EDE is estimated be 7×10-3 person-Sv (0.7 person-rem).
4–21

4.2.4.3 Disposal

Although the 85Kr, 106Ru, and 204Tl sources in beta backscatter and transmission devices are

discarded as radioactive waste by the manufacturer or distributor under the current regulatory

scheme, for the purposes of evaluating a possible exemption, all sources distributed are

assumed to be disposed as ordinary waste, as there are usually no controls over disposal under

an exemption.

To estimate potential doses from disposal of these sources as ordinary waste in landfills and

incinerators, the generic disposal methodology in Appendix A.2 is used, and the very

conservative assumption is made that all sources are disposed as ordinary waste after one halflife of radioactive decay. Thus, the 20 sources of 85Kr would contain a total of 0.37 GBq

(10 mCi) at the time of disposal, the 1000 sources of 106Ru would contain a total of 0.19 GBq

(5 mCi), and the 300 sources of 204Tl would contain a total of 0.56 GBq (15 mCi). It is assumed

that 80% of the sources are disposed in a landfill and 20% go to incineration.

4.2.4.3.1 Landfill Disposal

In applying the methodology of Appendix A.2 to disposal at landfills, it is assumed that the 106Ru

and 204Tl sources normally remain intact and that waste collectors or workers at landfills do not

touch these discarded sources with their hands. For 85Kr sources, it is assumed that half of the

activity is released to air during disposal at landfills, due to rupture of the sealed sources, and

only half of the activity is retained in the intact sealed sources. Based on these assumptions

and the relatively short half-lives of 85Kr, 106Ru, and 204Tl, the only significant exposure pathway

is external whole-body irradiation of waste collectors and landfill workers by photons from these

sources. Since the number of sources disposed annually is less than 3500, the assumed

number of landfills, the applicable dose-to-source ratios (DSRs) from Appendix A.2 are

multiplied by the ratio of 3500 to the number of items. (See discussion in Appendix A.2

addressing disposal of a limited number of items).

Estimates of potential individual and collective doses from landfill disposal of 85Kr sources are

summarized as follows:

 The annual EDE to waste collectors could be 3×105

 mSv (0.003 mrem). For workers at

landfills, off-site members of the public near landfills, and future on-site residents, the

individual doses would be less.

 The collective EDE could be 7×107

 person-Sv (7×105 person-rem), due almost entirely

to exposure to waste collectors and landfill workers.

Estimates of individual and collective doses from landfill disposal of 106Ru sources are

summarized as follows:

 The annual EDE to a waste collector could be 3×105

 mSv (0.003 mrem). For workers

at landfills, off-site members of the public near landfills, and future on-site residents, the

individual doses would be less.

 The collective EDE could be 5×105

 person-Sv (5×103 person-rem), due almost entirely

to exposure to waste collectors and landfill workers.
4–22

Estimates of individual and collective doses from landfill disposal of 204Tl sources are

summarized as follows:

 The annual EDE to a waste collector, workers at landfills, off-site members of the public

near landfills, and future on-site residents, would be less than 1×105

 mSv

(<0.001 mrem).

 The collective EDE could be 7×107

 person-Sv (7×105 person-rem), due almost entirely

to exposure to waste collectors and landfill workers.

4.2.4.3.2 Incineration

In applying the methodology of Appendix A.2 to incineration, the assumption is made that there

is no exposure to waste collectors by either inhalation or ingestion of byproduct material from

the sealed sources, but it is assumed that all of the various pathways of exposure to workers

and off-site members of the public are fully operative. An incineration fraction of 0.2 is

assumed for 85Kr and 204Tl. Since the number of items incinerated is less than the number of

incinerator facilities assumed, a correction to the DSRs in Appendix A.2 is applied. (See

discussion in Appendix A.2 addressing incineration of a limited number of items).

Estimates of potential individual and collective doses from incineration of 85Kr sources are

summarized as follows:

 The annual EDE to waste collectors could be 3×105

 mSv (0.003 mrem). For workers at

incinerators and off-site members of the public near incinerators, the individual doses

would be less.

 The collective EDE could be 1×107

 person-Sv (1×105 person-rem), due almost entirely

to exposure to waste collectors at incinerators.

Estimates of potential individual and collective doses from incineration of 106Ru sources are

summarized as follows:

 The annual EDE to waste collectors could be 4×105

 mSv (0.004 mrem). For workers at

incinerators and off-site members of the public near incinerators, the individual doses

would be less.

 The collective EDE could be 6×106

 person-Sv (6×104 person-rem), due almost entirely

to exposure to waste collectors at incinerators.

Estimates of potential individual and collective doses from incineration of 204Tl sources are

summarized as follows:

 The annual EDE to waste collectors, workers at incinerators and off-site members of the

public near incinerators, would be less than 1×105

 mSv (<0.001 mrem).

 The collective EDE could be 1×107

 person-Sv (1×105 person-rem), due almost entirely

to exposure to waste collectors at incinerators.
4–23

4.2.4.4 Accidents and Misuse

Three different pathways of exposure during accidents and misuse are considered in this

section. The first pathway involves exposure to radioactive materials released from sources

involved in a transportation fire. The second pathway involves exposure to radioactive material

leaking from a ruptured source. The third pathway involves exposure to a waste collector or

other person who finds a discarded source and carries it in his or her pocket for 3 hours before

storing it in an out-of-the-way place.

In the case of an accident involving fire, a transportation fire is considered that involves a typical

shipment of either one source of 85Kr containing 37 MBq (1000 Ci), four sources of 106Ru

containing 0.37 MBq (10 Ci) each, or one source of 204Tl containing 3.7 MBq (100 Ci) (see

Section 4.2.4.1). A release fraction of 100% is assumed for the 85Kr source, and a release

fraction of 0.1% is assumed for the 106Ru or 204Tl sources. Based on these assumptions and

the generic accident methodology in Appendix A.1, the estimates of individual dose from

radioactive materials released from these sources are summarized as follows:

 For a firefighter at a transportation fire, the EDE from submersion in the 85Kr released

from a single 37-MBq (1000-Ci) source could be 2×105

 mSv (0.002 mrem). 

 For a firefighter who wears a respirator at a transportation fire and for a worker who is

involved in cleanup following the fire and who does not wear a respirator, the EDE from

submersion in and inhalation of the 106Ru released from four 0.37-MBq (10-Ci) sources

would be less than 1×105

 mSv (<0.001 mrem).

 For a firefighter who wears a respirator at a transportation fire, the EDE from

submersion in and inhalation of the 204Tl released from a single 3.7-MBq (100-Ci)

source would be less than 1×105

 mSv (<0.001 mrem). For a worker who is involved in

cleanup following the fire and who does not wear a respirator, the individual dose from

inhalation could be 2×105

 mSv (0.002 mrem).

In the case of source leakage, potential radiation doses to the user of the source and to a waste

collector were considered. Potential radiation doses to users and waste collectors from

external exposure by submersion in the 85Kr from a ruptured source would be small compared

to those from internal exposure by ingestion or inhalation of either 106Ru or 204Tl. To estimate

the radiation dose to users of a ruptured 106Ru or 204Tl source, the generic accident

methodology developed in Appendix A.1 for ingestion of radioactivity following a spill of a

radioactive material in the form of a powder was used. First, it was assumed that 10% of the

material was deposited on the skin of an individual and, second, that 0.1% of this deposited

material would be ingested before bathing removed the material from the body. Based on

these assumptions and the general accident methodology of Appendix A.1 (see Table A.1.8),

the EDE to a user could be 3×104

 mSv (0.03 mrem) due to leakage from either a 0.37-MBq

(10-Ci) source of 106Ru or 3.7-MBq (100-Ci) source of 204Tl. 

To estimate the radiation dose to waste collectors, the generic disposal methodology in

Appendix A.2 (see Table A.2.1) was used. Using this methodology, the dose-to-source ratios

are divided by the number of landfills in the United States, multiplied by 3500 (i.e., the

estimated number of U.S. landfills) and then by the amount of activity in the 106Ru or 204Tl

sources (i.e., 370 kBq and 3700 kBq (10 Ci and 100 Ci), respectively). Thus, the EDE to the
16 The dose equivalent from beta-particle irradiation of skin by a discarded 0.19-MBq (5-Ci)

source of 106Ru and a discarded 1.9-MBq (50-Ci) source of 204Tl is estimated to be 0.03 Sv

(3 rem) and 0.4 Sv (40 rem), respectively.

4–24

waste collector would be less than 1×105

 mSv (<0.001 mrem) due to leakage from either the

106Ru source or the 204Tl source.

In the case of misuse, it is assumed that a waste collector or other person who finds a

discarded 85Kr source will carry it in his or her pocket for 3 hours before storing it in an out-ofthe-way place. The distance from the source to the surface of the body is assumed to be

10 cm during this time. If it is further assumed that the 37-MBq (1000-Ci) source has decayed

for one half-life before disposal (see Section 4.2.4.3), then the amount of radioactivity in the

discarded 85Kr source would be 19 MBq (500 Ci). Based on calculations using MicroShield

(Computer Codes, Grove Engineering, 1996), the EDE rate from photon irradiation of the whole

body is estimated to be 0.5 Sv/h (50 rem/h) for the discarded 19-MBq (500-Ci) source of

85Kr, and the dose equivalent rate from photon irradiation of the skin is estimated to be

70 Sv/h (7 mrem/h). In addition, the dose equivalent rate from beta-particle irradiation of skin

by the discarded 19-MBq (500-Ci) source of 85Kr is estimated to be 1.4 Sv/h (140 rem/h),

assuming a cloth thickness of 0.7 mm and a density of 0.4g/cm3

(VARSKIN MOD2) (Computer

Codes, Durham, 1992). 

The dose equivalent rates to skin are based on calculations for a separation distance of 1 cm

between the source and skin. EDE rates are based on calculations at a tissue depth of 10 cm,

which is considered a reasonable approximation for the average depth of the body organs

relative to a small source on the surface of the body (Refer to Appendix A.4). If the irradiated

area of the skin in close contact with the source is about 10 cm2

 and the organ weighting factor

for skin is 0.01 (International Commission on Radiological Protection (ICRP) 60), then the total

EDE from both photons and beta-particle irradiation of the whole body is estimated to be

0.02 mSv (2 mrem). However, the dose equivalent to the small 10-cm2

 area of skin on the

body’s surface from the discarded 19-MBq (500-Ci) source of 85Kr could be as much as 4 Sv

(400 rem), assuming minimal shielding by articles of clothing or other materials between the

source and skin surface.16

Reports of leaking sources—submitted by licensees under the requirements of

10 CFR 31.5(c)(5) to the NRC between 1990 and 1996—(NRC, Databases, NMED, Reports),

indicate one instance of a leaking source in a density gauge and three instances of leaking

sources in other unspecified gauges containing byproduct materials used in beta backscatter

and transmission devices. A leak test requirement for the 85Kr, 106Ru, and 204Tl sources

considered in this assessment is unlikely, because sources containing only krypton or not more

than 3.7 MBq (100 Ci) of a beta- and/or gamma-emitting material are already exempt from

leak test requirements (see Section 4.2.1). However, there are other circumstances under

10 CFR 31.5(c)(5) and (10) for which actions are required to control exposures to users and the

public that would not be applicable under this potential exemption. For example, labeling

requirements for exempt products are different than for generally licensed devices. 
4–25

4.2.5 Summary

Table 4.2.2 presents the results of the analysis of potential radiological impacts for an

exemption that would allow beta backscatter and transmission devices containing not more than

37 MBq (1000 Ci) of 85Kr, 0.37 MBq (10 Ci) of 106Ru, or 3.7 MBq (100 Ci) of 204Tl. Radiation

doses estimated in this assessment are based on typical designs of devices distributed under

the requirements applicable to distributors and applicants for license to distribute such devices

for use under 10 CFR 31.5. The details of the designs are important to ensuring control of

exposure. For the radiation doses resulting under an exemption of these devices to be

comparable to those estimated, similar controls over the distributors would be necessary to

ensure that the designs of the devices are comparable in minimizing exposure to users and the

members of the public.

The annual EDEs to individuals could be about 0.004 mSv (0.4 mrem) for distribution and

transport, 0.04 mSv (4 mrem) for routine use of a beta transmission or backscatter device, and

4×105

 mSv (0.004 mrem) for disposal in landfills and by incineration. For all of these activities

combined, the collective EDE to all users and members of the public is estimated to be

5×102

 person-Sv (5 person-rem). This collective dose estimate assumes an annual distribution

of 20 sources of 85Kr initially containing 37 MBq (1000 Ci) each, 1000 sources of 106Ru initially

containing 0.37 MBq (10 Ci) each, and 300 sources of 204Tl initially containing 3.7 MBq

(100 Ci) each. The effective lifetime of the sources is assumed to be equal to one half-life of

the radioactive decay, and the effective lifetime of the devices using these sources is assumed

to be 10 years. For accidents involving fire or leakage of radioactive material from a source,

the EDE to an individual could be 3×104

 mSv (0.03 mrem) due to leakage from either a

0.37-MBq (10-Ci) source of 106Ru or 3.7-MBq (100-Ci) source of 204Tl.

For the situation in which the sources can be replaced by the user and disposed of as ordinary

waste, however, a potentially serious problem has been identified resulting from the loss of

control and accountability over the discarded sources. If a waste collector or other person

carried a discarded source in a pocket for as few as 3 hours, the dose equivalent to a small

area of skin could be as much as 4 Sv (400 rem) (see footnotes to Table 4.2.2), which could

cause minor radiation burns to the skin (Potten, 1985). Thus, some control and accountability

over discarded sources should probably be maintained in any potential exemption to prevent

such potential misuse from occurring.
4–26

Table 4.2.1 Summary of Radionuclides, Source Activities, and Dose Rates Near

Beam Port of Example Existing Backscatter and Transmission Devices

Radionuclide

Source Activity 

(Ci)a

Dose Rateb

(mrem/h)c

BETA BACKSCATTER DEVICES

90Sr

106Ru

109Cd

147Pm

204Tl

210mBi

5

5

20

600

60

400

30

50

150

45

58

9

14

16

22

BETA TRANSMISSION DEVICES

85Kr

90Sr

147Pm

500

500

1,500

1,800

270d

80

62

12

a

 1 Ci = 0.037 MBq.

b

 Dose rate to skin in beta-particle beam at a distance of 10 cm from unshielded source (i.e., no

protective cap over beam port or shutter opened). 

c

 1 mrem/h = 0.01 mSv/h.

d

 Dose rate to skin at 10 cm from a shielded 19-MBq (500-Ci) source of 85Kr—before it is

installed in a fixed source holder—is about 0.8 mSv/h (80 mrem/h).
4–27

Table 4.2.2 Potential Radiation Doses From Beta Backscatter and Transmission

Devices Using 85Kr, 106Ru, and 204Tl Sources

Exposure Pathway

Individual Annual

Effective Dose 

Equivalent

(mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport 0.4c 0.008

Routine use

 Transmission devices

 Backscatter devices

 1d

 4e

 

0.3

5

Disposal as ordinary waste

 Landfills

 Incinerators

0.003f

0.004g 

0.005

<0.001

Accidents and misuse

 Accidents involving fire

 Source leakage

 Carrying source in pocket

0.002h

0.03i

2j

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Refer to text for time period of collection dose assessment.

c

 Dose estimate applies to local parcel-delivery driver, who is assumed to pick up the 1,000

sources of 106Ru from a single supplier; dose estimates are less for other truck drivers, terminal

workers, and members of public along truck routes (see Section 4.2.4.1).

d

 Dose estimate applies to routine users of beta transmission devices; dose estimate for annual

dose equivalent to hands (fingers) from beta particles is 5 mSv (500 mrem) (see

Section 4.2.4.2.1).

e

 Dose estimate applies to routine users of beta backscatter devices; dose estimate for annual

dose equivalent to hands (fingers) from beta particles is 0.04 Sv (4 rem) (see

Section 4.2.4.2.2).

f

 Dose estimate applies to waste collectors at landfills; dose estimates are less for workers at

landfills, off-site members of the public near landfills, and future on-site residents (see

Section 4.2.4.3.1).

g

 Dose estimate applies to waste collectors at incinerators; dose estimates are less for workers

at incinerators and off-site members of the public near incinerator sites (see Section 4.2.4.3.2).

h

 Dose estimate applies to the submersion of a firefighter in the 85Kr from a single 37-MBq

(1,000-Ci) source; dose estimates are less to a firefighter or worker during cleanup following a

transportation fire involving 106Ru or 204Tl sources (see Section 4.2.4.4).

i

 Dose estimates apply to internal exposure to a user of a beta backscatter device containing a

leaking source of either 106Ru or 204Tl (see Section 4.2.4.4).

j

 Dose estimate applies to whole-body irradiation of a person who carries a discarded 19-MBq

(500-Ci) source of 85Kr in his or her pocket for 3 hours; dose estimate for a small area of skin

on the whole body is 4 Sv (400 mrem) (see Section 4.2.4.4).

4–29

4.3 Electron Capture Detectors for Gas Chromatographs

4.3.1 Introduction

A general license is granted in 10 CFR 31.5 to acquire, possess, use, or transfer byproduct

material contained in devices designed and manufactured for measuring chemical composition,

either qualitatively or quantitatively. Included in the general license are requirements for

labeling, leak testing, and proper storage or disposition of the device. The licensee is also

subject to terms and conditions set forth in 10 CFR 31.2 dealing with general license

requirements, transfer of byproduct material, reporting and recordkeeping, and inspection. 

Leak testing is required except for (1) devices containing only krypton, (2) devices containing

only tritium (3

H) or not more than 3.7 megabecquerel (MBq) (100 microcurie (Ci)) of a betaand/or gamma-emitting material or 0.37 MBq (10 Ci) of an alpha-emitting material, and (3)

devices held in storage in the original shipping container prior to initial installation.

Electron capture detectors (ECDs) are used to identify molecules in the effluent stream from

gas chromatographs. The electrons are typically provided by low-energy beta particles from 3

H

sources containing not more than 11 GBq (300 mCi) or 63Ni sources containing not more than

0.74 GBq (20 mCi). There are no limits in 10 CFR 31.5 on the amount or kind of byproduct

materials that can be used in ECDs, but an applicant for a specific license to manufacture or

initially transfer ECDs for use under 10 CFR 31.5 must demonstrate that the devices will meet

certain requirements contained in 10 CFR 32.51. These requirements are described below:

 The device can be safely operated by persons without training in radiological protection.

 Under ordinary conditions of handling, storage, and use of the device, the byproduct

material contained in the device will not be released or inadvertently removed from the

device, and it is unlikely that any person will receive in any 1-year period a total effective

dose equivalent (EDE) in excess of 5 millisieverts (mSv) (500 mrem), or the sum of the

deep-dose equivalent and committed dose equivalent to any individual organ or tissue

other than the lens of the eye in excess of 50 mSv (5 rem); an eye dose equivalent in

excess of 15 mSv (1.5 rem); or a shallow-dose equivalent in excess of 50 mSv (5 rem)

to the skin or to any extremity (i.e., hand, elbow, arm below the elbow, foot, knee, or leg

below the knee).

 Under accident conditions (such as fire and explosion) associated with handling,

storage, and use of the device, it is unlikely that any person would receive an external

dose equivalent or committed internal dose equivalent in excess of 0.15 Sv (15 rem) to

the whole body, head and trunk, active blood-forming organs, gonads or lens of the eye;

2 Sv (200 rem) to the hands and forearms, feet and ankles, or localized areas of skin

averaged over areas no larger than 1 cm2

; and 0.50 Sv (50 rem) to any other organs.

Some ECDs are potential candidates for exemption from the general licensing requirements of

10 CFR 31.5. This assessment evaluates the potential radiation doses that could result if the

receipt, possession, use, and transfer of certain of these products were exempt from licensing. 

The assumed conditions for this possible exemption are ECDs containing not more than

0.74 GBq (20 mCi) of 63Ni or 11 GBq (300 mCi) of 3

H per device.
4–30

4.3.2 Description of Items Considered for Exempt Distribution

Gas chromatographs are used to separate a gas mixture in order to identify the various

components and their concentrations (Littlewood, 1970). To enhance sensitivity and thus

further refine the analysis, the effluent stream from a gas chromatograph is passed through

other devices. Mass spectrometry is perhaps the most highly evolved technique, providing

identification by molecular mass after molecular ionization. The ECD is another commonly

used, negative-ion-based detector (Zlatkis and Poole, 1981).

The analyte in a carrier gas (ionization potential 15 eV) such as helium, argon, or nitrogen is

passed through the ECD. The electrons from low-energy beta-particle sources are thermalized

in the carrier gas, and produce negative ions of the analyte at a rate that depends on the

electron capture cross-section for different molecules or molecular subunits. The change in

electron current while passing through the gas stream is monitored, and a variety of signal

processing techniques can be used to detect concentrations of one part in 1014 to 1016. The

few molecular compounds that react rapidly with thermalized electrons are mostly compounds

that are either highly toxic or otherwise environmentally objectionable (e.g., pesticides, nitro

compounds, chlorofluorocarbons) (Lovelock, 1982). Thus, such high sensitivities make the

combination of a gas chromatograph and ECD important in environmental sampling for very

small amounts of pollutants.

The radioactive sources usually consist of either a thin layer of titanium tritide or 63Ni on a

metallic foil in the form of a cylindrical sleeve through which the gas stream flows. The foils are

mounted in basically solid, cylindrical containers with small inlet and outlet holes or tubes for the

analyte. The basic unit holding the radioactive source has a stainless steel body, plastic parts

for insulation between metal components, and ceramic parts for electrical leads. Although the

shape and size of the components vary among manufacturers, the basic devices are

approximately fist-sized” or smaller. The components of the device are assembled by the

manufacturer using special securing screws or other techniques to eliminate easy access to the

radioactive source. The ECD has no moving parts that might damage the foil or require

frequent servicing.

4.3.3 Summary of Previous Analyses and Assessments

There are no known previously published analyses or assessments of the radiation doses to

personnel operating gas chromatographs using ECDs with 63Ni sources. However, each

applicant for a specific license to manufacture or initially transfer these devices for use under

10 CFR 31.5 is required to submit information to the Nuclear Regulatory Commission (NRC) to

show that its product meets the criteria summarized in Section 4.3.1.

4.3.4 Present Exemption Analysis

NRC data indicate that about 200 sources initially containing 11 GBq (300 mCi) each of 3

H and

2000 sources initially containing 0.74 GBq (20 mCi) each of 63Ni could be distributed annually in

ECDs for use with gas chromatographs under the general license in 10 CFR 31.5. This

corresponds to an annual distribution of about 1.5 TBq (40 Ci) of 63Ni and 2.2 TBq (60 Ci) of 3

H. 

For purposes of this analysis, it is assumed that the same amount of 63Ni and 3

H could be

distributed under this potential exemption, but the removal of some requirements of a general
4–31

license could ultimately increase the amount of both 63Ni and 3

H distributed annually in the

ECDs for use with gas chromatographs.

The ECDs may be used in either fixed or portable gas chromatographs, and they may be

heated during normal operations to temperatures up to 400C in fixed instruments and 200C

in portable instruments. Tests by manufacturers at temperatures above these norms show that

63Ni source foils should retain their physical integrity and generate no leakage of radioactive

material. The nonleakage of 63Ni from ECDs under normal usage, assumed earlier by Howley

et al. (1970), was confirmed in subsequent tests by Carlton et al. (1975). Operation at elevated

temperatures does increase the leakage rate of 3

H from the titanium tritide foils to values

significantly above those at room temperature; this is discussed further below. Safety features

on the ECDs prevent temperature excursions above those required for normal operations.

The 3

H atoms in a titanium tritide foil will diffuse to the foil surface and escape at a rate that

depends on the foil temperature (Howley et al., 1970; Taylor, 1962). For example, an 11-GBq

(300-mCi) foil of titanium tritide will lose approximately 0.19 MBq/day (5 Ci/day) of 3

H at room

temperature, 1.9 MBq/day (50 Ci/day) of 3

H at 250C, and 3H at greatly accelerated rates at

higher temperatures (Taylor, 1962). There is an additional loss of about 1.9 MBq/day

(50 Ci/day) as a result of radioactive decay of 3

H with a half-life of 12.28 years (see Table

2.1.1). Due to both radioactive decay and diffusion of 3

H from a foil source during normal

operation of an ECD, the effective lifetime of a source foil containing 3

H would be smaller than

that for a source foil containing 63Ni with a 100.1-year half-life (see Table 2.1.1) and no

measurable leakage (Carlton et al., 1975). Thus, the effective lifetime of a 3

H source in an ECD

is assumed to be about 5 years, and the effective lifetime of a 63Ni source in an ECD is

assumed to be about 10 years.

The wall materials of the ECDs surrounding the source foils are generally thick enough to stop

all beta particles from 3

H or 63Ni, since the range of even the most energetic 66-keV beta

particles from 63Ni is less than 0.1 mm in iron or plastic. In addition, these beta particles are not

energetic enough to contribute to the shallow-dose equivalent (or dose equivalent at 0.007 cm

in skin), even in close proximity to a source. Radiative energy loss by the beta particles

(bremsstrahlung) also comprises a faction less than 4×103

, even for the most energetic 66-keV

beta particles from 63Ni. The low radiation fields expected around the ECDs have been

confirmed by measurements that show no detectable radiation at the surface of an ECD

containing a 7.4-GBq (200-mCi) source of 3

H and an exposure rate of 34 nanocoulomb

(nC)/kg-h (0.13 milliroentgen (mR)/h) at the surface of an ECD containing a 0.37-GBq (10-mCi)

source of 63Ni (Howley et al., 1970).

In Section 2.15.5, the dose equivalent rates due to bremsstrahlung at distances of 30 cm and

1 meter from a 0.37-GBq (10-mCi) source of 63Ni using CONDOS (Computer Codes, O’Donnell

et al., 1975) were estimated. At 30 cm, the dose rate was estimated to be 7×106

 mSv/h

(7×104

 mrem/h), and at 1 meter, the dose rate was estimated to be 5×107 mSv/h

(5×105

 mrem/h). However, shielding of the 63Ni source by its protective housing, the external

case of the detector, and various other detector components will reduce the dose rates from the

63Ni source to essentially zero (0).

These dose rates are not inconsistent with the above measured value of 34 nC/kg-h

(0.13 mR/h) at the surface of an ECD containing a 0.37-GBq (10-mCi) source of 63Ni if the
4–32

radius of the outer surface of the cylindrical detector is about 2 cm and the inverse square law

is applicable to the radiation field from the detector.

4.3.4.1 Distribution and Transport

As discussed in Section 4.3.4, the dose rate from63Ni while within its detector and instrument

housing is essentially zero (0); there is no distribution and transport dose. For 3H sources in

ECDs, the hypothetical radiation doses from distribution and transport are estimated in this

section using the generic methodology of Appendix A.3. In applying this methodology, it is

assumed that local parcel-delivery drivers in large trucks pick up the ECDs from manufacturers

or suppliers and take them to local terminals, where they are shipped by semi-truck to other

local terminals for delivery to customers. It is also assumed that each shipment passes through

an average of four regional terminals before reaching its final destination. A typical shipment

from a single manufacturer or supplier is assumed to consist of one ECD containing 11 GBq

(300 mCi) of 3

H.

It is assumed further that (1) the radiation doses to workers at both local terminals and regional

terminals are the same as those estimated for workers in a large warehouse, (2) a local

parcel-delivery driver could annually pick up as many as 100 ECDs containing 11 GBq

(300 mCi) of 3

H per device from a single manufacturer or supplier, and (3) the leakage rate

from the ECDs containing 3

H is 0.7 ppm/h or slightly less than the value of 1 ppm/h used in the

development of the generic methodology in Appendix A.3. A 3

H leakage rate of 0.7 ppm/h is

based on the estimated leakage rate of 0.19 MBq (5 Ci) per day at room temperatures from an

ECD containing 11 GBq (300 mCi) of 3

H (see Section 4.3.4).

Based on these assumptions and the generic methodology in Appendix A.3, the annual

individual EDE to the local parcel-delivery driver for ECDs containing 3

H could be 8×104 mSv

(0.08 mrem). Individual doses would be less to other truck drivers, workers in the truck

terminals, and members of the public along truck routes. The collective EDEs from 1 year’s

distribution of ECDs could be 1×105

 person-Sv (1×103 person-rem).

4.3.4.2 Routine Use

Three different exposure pathways during routine use are considered in this section. The first

pathway is exposure during sample analysis. The second is exposure during transport of the

detectors for field use. The third is exposure during replacement of the 3

H sources in these

devices. The resulting exposures from devices distributed under an exemption are assumed to

be the same as those from devices being distributed under the current regulatory scheme, but

there could be differences depending on particular design and other requirements that might be

imposed on the device manufacturers or distributors.

 4.3.4.2.1 Sample Analysis

As discussed in Section 4.3.4, the dose rate from 63Ni while within its detector and instrument

housing is essentially zero (0); there is no dose from sample analysis. To estimate the potential

radiation doses during sample analysis, it is assumed that one-half of the 3

H sources are used

with fixed gas chromatographs and the other one-half of the 3

H sources are used with portable

gas chromatographs. In the case of fixed gas chromatographs, the assumption is that the

operator is exposed to three ECDs as follows: (1) the operator is located at an average distance
4–33

of 1 meter from one ECD and an average distance of 2 meters from the other two ECDs, and

(2) the ECDs are located in a laboratory with a ventilation rate of 6 volume changes per hour

(see Appendix A.1). In the case of portable gas chromatographs, it was assumed that the

operator was exposed to the ECD at an average distance of 30 cm for 500 h/yr during sample

analysis, and a typical ventilation rate for varying field type conditions is 1 volume change per

hour.

It was assumed that the users of the ECDs inhale 3

H as it escapes into a hemispherical air

space about the ECD and the escaping 3

H is all converted to its oxide form, HTO (i.e., 3H water

vapor). For a user of a portable instrument or a user of a fixed instrument located at a distance

of 1 meter, the escape of 3

H into a hemispherical air space with a radius of 1.5 meters is

considered, and for a fixed instrument located at a distance of 2 meters, the escape of 3

H into a

hemispherical air space with a radius of 2.5 meters is considered. Also, during sample

analysis, a 3

H leakage rate of 7 ppm/h or 1.9 MBq/day (50 Ci/day) from an 11-GBq (300-mCi)

source of 3

H at 250C (see Section 4.3.4) and a breathing rate of 1.2 m3/h for a user engaged in

light physical activity are assumed. For a portable instrument, the annual EDE to the user could

be 0.17 mSv (17 mrem) if the user operates the instrument for a total of 500 h/yr during sample

analysis. For a user of three fixed instruments, the calculated annual EDE is 0.08 mSv

(8 mrem), if the user operates the three instruments for a total of 1000 h/yr during sample

analysis. 

If the ECDs on fixed gas chromatographs are vented to fume hoods or to the outside as

recommended by manufacturers, and a protection factor of at least 1000 is assumed, the

annual EDE to the user is 8×105

 mSv (0.008 mrem). Based on this latter dose estimate for

users of fixed instruments, the estimated collective EDE is 0.017 person-Sv (1.7 person-rem) to

users of both fixed and portable instruments over the first year of operation of the ECDs. The

total collective EDE is about 0.074 person-Sv (7.4 person-rem) over an assumed effective

lifetime of 5 years for ECDs containing 3

H sources and an annual distribution of 200 ECDs

containing 11 GBq (300 mCi) of 3

H each.

4.3.4.2.2 Transport

To estimate individual and collective doses from site-to-site transport of the portable gas

chromatographs during field usage, the generic methodology developed in Appendix A.3 for

express delivery via small commercial trucks was used. The driver of the small commercial

truck and members of the public along the truck route are assumed to be exposed for about

2 hours per day (see Table A.3.5), and the estimates of the individual EDE to members of the

public along the truck routes are less than those for the drivers of the small truck based on

distances and times of exposure (see Table A.3.5). As discussed in Section 4.3.4.1, for 63Ni

within the ECD and instrument housing, the external dose rate is essentially zero (0).

In the case of 3

H, the maximum individual dose to the driver (or user in this application) is

estimated to be 5.7×1010 Sv/GBq (2.1×1012 rem/Ci) of 3

H if the leakage rate is 1 ppm/h (see

Appendix A.3 and Table A.3.1). For an 11-GBq (300-mCi) source with a leakage rate of

0.19 MBq/day (5 Ci/day) or 0.7 ppm/h at room temperature (see Section 4.3.4.1), the dose for

a single transport (once per day) could be (5.7×1010 Sv/GBq (2.1×1012 rem/Ci))×(11 GBq

(3×105 Ci))×(0.7 ppm/h ÷ 1 ppm/h)×(1000 mSv/Sv (1000 mrem/rem), or 4.4×106

 mSv

(4.4×104

 mrem). Assuming that the individual transports the source 250 day/yr, the annual

dose could be 0.001 mSv (0.1 mrem).
4–34

Applying the collective dose factor from Table A.3.3, the corresponding collective dose for a

single source during the first year could be 1.1×106 person-Sv (1.1×104 person-rem). If the

100 3

H sources in the portable instruments are transported over their assumed 5-year effective

lifetime, then the total collective dose to all users and members of the public could be

5×104

 person-Sv (5×102 person-rem), considering reduction of 3H in the ECDs by decay.

 4.3.4.2.3 Source Replacement

Both 63Ni and 3

H sources in the ECDs may need to be changed because radioactive decay and

leakage may reduce the activity of the 3

H sources to levels that are not adequate for their

intended use within 5 to 10 years (see Section 4.3.4) and both types of sources may be

damaged by the corrosive action of the molecular material being analyzed (Howley et al., 1970). 

Currently, the general licensee may replace the source, using instructions provided by the

manufacturer, or have it replaced by a specific licensee, such as the manufacturer. Since the

exemption may allow the user to replace sources in all devices, the potential individual and

collective doses from this activity are considered. The 63Ni and 3

H sources emit only beta

particles (see Section 2.1), and the beta particles from the 63Ni and 3

H sources are not

energetic enough to contribute to the shallow-dose equivalent (or dose equivalent at a tissue

depth of 0.007 cm in skin). Thus, the only significant modes of exposures are those due to the

low-energy bremsstrahlung from the 63Ni sources and the leakage of 3

H from the 3H sources

(see Section 4.3.4).

In the case of 63Ni, it is assumed that the user’s hands are exposed at an average distance of

5 cm and the user’s body is exposed at an average distance of 30 cm from the unshielded

source. Based on calculations using CONDOS (Computer Codes, O’Donnell et al., 1975), and

correcting for the low energy bremsstrahlung (see Appendix A.4), the EDE rate at 30 cm from

the source is estimated to be 1×105

 mSv/h (0.001 mrem/h), and the dose equivalent rate to the

skin of the hand at 5 cm from the source is estimated to be 0.005 mSv/h (0.5 mrem/h). If the

user’s hands and body are exposed to the unshielded source for 5 minutes during a source

replacement, the dose equivalent to the user’s hands is estimated to be (0.005 mSv/h) ×

(5 min) ÷ (60 min/h) or less than 1×105

 mSv (<0.001 mrem), and the EDE from irradiation of

the user's body is estimated to be 1×105

 mSv × (5 min) ÷ (60 min/h), also less than

1×105

 mSv (<0.001 mrem). The replacement of 63Ni sources in ECDs should be rare events,

and the collective dose from this activity should be essentially zero (0).

In the case of 3

H, it is assumed that the user wears gloves and does not touch the source

directly so that 3

H is not absorbed though the skin of the hands, and the exposure is due

entirely to inhalation of 3

H leaking from the new and old sources into a hemispherical air space

with a radius of 1.5 meters. It is also assumed that: (1) the user is exposed to 3

H leakage from

both the new and old sources for 30 minutes during source replacement, (2) the leakage from

the old source is one-half of that from a new 11-GBq (300-mCi) source of 3

H at room

temperature, (3) the combined leakage from both the old and new 3

H sources is 0.28 MBq/day

(7.5 Ci/day), (4) the 3

H leaking from the two sources is all converted to 3H water vapor, (5) the

ventilation rate is 1 volume change per hour, and (6) the breathing rate is 1.2 m3

/h for a user

engaged in light activity. Based on these assumptions, the individual EDE from source

replacement could be 3×105

 mSv (0.003 mrem), and the collective EDE could be

5×106

 person-Sv (5×104 person-rem) from source replacement of 1 year’s distribution of

200 sources containing 11 GBq (300 mCi) of 3

H each.
4–35

4.3.4.3 Disposal

Although the 3

H and 63Ni sources in the ECDs are now being discarded as radioactive waste by

the manufacturer under the current regulatory scheme, for purposes of evaluating a possible

exemption, all sources distributed are assumed to be disposed as ordinary waste, as there are

usually no controls over disposal under an exemption. To estimate the potential doses from the

disposal of the 3

H and 63Ni sources as ordinary waste at landfills and incinerators, the generic

disposal methodology of Appendix A.2 is used. It is assumed that all potential pathways of

exposure are fully operative during disposal at both landfills and incinerators.

In applying the methodology of Appendix A.2, it is assumed that: (1) each 3

H source is

discarded at the end of 5 years and contains only half of the initial activity of 11 GBq (300 mCi)

because of the radioactive decay of 3

H and its leakage from the sources, and (2) each 63Ni

source is discarded at the end of 10 years and contains essentially the full initial activity of

0.74 GBq (20 mCi) because of the nonleakage of 63Ni from the sources and its long half-life of

100.1 years. Thus, the total activity in the 200 3

H sources at the time of disposal is 1.1 TBq

(30 Ci) and the total activity in the 2000 63Ni sources at the time of disposal is 1.5 TBq (40 Ci). 

It is assumed that 80% of the sources are landfill disposal and 20% are incinerated. Since the

number of sources disposed of annually is less than 3500, the assumed number of landfills, the

applicable dose-to-source ratios (DSRs) for individual dose in Appendix A.2 are multiplied by

the ratio of 3500 to the number of items annually disposed. A similar correction is made for the

incineration of the 3

H sources.

The estimates of potential individual and collective doses from landfill disposal of 3

H sources

are summarized as follows:

 The individual EDE to collectors at landfills could be 2×105

 mSv/yr (0.002 mrem/yr). 

For workers at landfills, off-site members of the public near landfills, and future on-site

residents, the individual doses would be less.

 The total collective EDE could be 3×104

 person-Sv (3×102 person-rem), due almost

entirely to exposure to off-site members of the public from groundwater releases.

The estimates of potential individual and collective doses from incineration of 20% of the 3

H

sources are summarized as follows:

 The individual EDE to collectors at incinerators could be 2×105

 mSv/yr (0.002 mrem/yr). 

For workers at incinerators and off-site members of the public near incinerators, the

individual doses would be less.

 The total collective EDE could be 3×105

 person-Sv (3×103 person-rem), due almost

entirely to exposure to off-site members of the public from airborne releases during

incinerator operations.

The estimates of individual and collective doses from landfill disposal of 80% of the 63Ni sources

are summarized as follows:
4–36

 The individual EDE to collectors at landfills could be 2×105

 mSv/yr (0.002 mrem/yr). 

For workers at landfills, off-site members of the public near landfills, and future on-site

residents, the individual doses would be less.

 The total collective EDE could be 1×104

 person-Sv (0.01 person-rem), due almost

entirely to exposure to collectors and workers at landfills.

The estimates of potential individual and collective doses from incineration of 20% 63Ni sources

are summarized as follows:

 The individual EDE to collectors at incinerators could be 6×105

 mSv/yr (0.006 mrem/yr). 

For workers at incinerators and off-site members of the public near incinerators, the

individual doses would be less.

 The total collective EDE could be 1×105 person-Sv (1×103 person-rem), due almost

entirely to exposure to collectors at incinerators.

4.3.4.4 Accidents and Misuse

Three different pathways for exposure during accidents and misuse of gas chromatographs

containing ECDs are considered in this section. The first pathway involves radioactive material

released from an ECD source during an accident involving a fire. The second pathway involves

radioactive material leaking from a damaged ECD source. The third pathway involves

vandalism of a portable gas chromatograph containing an ECD.

NRC records (NRC, Databases, NMED, Reports) indicate about 80 instances of leaking 63Ni

sources in ECDs for gas chromatographs between 1990 and 1996. Only leakage from 63Ni

sources is normally reported because leakage from 3

H sources is expected during routine

usage (see Section 4.3.4) and leak testing is not required for these sources (see Section 4.3.1). 

An exemption also would not include a leak testing requirement for 63Ni sources in ECDs and

without it, leakage from a damaged 63Ni source may go unnoticed. In addition, there are other

circumstances under 10 CFR 31.5(c)(5) and (10) for which actions are required to control

exposures to users and members of the public that would not be applicable under this potential

exemption. For example, if the labeling requirements of the exemption are different than under

the current regulatory scheme, it could increase the possibility of misuse of the 63Ni and 3

H

sources from portable gas chromatographs containing ECDs.

In the case of an accident involving fire, the following scenarios are considered: (1) a user takes

a portable gas chromatograph home at night and the ECD in the gas chromatograph is involved

in a residential fire, and (2) a transportation accident occurs involving a typical shipment of

either four ECDs containing 0.74 GBq (20 mCi) of 63Ni each or one ECD containing 11 GBq

(300 mCi) of 3

H. A release fraction of 0.1% is assumed for a 63Ni source and a release fraction

of 100% is assumed for a 3

H source. Based on these assumptions and the generic accident

methodology in Appendix A.1, the estimates of individual dose from inhalation of radioactive

materials released from the sources are summarized as follows:

 For a person trying to escape from a residential fire or a neighbor trying to rescue a

person from a residential fire, the individual dose could be 3×104 mSv (0.03 mrem) from
4–37

a 0.74-GBq (20-mCi) source of 63Ni and 0.1mSv (10 mrem) from an 11-GBq (300-mCi)

source of 3

H.

 For a firefighter who wears respiratory equipment and protective clothing at a residential

fire, the individual dose would be less than 1×105

 mSv (<0.001 mrem) from a 0.74-GBq

(20-mCi) source of 63Ni and 3×104

 mSv (0.03 mrem) from an 11-GBq (300-mCi) source

of 3

H. For a worker who is involved in cleanup following the fire and who does not wear

respiratory equipment, the individual doses from the 3

H and 63Ni are estimated to be less

than 1×105

 mSv (<0.001 mrem).

 For a firefighter who wears respiratory equipment and protective clothing at a

transportation fire, the individual dose would be less than 1×105

 mSv (<0.001 mrem)

from the four 0.74-GBq (20-mCi) sources of 63Ni and 3×104

 mSv (0.03 mrem) from the

single 11-GBq (300-mCi) source of 3

H. For a worker who is involved in cleanup

following the fire and who does not wear respiratory equipment, the individual doses

from the 3

H and 63Ni are estimated to be essentially zero (0) and 2×105 mSv

(0.002 mrem), respectively.

In the case of source leakage, the potential radiation dose to a user of the source and to a

waste collector are considered. To estimate the radiation dose to a user of a source, it is

assumed that a damaged 3

H source in the ECD of a portable gas chromatograph leaks at 10

times the usual rate during sample analysis (see Section 4.3.4.2.1). Thus, the individual EDE

would be 2 mSv/yr (200 mrem/yr) and the leakage would amount to about 3% of the 3

H from an

11-GBq (300-mCi) source over the assumed sample-analysis period of 500 hours. If 3% of the

material leaks from a damaged 63Ni source during sample analysis, then the EDE to a user of a

portable instrument could be about 3 mSv/yr (300 mrem/yr). However, this is an extremely

conservative estimate for an individual dose from a damaged 63Ni source. The best estimate of

the potential individual doses to a user of a portable gas chromatograph and ECD with a

damaged 63Ni or 3

H source would be about 1 mSv/yr (100 mrem/yr). The potential individual

dose to a user of a fixed gas chromatograph and ECD with a damaged 63Ni or 3

H source would

be less. To estimate the radiation dose to a waste collector, the generic disposal methodology

in Appendix A.2 (see Table A.2.1) was used. Because the dose-to-source ratios are divided by

the number of landfills in the United States, the first thing that must be done is to multiply by

3500 (i.e., the estimated number of U.S. landfills) and then by the amount of activity in the 63Ni

or 3

H sources (i.e., 0.74 GBq (20 mCi) and 11 GBq (300 mCi), respectively). Thus, the

individual dose to the waste collector could be 2×105

 mSv(0.002 mrem) due to leakage from a

damaged 63Ni source and 3×105

 mSv (0.003 mrem) due to leakage from a damaged 3H

source.

In the case of misuse, the exposure to a person who finds and vandalizes a portable gas

chromatograph with an ECD containing either a 0.74-GBq (20-mCi) source of 63Ni or 11-GBq

(300-mCi) source of 3

H is considered. It is assumed that the person deliberately destroys the

gas chromatograph and ECD out of curiosity and handles the 63Ni or 3

H source for 30 minutes

before discarding both the source and other detector parts. For the dose to the hands, the

estimate of the dose equivalent is essentially zero (0) for the 3

H source (see Section 4.3.4),

whereas the dose equivalent is about 0.003 mSv (0.3 mrem) for the 63Ni source based on

calculations using CONDOS (Computer Codes, O’Donnell et al., 1975) and correcting for low

energy bremsstrahlung (see Appendix A.4). For potential ingestion of material due to handling

of the 63Ni or 3

H sources, the generic accident methodology developed in Appendix A.1 was
4–38

used for spills of radioactive material in the form of a powder. It is assumed, first, that 10% of

the material on the source is deposited on the body and, second, that 0.1% of the deposited

material is ingested by mouth before it is removed from the body by washing. In the case of 3

H,

it is further assumed that the 3

H is in the form of a titanium tritide powder and it is treated as the

other powders were treated in Appendix A.1 or A.2. Thus, the estimated EDE could be

0.01 mSv (1 mrem) from the ingestion of 63Ni and 0.02 mSv (2 mrem) from ingestion of 3

H in

the form of titanium tritide powder.

4.3.5 Summary

Table 4.3.1 presents the results of the analysis of potential radiological impacts from the

potential exemption of ECDs containing not more than 0.74 GBq (20 mCi) of 63Ni or 11 GBq

(300 mCi) of 3

H. The radiation doses estimated in this assessment are based on typical

designs of devices distributed under the requirements applicable to distributors and applicants

for license to distribute such devices for use under 10 CFR 31.5. The details of the designs are

important to ensuring control of exposure. For the radiation doses resulting under an

exemption of these devices to be comparable to those estimated, similar controls over the

distributors would be necessary to ensure that the designs of the devices are comparable in

minimizing exposures to users and members of the public.

Annual EDEs to individuals are estimated to be about 8×104

 mSv (0.08 mrem) for distribution

and transport; 0.17 mSv (17 mrem) for routine use involving sample analysis, transport for field

use, and source replacement; and 6×105

 mSv (0.006 mrem) for disposal in landfills and by

incineration. For all of these activities combined, the collective EDE to all users and members

of the public is estimated to be 0.08 person-Sv (8 person-rem). This collective dose estimate

assumes an annual distribution of 2000 sources of 63Ni initially containing 0.74 GBq (20 mCi)

each and 200 sources of 3

H initially containing 11 GBq (300 mCi) each. The effective lifetime of

the 200 sources of 3

H was taken to be 5 years and the effective lifetime of the 2000 63Ni

sources was taken to be 10 years. For accidents involving fire and for misuse involving

vandalism, the maximum EDEs were estimated to be 0.1 mSv (10 mrem) and 0.02 mSv

(2 mrem) for 63Ni and 3

H, respectively.

For damaged source leakage, the individual EDEs to users could be as much as 2 mSv/yr

(200 mrem/yr). NRC records (NRC, Databases, NMED, Reports) indicate about 80 instances

of leaking 63Ni sources between 1990 and 1996. Only leakage from 63Ni sources is normally

reported because leakage from 3

H sources is expected during normal usage and leak testing is

not required for these sources.
4–39

Table 4.3.1 Potential Radiation Doses From Electron Capture Detectors 

Using 63Ni and 3

H Sources

Exposure Pathway

Individual Annual

Effective

Dose Equivalent

 (mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport 0.8c 0.001

Routine use

 Sample analysis

 Transport

 Source replacement

17d

 0.1e

 0.003f

17

 0.05

 <0.001

Disposal as ordinary waste

 Landfills

 Incinerators

 0.002g

 0.006h

 0.04

 0.004

Accidents and misuse

 Accidents involving fire

 Vandalism of instrument

 Source leakage

 10i

 2j

 200k

See following page for footnotes.
4–40

Footnotes to Table 4.3.1

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Collective doses for 1 year’s distribution. Refer to text for time period of collective dose

assessment. 

c

 Dose estimate applies to local parcel-delivery drivers; dose estimates are less for long-haul

semi-truck drivers, terminal workers, and members of the public along truck routes (see

Section 4.3.4.1).

d

 Dose estimate applies to a user of portable gas chromatographs containing ECDs; dose

estimates are less for users of fixed ECDs and gas chromatographs (see Section 4.3.4.2.1).

e

 Dose estimate applies to users of portable gas chromatographs containing ECDs; dose

estimates are less for members of the public during site-to-site transport for field usage (see

Section 4.3.4.2.2).

f

 Dose estimate applies to internal exposure to user from 3H inhalation during source

replacement; dose estimates for a 63Ni source replacement are less than 1×105

 mSv

(<0.001 mrem) for effective dose equivalent (EDE) from external irradiation of user’s body and

for dose equivalent to user's hands (see Section 4.3.4.2.3).

g

 Dose estimate applies to waste collectors at landfills; dose estimates are less for workers at

landfills, off-site members of the public near landfills, and future on-site residents (see

Section 4.3.4.3).

h

 Dose estimate applies to waste collectors at incinerators; dose estimates are less for workers

at incinerators and off-site members of the public near incinerator sites (see Section 4.3.4.3).

i

 Dose estimate applies to inhalation by a person escaping from a residential fire or local

neighbor trying to rescue a person from a residential fire; dose estimates are less to a firefighter

or worker during cleanup following the fire (see Section 4.3.4.4).

j

 Dose estimate applies to internal exposure to a person who finds and destroys a portable gas

chromatograph with an ECD containing 3

H: dose estimates for a 63Ni source are 0.003 mSv

(0.3 mrem) for the dose equivalent to the hands and 0.01 mSv (1 mrem) for the EDE from

internal exposure due to ingestion (see Section 4.3.4.4).

k

 Dose estimate applies to internal exposure to a user of a portable gas chromatograph with an

ECD containing a leaking source of either 63Ni or 3

H (see Section 4.3.4.4).
4–41

4.4 X-ray Fluorescence Analyzers

4.4.1 Introduction

A general license is granted in 10 CFR 31.5 to acquire, possess, use, or transfer byproduct

material contained in devices designed and manufactured for measuring chemical composition,

either qualitatively or quantitatively. Included in the general license are requirements for

labeling, leak testing, and proper storage or disposition of the device. The licensee is also

subject to terms and conditions set forth in 10 CFR 31.2 dealing with general license

requirements, transfer of byproduct material, reporting and recordkeeping, and inspection. 

Leak testing is required if the byproduct material source contains more than

0.37 megabecquerel (MBq) (10 microcurie (Ci)) of a material emitting alpha particles or

3.7 MBq (100 Ci) of a material emitting only beta particles and photons.

X-ray fluorescence analyzers are designed for use in nondestructive analysis to determine the

elemental chemical composition of solid and liquid samples. A variety of radionuclides,

including byproduct materials, can be used as the source of X-rays. There are no limits in

10 CFR 31.5 on the amount of byproduct material that can be used in X-ray fluorescence

analyzers. However, an applicant for a specific license to manufacture or initially transfer X-ray

fluorescence analyzers for use under 10 CFR 31.5 must demonstrate that the devices will meet

certain requirements contained in 10 CFR 32.51. These requirements are described below:

 The device can be safely operated by persons without training in radiological protection.

 Under ordinary conditions of handling, storage, and use of the device, the byproduct

material contained in the device will not be released or inadvertently removed from the

device, and it is unlikely that any person will receive in any 1-year period a total effective

dose equivalent (EDE) in excess of 5 millisieverts (mSv) (500 mrem), or the sum of the

deep-dose equivalent and committed dose equivalent to any individual organ or tissue

other than the lens of the eye in excess of 50 mSv (5 rem); an eye dose equivalent in

excess of 15 mSv (1.5 rem); or a shallow-dose equivalent in excess of 50 mSv (5 rem)

to the skin or to any extremity (i.e., hand, elbow, arm below the elbow, foot, knee, or leg

below the knee).

 Under accident conditions (such as fire and explosion) associated with handling,

storage, and use of the device, it is unlikely that any person would receive an external

dose equivalent or committed internal dose equivalent in excess of 0.15 Sv (15 rem) to

the whole body, head and trunk, active blood-forming organs, gonads or lens of the eye;

2 Sv (200 rem) to the hands and forearms, feet and ankles, or localized areas of skin

averaged over areas no larger than 1 cm2

; and 0.50 Sv (50 rem) to any other organs.

Some X-ray fluorescence analyzers are potential candidates for exemption from the general

licensing requirements of 10 CFR 31.5. This assessment evaluates the potential radiation

doses that could result if the receipt, possession, use, and transfer of certain of these products

were exempt from licensing. The assumed conditions for this possible exemption are 1.5 GBq

(40 mCi) of 55Fe or 1.9 GBq (50 mCi) of 109Cd per device.
4–42

4.4.2 Description of Items Considered for Exempt Distribution

An early, useful review of X-ray fluorescence techniques and applications is provided by Russ

(1971), and a more recent, elementary description is given in a self-study text by Whiston

(1987). Since the recognition and understanding of atomic emission spectra in the early days

of quantum physics, X-ray fluorescence has been an important tool for elemental chemical

analysis. When an atom is bombarded with beta particles or X-rays, an inner orbital electron

may be displaced, leaving the atom in an excited state. The atom can regain stability by

rearrangement of its electrons. Inner shell vacancies may be filled with electrons from outer

shells, leading to the emission of characteristic X-ray energies that can be used to determine

elemental chemical compositions of the materials being tested.

An analyzer consists of a radiation source to irradiate the sample and either a solid-state

detector or gas proportional counter to detect the X-ray fluorescence from the sample. The

output signal from the solid-state detector or gas proportional counter is fed into a multichannel

pulse-height analyzer for separation by energy, coupled with a microcomputer for processing

the data. The geometrical arrangement of the components of an X-ray fluorescence analyzer

can vary widely, depending on the intended uses of the instrument. The devices can range in

size from small handheld, dedicated instruments (e.g., to measure lead content on surface

materials) to large multisource, multisample, multipurpose laboratory instruments.

The introduction and increasingly widespread use of high-efficiency, high-resolution solid-state

detectors, initiated more than 30 years ago, led to devices both smaller in size and requiring

lower source strengths for producing the requisite inner shell electron vacancies. In particular,

sealed radionuclide sources can be used in place of X-ray tubes. Radionuclides commonly

used in X-ray fluorescence analyzers have included the low-energy photon emitters 55Fe, 57Co,

109Cd, 153Gd, 238Pu, 241Am, or 244Cm, and the beta-particle emitters 3

H or 147Pm combined with a

stopping material to provide a broad band of bremsstrahlung (Tertian and Claisse, 1982). The

radionuclides 55Fe and 109Cd in equilibrium with its short-lived decay product, 109mAg, emit very

low-energy Auger electrons and photons. A photon with an energy of only 6 keV is emitted in

the radioactive decay of 55Fe with an intensity of approximately 28%. Photons with energies of

about 21 keV and 88 keV are emitted in the decays of 109Cd/109mAg with intensities of

approximately 110% and 4%, respectively (International Commission on Radiological Protection

(ICRP) 38). The radioactive half-lives are 2.7 years for 55Fe, 464 days (1.27 years) for 109Cd,

and 39.6 seconds for 109mAg (see Section 2.1).

Two different types of portable X-ray fluorescence analyzers may be distributed under this

possible exemption. With one type of analyzer, a sample of a liquid or solid is inserted into the

instrument for analysis. With the other type, the instrument is placed on the surface of the

object or material to be analyzed. The 55Fe and 109Cd sources used in these instruments are

electroplated on a metallic disk, which is heat treated to firmly affix the radioactive materials to

the disk and covered with a thin metallic window to form a sealed source. The sealed source is

then mounted inside a source housing with an aperture to emit low-energy photons in the

direction of the sample or material to be analyzed. The aperture is covered by a shutter to

provide additional shielding of the source when not in use. The sealed sources are well

shielded in directions other than the sample direction to ensure a low background for the

detector, and access to the sealed sources is normally restricted by the use of tamper-proof
4–43

screws on the source housing. Automatic shutters, interlocks, and other safety features are

also included as a part of the instruments to minimize operator exposure.

4.4.3 Summary of Previous Analyses and Assessments

Each applicant for a specific license to manufacture or initially transfer X-ray fluorescence

analyzers for use under 10 CFR 31.5 is required to submit information to the Nuclear

Regulatory Commission (NRC) to show that its product meets the dose criteria summarized in

Section 4.4.1.

4.4.4 Present Exemption Analysis

NRC data (NRC, Memoranda, Bernero, 1994; NRC, E-mail Message) indicate that

approximately 700 sources initially containing 1.5 GBq (40 mCi) each of 55Fe and 1600 sources

initially containing 1.9 GBq (50 mCi) each of 109Cd could be distributed annually for use in X-ray

fluorescence analyzers under the general license in 10 CFR 31.5. This corresponds to an

annual distribution of about 1 TBq (28 Ci) of 55Fe and 3 TBq (80 Ci) of 109Cd. For purposes of

this analysis, it is assumed that the same number of such devices could be distributed under an

exemption. However, the removal of some requirements of a general license could ultimately

increase the number of such devices distributed annually. The effective lifetimes of the 55Fe

and 109Cd sources are assumed to be 3 to 4 years.

4.4.4.1 Distribution and Transport

During distribution and transport, the radiation doses to individuals from external exposure to

55Fe sources are essentially zero (0), due to the very low energy of photons emitted by 55Fe. 

Hence, this section estimates only the radiation doses to individuals from external exposure to

109Cd sources using the generic distribution methodology of Appendix A.3.

In applying this methodology, it is assumed that the 109Cd sources are shipped primarily by a

parcel-delivery service, and that a driver in a large van picks up the sources and takes them to

a local terminal for shipment to other local terminals for delivery to customers. A typical

shipment from a manufacturer or supplier is assumed to consist of a single source containing

1.9 GBq (50 mCi) of 109Cd. It is further assumed that semi-trucks are used to transport the

sources between local terminals, and that the sources pass through an average of four regional

terminals before reaching their final destination. Radiation exposures to workers at both local

and regional terminals are assumed to be similar to those to workers at a large warehouse (see

Appendix A.3).

Based on the above assumptions and the generic methodology of Appendix A.3, the individual

receiving the largest dose is the local driver who is assumed to pick up an average of

100 sources containing 1.9 GBq (50 mCi) each of 109Cd during the year from the same

manufacturer or supplier. The annual EDE to this individual is estimated to be 0.05 mSv

(5 mrem). Individual doses to long-haul semi-truck drivers, terminal workers, and members of

the public along the truck routes will be less. The annual collective dose to all truck drivers,

terminal workers, and members of the public along the truck routes is estimated to be

6×103

 person-Sv (0.6 person-rem).
4–44

4.4.4.2 Routine Use

Three different exposure pathways during routine use are considered in this section. The first

pathway is exposure during sample analysis. The second is exposure during transport of the

X-ray fluorescence analyzers for field use. The third is exposure during replacement of sources

in these devices. The resulting exposures from devices distributed under an exemption are

assumed to be the same as those from devices distributed under the present regulatory

scheme, but there could be differences depending on particular design and other requirements

that might be imposed on the manufacturers or distributors of the devices.

4.4.4.2.1 Sample Analysis

During sample analysis, radiation doses to users of portable instruments with shielded 55Fe

sources are essentially zero (0), due to the very low energy of the photons emitted by 55Fe. 

Thus, the following discussions are relevant only for portable instruments with shielded 109Cd

sources.

To estimate the potential radiation doses to users of portable instruments with shielded 109Cd

sources, NRC records of dose measurements reported by manufacturers were surveyed. For

portable instruments into which the samples are inserted for analyses, the dose rates at 5 and

30 cm from a 1.9-GBq (50-mCi) source of 109Cd can be reduced by shielding to about

0.002 mSv/h (0.2 mrem/h) and 5×104

 mSv Sv/h (0.05 mrem/h), respectively, with the shutter

either open or closed. For portable instruments that are placed on the surface of an object or

material to be analyzed, the dose rates at 5 and 30 cm are essentially the same with the shutter

closed, but the dose rates with the shutter open can be much greater due to backscattering of

the photons from the surface of the object or material to be analyzed. However, the

backscattered photons have very low energies and can only deliver a significant radiation dose

to hands, as discussed later in this section.

To estimate the EDE to a user during sample analysis, it is assumed that a portable instrument

with a shielded 1.9-GBq (50-mCi) source of 109Cd was located at a distance of 30 cm from the

user’s body for about 500h/yr during sample analysis. Thus, the EDE to the individual is

estimated to be (5×104

 mSv/h (0.05 mrem/h)) × (500 h/yr), or 0.25 mSv/yr (25 mrem/yr) if

decay of the 109Cd source is not considered and 0.2 mSv/yr (20 mrem/yr) if decay of the source

is considered over the first year of usage. For the collective EDE to all users, the estimated

value was approximately 0.6 person-Sv (60 person-rem) based on an EDE to a user of 0.4 mSv

(40 mrem) over an effective lifetime of 3 years for a 109Cd source and an annual distribution of

1600 sources initially containing 1.9 GBq (50 mCi) each of 109Cd.

To estimate the dose equivalent to the hands of a user during sample analysis, the data used

were obtained from NRC records (NRC, Databases, SSDR) for a small X-ray fluorescence

analyzer designed to test for lead in painted surfaces. The device is held in the palm of the

hand with the fingers in close proximity to the backscattered photons from a painted test

surface. For a shielded 109Cd source loading of 370 MBq (10 mCi) and thick wood substrate to

simulate the test surface, measurements indicated that the dose equivalent rate to the fingers

could be 0.01 mSv/h (1 mrem/h) with the shutter open and 3×104

 Sv/h (0.03 mrem/h) with the

shutter closed. If the instrument is held a total of 500 h/yr during sample analysis with the

shutter open 10% of the time, then the radiation dose to the fingers is estimated to be

(0.01 mSv/h (1 mrem/h)) × (50 h/yr) + (0.0003 mSv/h (0.03 mrem/h)) × (450 h/yr), or 0.6 mSv/yr
4–45

(60 mrem/yr) if decay of the 109Cd source is not considered and 0.5 mSv (50 mrem) if decay of

the source is considered over the first year of usage. Thus, the radiation dose to the fingers

could be as much as 2.5 mSv/yr (250 mrem/yr) if a source loading of 1.9 GBq (50 mCi) of 109Cd

was used in this type of portable instrument.

4.4.4.2.2 Transport

During transport for field use, radiation doses to users of portable instruments with shielded

55Fe sources are essentially zero (0), due to the very low energy of the photons emitted by 55Fe. 

Thus, the following discussions are relevant only for portable instruments with shielded 109Cd

sources.

To estimate individual and collective doses during transport of the portable instruments for field

use, the generic methodology developed in Appendix A.3 for express delivery via small

commercial trucks was used. During express delivery with a small commercial truck, the driver

and members of the public along the truck route are assumed to be exposed about 2 h/day

(see Table A.3.5). The maximum individual dose to the driver (or user in this application) is

estimated to be 3.8×107

 Sv/GBq (1.4×109 rem/Ci) of 109Cd (see Table A.3.1). Thus, the dose

rate to the user is 7.0×104

 mSv/day (7.0×102 mrem/day) and 3.5×104 mSv/h

(3.5×102

 mrem/h) for a portable instrument with a 1.9-GBq (50-mCi) source of 109Cd. If the

user transports the portable instrument 1 h/day for 250 day/yr (i.e., 250 h/yr), then the

maximum individual dose to a user could be (3.5×104

 mSv/h (0.035 mrem/h)) × (250 h/yr), or

about 0.09 mSv/yr (9 mrem/yr) if the decay of the 109Cd source is not considered or 0.07 mSv

(7 mrem) per year if the decay of the 109Cd source is considered over the first year of usage.

The collective dose to a user and members of the public during transport for field use is

estimated to be 5.4×108

 person-Sv/GBq (2.0×1010 person-rem/Ci) of 109Cd (see Table A.3.3),

or 1.0×107

 person-Sv/day (1.0×105 person-rem/day) and 5.0×108 person-Sv/h

(5.0×106

 person-rem/h) from a portable instrument with a 1.9-GBq (50-mCi) source of 109Cd. If

a portable instrument is transported 1 h/day for 250 day/yr (i.e., 250 h/yr) over the assumed

3-year effective life of a 109Cd source, then the collective dose to all users and members of the

public could be (5.0×108

 person-Sv/h (5.0×106 person-rem/h)) × (250 h/yr) × (3 y/source) ×

(1600 sources), or 6×102

 person-Sv (6 person-rem) if the decay of the 109Cd sources is not

considered or 3×102

 person-Sv (3 person-rem) if the decay is considered over the assumed

3-year effective life of the 109Cd sources.

4.4.4.2.3 Source Replacement

Because the sources in these devices decay to levels that are not adequate for their intended

use within 3 to 4 years, the sources are usually replaced a number of times over the lifetime of

a device. Currently the general licensee may replace the source after its useful life, using

instructions provided by the manufacturer, or have it replaced by a specific licensee, such as

the manufacturer. Since the exemption may allow the user to replace sources in all devices,

the potential individual and collective doses from this activity during routine use are considered.

During source replacement, it is assumed that the user’s hands are exposed at an average

distance of 5 cm and the user’s body is exposed at an average distance of 30 cm from the

unshielded 55Fe and 109Cd sources. The radiation dose to the hands and skin of the whole

body, as approximated by the so-called shallow depth dose at a tissue depth of 0.007 cm, and
4–46

the EDE from irradiation of the whole body by an unshielded point source of 55Fe or 109Cd in air,

are as follows:

 The shallow depth dose rates are approximately 80 and 1 mSv/h (8000 and

100 mrem/h) at distances of 5 cm and 30 cm, respectively, from an unshielded 1.5-GBq

(40-mCi) point source of 55Fe in air.

 The shallow depth dose rates are approximately 30 and 0.82 mSv/h (3000 and

82 mrem/h) at distances of 5 cm and 30 cm, respectively, from an unshielded 1.9-GBq

(50-mCi) point source of 109Cd in air.

 The EDE rates are essentially zero (0) and approximately 0.17 mSv/h (17 mrem/h) at

30 cm from an unshielded 1.5-GBq (40-mCi) point source of 55Fe and an unshielded

1.9-GBq (50-mCi) point source of 109Cd in air, respectively.

The above dose rates for an unshielded 109Cd point source in air were calculated by using

MicroShield (Computer Codes, Grove Engineering, 1996). It was not possible, however, to use

MicroShield to calculate dose rates for 55Fe due to the extremely low energy of the 6-keV

photons. Thus, the above dose rates for an unshielded 55Fe point source in air were calculated

using the photon mass attenuation and energy-absorption coefficients of Hubbell (1982) and a

calculated energy fluence, where dose rate equals fluence times the mass energy-absorption

coefficient.

To estimate the dose equivalent to the hands of a user from source replacement, it is assumed

that the user’s hands are exposed at 5 cm from the unshielded sources for 5 minutes during a

source change (2.5 minutes during removal of the old source and 2.5 minutes during the

installation of the new source). The initial dose rates for the new sources are given above. The

dose rates from the old source after an assumed decay period of 3 years are 0.46 and 0.20

times the initial dose rates of the 55Fe and 109Cd sources, respectively. Thus, the skin dose to a

user’s hands is estimated to be [(80 mSv/h (8000 mrem/h)) × (1 + 0.46) × (2.5 min) ÷

(60 min/h)], or about 5 mSv (500 mrem) for 55Fe source replacement, and [(30 mSv/h

(3000 mrem/h)) × (1 + 0.20) × (2.5 min) ÷ (60 min/h)], or about 2 mSv (200 mrem) for 109Cd

source replacement.

To estimate the EDE to a user from source replacement, it is assumed that the user’s body is

exposed at 30 cm from the unshielded sources for 5 minutes during a source change. The

estimated total EDE is that due to the whole body plus 1% of the skin dose to the whole body,

which does not include the so-called extremities (i.e., the hands and arms below the elbow or

the feet and legs below the knees). Thus, the total EDE to individual users is estimated to be

[(0.17 mSv/h (17 mrem/h)) × (1 + 0.20) × (2.5 min) ÷ (60 min/h)] + [(0.01) × (0.82 mSv/h) × (1 +

0.20) × (2.5 min) ÷ (60 min/h)], or about 0.009 mSv (0.9 mrem) for a 109Cd source replacement,

and [(0.01) × (1 mSv/h) × (1 + 0.46) × (2.5 min) ÷ (60 min/h)], or about 6×104

 mSv (0.06 mrem)

for a 55Fe source replacement. These values, especially the one for 55Fe, are very conservative

because the weighting factor of 1% for the skin dose to whole body assumes that the skin of

the whole body is uniformly irradiated, which is not the case here.

For the collective EDE to all users from source replacement, the estimated value is

0.01 person-Sv (1 person-rem) based on the above EDEs to individual users during source

replacement and the annual distribution of 700 sources (NRC, Memoranda, Bernero, 1994;
4–47

NRC, E-mail Message) initially containing 1.5 GBq (40 mCi) each of 55Fe and 1600 sources

initially containing 1.9 GBq (50 mCi) each of 109Cd.

4.4.4.3 Disposal

Although the 55Fe and 109Cd sources in many instruments are now being replaced and

discarded as radioactive waste by the manufacturer under the current regulatory scheme, for

purposes of evaluating a possible exemption, all sources distributed are assumed to be

disposed as ordinary waste. There is because there are usually no controls over disposal

under an exemption. Exactly what fraction are disposed of as ordinary waste may depend on

the specific conditions of the exemption (e.g., whether the device is designed to accommodate

source change by the user). If the conservative assumption is made that all sources are

disposed as ordinary waste at the end of 3 years (NRC, Memoranda, Bernero, 1994; NRC,

E-mail Message), the 700 sources of 55Fe would contain about 0.5 TBq (13 Ci) at the time of

disposal and the 1600 sources of 109Cd would contain about 0.6 TBq (16 Ci) at the time of

disposal.

To estimate the potential doses from disposal of the 55Fe and 109Cd sources as ordinary waste,

the generic disposal methodology in Appendix A.2 was used, with 80% being disposed of at

landfills and 20% being incinerated. During waste collection at both incinerators and landfills

and disposal at landfills, it is assumed that the sealed sources normally remain intact and that

waste collectors or workers at landfills do not touch the discarded sources with their hands. 

Based on these assumptions and the short half-lives of 55Fe and 109Cd, the only significant

exposure pathway is direct whole-body irradiation of waste collectors and landfill workers by

photons from the 109Cd sources. The radiation doses from direct whole-body exposure to the

55Fe sources are essentially zero (0), due to the very low-energy photons from 55Fe. All

exposure pathways are assumed, however, to be operative in the case of workers at

incinerators and off-site members of the public near the incinerator sites (see Appendix A.2). 

Since the number of sources disposed of annually is less than 3500, the assumed number of

landfills, the applicable dose-to-source ratios (DSRs) for individual dose in Appendix A.2 are

multiplied by the ratio of 3500 to the number of items annually disposed. A similar correction is

made for the incineration of the 55Fe sources.

The estimates of potential individual and collective doses from landfill disposal of 109Cd sources

are summarized as follows:

 For collectors at landfills, the individual EDE could be 0.002 mSv/yr (0.2 mrem/yr), and

the collective EDE could be 4×103

 person-Sv/yr (0.4 person-rem/yr).

 For workers at landfills, the individual EDE could be 1×104

 mSv/yr (0.01 mrem/yr), and

the collective EDE could be 1×103

 person-Sv/yr (0.1 person-rem/yr).

The estimates of potential individual and collective doses from incineration of 20% of the 109Cd

sources are summarized as follows:

 For collectors at incinerators, the individual EDE could be 0.006 mSv/yr (0.6 mrem/yr),

and the collective EDE could be 9×104

 person-Sv/yr (0.09 person-rem/yr).
4–48

 For workers at incinerators, the individual EDE would be less than 1×105

 mSv/yr

(<0.001 mrem/yr), and the collective EDE could be 8×107

 person-Sv/yr (8×105 personrem/yr).

 For off-site members of the public near incinerators, the individual EDE would be less

than 1×105

 mSv/yr (<0.001 mrem/yr), and the collective EDE could be

6×106

 person-Sv/yr (6×104 person-rem/yr).

The estimates of individual and collective doses from incineration of 55Fe sources are

summarized as follows:

 For workers at incinerators, the individual EDE would be less than 1×105

 mSv/yr

(<0.001 mrem/yr), and the collective EDE could be 2×109

 person-Sv/yr

(2×107

 person-rem/yr).

 For off-site members of the public near incinerators, the individual EDE would be less

than 1×105

 mSv/yr (<0.001 mrem/yr), and the collective EDE could be

2×107

 person-Sv/yr (2×105 person-rem/yr).

Because of the low-energy photons from 55Fe, the individual and collective doses to waste

collectors from 55Fe sources are essentially zero (0).

4.4.4.4 Accidents and Misuse

Three different pathways of exposure during accidents and misuse are considered in this

section. The first pathway of exposure is inhalation of radioactive material released from a

source during an accident involving fire. The second pathway of exposure is ingestion of

radioactive material leaking from a ruptured source. The third pathway of exposure is external

irradiation of the hands to a waste collector or other person who finds a discarded source and

carries it in a pocket for several hours (3 hours) before storing it in an out-of-the-way place.

NRC records indicate six instances of leaking sources in X-ray fluorescence analyzers between

1990 and 1996. An exemption would not include a leak testing requirement for these sources

(see Section 4.4.1) and without it, leakage from a failed source may go unnoticed. In addition,

there are other circumstances under 10 CFR 31.5(c)(5) and (10) for which actions are required

to control exposures to the users and members of the public that would not be applicable under

this potential exemption. For example, if the labeling requirements of the exemption are

different than under the current regulatory scheme, it could increase the possibility of misuse of

a source from a portable instrument.

In the case of an accident involving fire, the following is considered: (1) a user takes a portable

instrument home at night and it is involved in a residential fire, and (2) a single source is

involved in a transportation accident. A release fraction of 0.1% is assumed for the 55Fe or

109Cd sources used in the portable instruments. Based on these assumptions and the generic

accident methodology in Appendix A.1, the estimates of individual dose from inhalation of

radioactive materials released from the sources are summarized as follows:

 For a person trying to escape from a residential fire or a neighbor trying to rescue a

person from a residential fire, the individual dose could be 4×104 mSv (0.04 mrem)
4–49

from a 1.5-GBq (40-mCi) source of 55Fe and 0.02 mSv (2 mrem) from a 1.9-GBq

(50-mCi) source of 109Cd.

 For a firefighter who wears a respirator at a residential fire, the individual dose would be

less than 1×105

 mSv (<0.001 mrem) from a 1.5-GBq (40-mCi) source of 55Fe and

6×105

 mSv (0.006 mrem) from a 1.9-GBq (50-mCi) source of 109Cd. For a worker who

is involved in the cleanup following the fire and who does not wear a respirator, the

individual dose from these 55Fe and 109Cd sources would be less than 1×105

 mSv

(<0.001 mrem) and 3×104

 mSv (0.03 mrem), respectively.

 For a firefighter who wears a respirator at a transportation fire, the individual dose would

be less than 1×105

 mSv (<0.001 mrem) from a 1.5-GBq (40-mCi) source of 55Fe and

6×105

 mSv (0.006 mrem) from a 1.9-GBq (50-mCi) source of 109Cd. For a worker who

is involved in the cleanup following the fire and who does not wear a respirator, the

individual dose from these 55Fe and 109Cd sources could be 1×105

 mSv (0.001 mrem)

and 6×104

 mSv (0.06 mrem), respectively.

In the case of source leakage, the potential radiation doses to a user of the source and to a

waste collector are considered. To estimate the radiation dose to a user of the source, the

generic accident methodology developed in Appendix A.1 is used for the ingestion of

radioactivity following a spill of a radioactive material in the form of a powder. First, it is

assumed that 10% of the material is deposited on the skin of an individual and, second, that

0.1% of this deposited material is ingested before bathing removed the material from the body. 

Based on these assumptions and the general accident methodology of Appendix A.1 (see

Table A.1.8), the individual dose to a user of a portable instrument could be 0.03 mSv (3 mrem)

due to leakage from a 1.5-GBq (40-mCi) source of 55Fe and 0.7 mSv (70 mrem) due to leakage

from a 1.9-GBq (50-mCi) source of 109Cd.

In the case of misuse, it is assumed that a waste collector or other person finds a discarded

source and carries it in his or her pocket for 3 hours before storing it in an out-of-the-way place. 

Assuming that a 1.5-GBq (40-mCi) source of 55Fe or 1.9-GBq (50-mCi) source of 109Cd had

decayed for 3 years, then the amount of 55Fe and 109Cd in the discarded sources would be

about 740 MBq (20 mCi) and 370 MBq (10 mCi), respectively. The EDE rates to the whole

body are estimated to be essentially zero (0) and 0.036 mSv/h (36 mrem/h) for the discarded

55Fe and 109Cd sources, respectively. Dose equivalent rates to the skin of the whole body are

estimated to be 1.1 Sv/h (110 rem/h) and 0.15 Sv/h (15 rem/h) for the discarded 55Fe and 109Cd,

respectively. Dose equivalent rates to skin are based on calculations for a separation distance

of 1 cm between the source and skin. EDE rates are based on calculations at a body depth of

10 cm, which is considered a reasonable approximation for the average depth of the body

organs relative to a small source on the surface of the body. (Refer to Appendix A.4). Because

of the small area of skin irradiated by a small source on the body’s surface, the contribution of

the skin dose to the EDE for the whole body is 0.006 mSv/h (0.06 mrem/h) for 55Fe and

8×104

 mSv/h (0.08 mrem/h) for 109Cd, assuming a 10cm2 exposed skin area and a skin

weighting factor of 0.01 (ICRP60). The total EDE is estimated to be 0.02 mSv (2 mrem) for a

discarded 55Fe source and 0.1 mSv (10 mrem) for a 109Cd source, carried in the pocket for

3 hours. However, the dose equivalent to a small area of skin on the body’s surface could be

as much as 0.4 Sv (40 rem) for the discarded 109Cd source and 3 Sv (300 rem) for the

discarded 55Fe source, assuming minimal shielding by articles of clothing or other materials

between the source and skin surface.
4–50

4.4.5 Summary

Table 4.4.1 presents the results of the analysis of potential radiological impacts for an

exemption that would allow for X-ray fluorescence analyzers containing 1.5 GBq (40 mCi) of

55Fe or 1.9 GBq (50 mCi) of 109Cd. The radiation doses estimated in this assessment are based

on typical designs of devices distributed under the requirements applicable to distributors and

applicants for license to distribute such devices for use under 10 CFR 31.5. The details of the

designs are important to ensuring control of exposure. For the radiation doses resulting under

an exemption of these devices to be comparable to those estimated, similar controls over the

distributors would be necessary to ensure that the designs of the devices are comparable in

minimizing exposures to users and members of the public.

Annual EDEs to individuals are estimated to be 0.05 mSv (5 mrem) for distribution and

transport; 0.3 mSv (30 mrem) for routine use involving sample analysis, transport for field use,

and source replacement; and 0.008 mSv (0.8 mrem) for disposal in landfills and by incineration. 

For all of these activities combined, the collective EDE to all users and members of the public is

estimated to be approximately 0.6 person-Sv (60 person-rem). This collective dose estimate

assumes a 3-year effective lifetime for the 55Fe and 109Cd sources and an annual distribution of

700 sources initially containing 1.5 GBq (40 mCi) each of 55Fe and 1600 sources initially

containing 1.9 GBq (50 mCi) of 109Cd. For accidents involving fire or leakage of radioactive

material from a source, the maximum EDEs to individuals are estimated to be 0.02 mSv

(2 mrem) and 0.7 mSv (70 mrem), respectively.

For the situation in which the sources can be replaced by the user and disposed of as ordinary

waste, a potentially serious problem has been identified that results from the potential loss of

control and accountability over the discarded sources. If a waste collector or other person

accidentally carried a discarded source in a pocket for as few as 3 hours, the dose equivalent to

the skin could be as much as 3 Sv (300 rem) (see footnotes to Table 4.4.1). Such a skin dose

could cause minor radiation burns to the skin (Potten, 1985).
4–51

Table 4.4.1 Potential Radiation Doses From X-Ray Fluorescence Analyzers

Using 55Fe and 109Cd Sources

Exposure Pathway

Individual Annual

Effective

Dose Equivalent

(mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport 5c 0.6

Routine use

 Sample analysis

 Transport

 Source replacement

 20d

 7e

 0.9f

 60

3

 1

Disposal as ordinary trash

 Landfills

 Incinerators

 0.2g

 0.6h

 0.5

 0.09

Accidents and misuse

 Accidents involving fire

 Carrying source in pocket

 Source leakage

 2i

 10j

 70k

a

 1mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Collective doses for 1 year’s distribution. Refer to text for time period of collective dose

calculation.

c

 Dose estimate applies to local parcel-delivery driver; dose estimates would be less for

long-haul semi-truck drivers, terminal workers, and members of the public along truck routes

(see Section 4.4.4.1).

d

 Dose estimate applies to whole-body irradiation of a user; dose estimate for the annual dose

equivalent to fingers of a user is 2.5 mSv (250 mrem) (see Section 4.4.4.2.1).

e

 Dose estimate applies to user; dose estimates are significantly less for members of the public

from site-to-site transport during routine usage (see Section 4.4.4.2.2).

f

 Dose estimate applies to whole-body irradiation of a user; dose estimate for hands of a user is

5 mSv (500 mrem) (see Section 4.4.4.2.3).

g

 Dose estimate applies to either waste collectors or workers at landfills. Dose estimates for

future on-site residents or off-site residents from well water ingestion are essentially zero (0),

and dose estimates for off-site residents from inhalation are zero (0) because the 55Fe and

109Cd sources are assumed to remain intact during landfill disposal (see Section 4.4.4.3).

h

 Dose estimate applies to waste collectors at incinerators; dose estimates are less for workers

at incinerators and members of the public near incinerator sites (see Section 4.4.4.3).

i

 Dose estimate applies to inhalation of radioactive material by a person escaping from a

residential fire or local neighbor trying to rescue a person from a residential fire; dose estimates

are less to firefighter or worker during cleanup following the fire (see Section 4.4.4.4).

j

 Dose estimate applies to whole-body irradiation of an individual who carries a discarded

3-year-old source in his her pocket for 3 hours; dose estimate for a small area of skin on whole

body is 3 Sv (300 rem) (see Section 4.4.4.4).

k

 Dose estimate applies to a routine user who ingests radioactive material from a leaking source

(see Section 4.4.4.4).

4–53

4.5 Calibration and Reference Sources

4.5.1 Introduction

Calibration and reference sources may be made from solid byproduct material that is either

encapsulated, embedded in another material, or plated on a metal surface, and from liquid

byproduct material that is contained in sealed glass vials to prevent leakage or dispersion of the

materials during normal handling and usage. A person who acquires, receives, possesses,

owns, uses, or transfers such sources may be exempted from licensing under conditions

discussed in 10 CFR 30.15(a)(9), Ionizing Radiation Measurement Instruments (see Section

2.10) or 10 CFR 30.18, Exempt Quantities (see Section 2.13). Other calibration and reference

sources are used under either a general license or a specific license as described in

10 CFR 30.31. This section deals with the potential exemption of certain calibration or

reference sources that are currently being distributed for use under a general or specific

license.

A general license is granted in 10 CFR 31.5 to acquire, receive, possess, use, or transfer

byproduct material contained in devices designed and manufactured for a number of specific

purposes, including measuring radiation or producing light. Included in the general license are

requirements for labeling, leak testing, and proper storage and disposition of the device. The

licensee is also subject to terms and conditions set forth in 10 CFR 31.2 dealing with general

license requirements, transfer of byproduct material, reporting and recordkeeping, and

inspection. Leak testing is required except for devices containing only krypton, devices

containing only tritium or not more than 3.7 megabecquerel (MBq) (100 microcurie (Ci)) of a

beta- and/or gamma-emitting material or 0.37 MBq (10 Ci) of an alpha-emitting material, and

devices held in storage in the original container prior to installation.

Examples of byproduct materials being distributed for use under 10 CFR 31.5 are sealed 133Ba

or 152Eu sources for calibration of a liquid scintillation counter and 14C contained in a phosphor

for use as a reference light source. There are no limits in 10 CFR 31.5 on the amount of

byproduct material that may be used in a calibration or reference source in a device. However,

an applicant for a specific license to manufacture or initially transfer such devices for use under

10 CFR 31.5 must demonstrate that they will meet certain requirements contained in

10 CFR 32.51. These requirements are described below:

 The device can be safely operated by persons without training in radiological protection.

 Under ordinary conditions of handling, storage, and use of the device, the byproduct

material contained in the device will not be released or inadvertently removed from the

device, and it is unlikely that any person will receive in any 1-year period a total effective

dose equivalent (EDE) in excess of 5 millisieverts (mSv) (500 mrem), or the sum of the

deep-dose equivalent and committed dose equivalent to any individual organ or tissue

other than the lens of the eye in excess of 50 mSv (5 rem); an eye dose equivalent in

excess of 15 mSv (1.5 rem); or a shallow-dose equivalent in excess of 50 mSv (5 rem)

to the skin or to any extremity (i.e., hand, elbow, arm below the elbow, foot, knee, or leg

below the knee).
17 CONDOS calculations for 14C have been reduced by a factor of 20 to correct for its over

estimation for low energy bremsstrahlung radiation. (Refer to Appendix A.4).

4–54

 Under accident conditions (such as fire and explosion) associated with handling,

storage, and use of the device, it is unlikely that any person would receive an external

dose equivalent or committed internal dose equivalent in excess of 0.15 Sv (15 rem) to

the whole body, head and trunk, active blood-forming organs, gonads, or lens of the

eye; 2 Sv (200 rem) to the hands and forearms, feet and ankles, or localized areas of

skin averaged over areas no larger than 1 cm2

; and 0.50 Sv (50 rem) to any other

organs.

A specific license under 10 CFR 30 is required when a device containing an internal calibration

or reference source does not meet the above conditions (e.g., the device cannot be operated

safely by persons without training in radiological protection) or the calibration or reference

source is not incorporated within a device and contains more than a quantity of byproduct

material as defined in 10 CFR 30.71, Schedule B (e.g., the loose calibration or reference

source contains more than 0.037 MBq (1 Ci) of 60Co, 0.37 MBq (10 Ci) of 137Cs, etc.). 

Exposures to individuals working near such devices and sources are monitored routinely, any

excessive doses to these individuals are detected and appropriate action taken to reduce

unwarranted exposures.

Some calibration or reference sources (either loose or internal to a device) are potential

candidates for exemption from the general and specific licensing requirements set out in

10 CFR 30 through 36 and 39. This assessment evaluates the potential radiation doses that

could result if the acquisition, receipt, possession, use, and transfer of certain of these

calibration and reference sources were exempt from licensing. The assumed conditions for this

potential exemption are loose or internal calibration and reference sources in the form of sealed

or plated sources containing not more than 10 times a quantity of byproduct material as defined

in 10 CFR 30.71, Schedule B. For devices with internal sources, the limit of 10 times a quantity

is assumed to apply to both the individual sources and the total within a device.

The quantity of a long half-life radionuclide commonly used in either calibration or reference

sources is determined primarily from considerations of the internal dose from intake (see

Section 2.13.1). The external dose rate from a single exempt quantity of a long half-life

radionuclide is typically less than 0.01 mSv/h (<1 mrem/h) at a distance of 10 cm and less than

0.1 mSv/h (<10 mrem/h) at 10 cm from a source containing 10 quantities. For example, the

dose rates as calculated by CONDOS (Computer Codes, O’Donnell et al., 1975) at 10 cm from

a source containing 10 quantities of either 14C, 133Ba, or 152Eu are approximately 1×105

 mSv/h

(0.001 mrem/h)17, 0.01 mSv/h (1 mrem/h), and 0.004 mSv/h (0.4 mrem/h), respectively. Also

considered was a generic source containing 10 quantities with a dose rate of 0.1 mSv/h

(10 mrem/h) at a distance of 10 cm to assess potential doses from external exposure during

routine use of a single calibration or reference source (or a combination of sources within a

single device) under this potential exemption.

Reports of leaking sources, submitted by licensees under the requirement of 10 CFR 31.5(c)(5)

between 1990 and 1996 and contained in the Nuclear Materials Events Database (NMED)

(Nuclear Regulatory Commission (NRC) Databases, NMED, Reports), do not include any cases

of leaking sources in liquid scintillation counters or thermoluminescent dosimetry readers. 
18 Examples of thermoluminescent materials include thallium-doped lithium flouride, LiF(Tl),

used in personnel dosimeters, and dysprosium-doped calcium sulfate, CaSO4(Dy), used in

environmental dosimeters.

4–55

Thus, external exposures to gamma rays, X-rays, and bremsstrahlung from the sources

dominate except in cases such as accidents, in which the source integrity may be

compromised. Similarly, for byproduct material contained in loose calibration sources, there is

no inhalation or ingestion concern during routine usage. The principal exposure pathway is

external irradiation of the whole body.

4.5.2 Description of Items Considered for Exempt Distribution

The calibration and reference sources that may be distributed under this potential exemption

include: (1) reference light sources contained in thermoluminescent dosimeter readers,

(2) calibration sources contained in liquid scintillation counters, and (3) loose calibration and

reference sources for general usage in instrument work or in research and teaching.

4.5.2.1 Thermoluminescent Dosimetry Readers

Thermoluminescent dosimetry (TLD) readers determine the radiation dose to an exposed piece

of thermoluminescent material18 by measuring the light output as the material is heated

(Duftschmid et al., 1986). Reference lights, described in a safety evaluation provided by one

manufacturer, contain either 2.2-MBq (60-Ci) or 8.9-MBq (240-Ci) sources of 14C and are

used for verifying instrument gain and stability and for troubleshooting. More sophisticated TLD

readers may contain up to four of the smaller (2.2-MBq (60-Ci)) reference lights, but no more

than one of the larger (8.9-MBq (240-Ci)) lights is used in a reader. TLD reader systems

incorporating these 14C reference lights have been manufactured since the early 1970s with no

reports of fracturing of the light source or inadvertent release of the hermetically sealed

radioactive material.

It is assumed that a nominal 60 TLD readers are sold annually by a manufacturer. The number

of instruments distributed annually by other manufacturers is believed to be about one-third of

this quantity. Since the TLD readers are configured to hold either a single 8.9-MBq (240-Ci)

source or up to four 2.2-MBq (60-Ci) sources, we assume that the configurations are equally

distributed among the five possibilities (those containing a single large source or one, two,

three, or four of the smaller sources). The maximum amount of 14C in a reader is 8.9 MBq

(240 Ci), and the average amount under these assumptions is 6.3 MBq (170 Ci). In this

assessment, however, it is assumed that each of the 80 TLD readers distributed annually

contains 10 times a quantity (see 10 CFR 30.71, Schedule B) or 37 MBq (1000 Ci) of 14C (see

Table 4.5.1).

Reference lights containing 90Sr/90Y are reportedly used in another TLD reader (Spanne, 1973),

but the source activity is not reported, and there is no evidence that this design is still used or

that the instrument was sold in the United States. Exposure from the 90Sr/90Y light sources was

observed as the thermoluminescent response of quartz components in the sample chamber of

the readers. No such thermoluminescence was reported in components of readers in which 14C

was used in the light source. The increased exposure potential from 90Sr/90Y is explained by the

ability of their more energetic beta particles to escape the scintillator matrix and impart radiation
4–56

dose to the surrounding materials, whereas the less energetic beta particles of 14C are unable

to escape the scintillator matrix to any appreciable degree.

4.5.2.2 Liquid Scintillation Counters

Liquid scintillation counters (LSCs) measure light emitted by a scintillator medium in which

radioactive materials are intimately dispersed and estimate the concentration of the radioactive

material from the light intensity. One manufacturer uses an LSC in which an external

gamma-ray source containing 0.74 MBq (20 Ci) of 152Eu is brought into close contact with the

sample vial containing the scintillating medium to provide a reproducible calibration and

reference light. The byproduct material is sealed within a 0.35-mm stainless steel capsule. The

capsule is further contained within a stainless steel cable cap by crimping the cap around a

stainless steel plug and, behind the plug, to a steel cable. The sealed source can only be used

in the LSC for which it was designed and not as part of another product. The source is

normally stored in a lead shield (40-mm wall thickness), except when extended into the LSC

sample chamber by the steel cable. The sample chamber is also enclosed by a lead shield, so

that the only time that the source is unshielded (except by the instrument housing) is during

transit from storage to the sample cell. Another manufacturer of LSCs uses a 133Ba or 137Cs

source in a similar configuration.

It is assumed that a nominal 600 LSCs using 0.37 to 0.74 MBq (10 to 20 Ci) of 133Ba are

distributed annually by one manufacturer. The average unit quantity of 133Ba is assumed to be

the approximate arithmetic average of these values, or 0.56 MBq (15 Ci). LSCs using

0.74 MBq (20 Ci) of 152Eu are not as prevalent in the United States, with only 20 to 25

distributed annually by a single manufacturer. In this assessment, however, it is assumed that

each of the LSCs contains 10 times an exemption quantity (see 10 CFR 30.71, Schedule B),

which is either 0.37 MBq (10 Ci) of 152Eu or 3.7 MBq (100 Ci) of 133Ba (see Table 4.5.1).

The two different LSC manufacturers report that both 226Ra and 241Am were once used in

calibration sources for these devices, but that 241Am is no longer in use and no 226Ra has been

distributed for this purpose for the past 6 or 7 years.

4.5.2.3 Loose Calibration Sources

Loose calibration sources that could be distributed under this potential exemption are primarily

gamma-ray emitters such as 60Co or 137Cs/137mBa and beta-particle emitters such as 90Sr/90Y or

204Tl (NCRP 112). The sources are small in physical size, namely, less than a few centimeters

in maximum dimensions. The byproduct material is either encapsulated or plated on metal and

covered with a thin window to prevent the spread of contamination during normal usage. For

calibration sources emitting beta particles, a thin window is necessary to permit the emergence

of a useful fraction of the radiation from the source. Extreme care in handling is vital to prevent

window damage to a beta-particle source.

The estimates in Table 4.5.2 for the number of loose calibration sources and the amount of

byproduct material distributed annually under this potential exemption were obtained as follows. 

First, it was assumed that the amount of byproduct material distributed annually for use under

an exemption would be the same as that distributed annually for use under 10 CFR 30.18 (see

Section 2.13, Table 2.13.1). Second, it was assumed that the amount of byproduct material per

source would be 10 times a quantity as specified in 10 CFR 30.71, Schedule B. Third, it was
4–57

assumed that the number of calibration sources for beta particles (i.e., 90Sr and 204Tl) would be

the same since very little 90Sr appears to be used under 10 CFR 30.18 (see Section 2.13,

Table 2.13.1). Fourth, it was assumed that two manufacturers produce an equal number of the

various loose calibration sources.

Assuming a quality factor of 1 for beta particles and photons (International Commission on

Radiological Protection (ICRP) 26), the dose rates from photons at 10 cm from the center of a

small encapsulated source containing 0.37 MBq (10 Ci) of 60Co or 3.7 MBq (100 Ci) of

137Cs/137mBa are approximately 0.01 mSv/h (1 mrem/h) and 0.03 mSv/h (3 mrem/h), respectively

(as derived from Table 4.3 in National Council on Radiation Protection and Measurements

(NCRP 112). Dose rates from beta particles at 10 cm from the radioactive surface of a small

plated source containing 37 kBq (1 Ci) of 90Sr/90Y or 3.7 MBq (100 Ci) of 204Tl are

approximately 0.08 mSv/h (8 mrem/h) and 4.5 mSv/h (450 mrem/h), respectively (as derived

from Table 5.1 in NCRP 112).

4.5.3 Summary of Previous Analyses and Assessments

There are no known previously published analyses or assessments of the radiation doses to

personnel using calibration or reference sources (either loose or internal to a device). However,

each applicant for a specific license to manufacture or initially distribute devices containing a

calibration or reference source for use under 10 CFR 31.5 is required to submit information

(i.e., a safety analysis) to the NRC to show that their product will meet the dose criteria

summarized in Section 4.5.1.

4.5.4 Present Assessment for Calibration and Reference Sources Distributed Internal to

a Device

Table 4.5.3 presents the results of the present assessment of the potential doses to members

of the public from calibration or reference sources distributed internal to a device under a

general license. Results are based on the annual distribution data in Table 4.5.1 and the

following useful lifetimes for the various devices and sources. The useful lifetime was assumed

to be 10 years for devices with internal sources containing 133Ba or 152Eu and 15 years for

devices with internal sources containing 14C.

The amount of activity per device in Table 4.5.1 is set equal to 10 times a quantity of a

byproduct material as defined in 10 CFR 30.71, Schedule B. These data are used to estimate

potential doses to individuals exposed to multiple sources during transport and disposal and to

estimate collective doses from all potential exposure pathways. Dose rates from photons at a

distance of 10 cm from the sources in Table 4.5.1 are less than the potential dose rates allowed

by this general license. Hence, also considered was a generic source containing 10 quantities

with a dose rate of 0.1 mSv/h (10 mrem/h) at a distance of 10 cm to assess the potential doses

from external exposure during routine use of a single source (or combination of sources within

a single device) as a potential exemption (see Section 4.5.1).

4.5.4.1 Distribution and Transport

A relatively small number of internal calibration and reference sources are expected to be

distributed under this potential exemption (see Table 4.5.1). Hence, the sources were
4–58

considered to be fabricated on demand and shipped directly to the user without intermediate

storage in a warehouse.

The individual and collective doses are based on the generic distribution methodology in

Appendix A.3 and the following assumptions. The distribution involves five steps: (1) express

delivery (small truck) from the manufacturer to a nearby airport, (2) processing at airport freight

terminal and loading on the outbound plane, (3) transport by plane, (4) unloading of the plane

and processing at the receiving airport, and (5) local delivery (small truck, within 400 km of the

airport) to the user.

Individual doses were evaluated based on the greatest annual quantity shipped by a single

manufacturer. A single driver is assumed to transport all sources in a small truck from a given

manufacturer to the same outbound airport. It is further assumed that the shipments are

distributed equally to 25 regional airports and that the sources are picked up at the receiving

airports and delivered to users by many local delivery drivers.

For calibration and reference sources distributed internal to devices, the shipment of sources,

individual doses, and collective doses can be summarized as follows:

 Sixty 14C sources containing 37 MBq (1000 Ci) each are distributed annually by a

single manufacturer (see Section 4.5.2.1), and the total distributed annually by all

manufacturers is 80 sources (see Table 4.5.1). The annual EDE to a local

express-delivery driver would be less than 1×105

 mSv (<0.001 mrem). Individual doses

to other truck drivers, terminal workers, and members of the public would also be less

than 1×105

 mSv (<0.001 mrem). The annual collective EDE to all truck drivers, terminal

workers, and members of the public is estimated to be 1×106

 person-Sv

(1×104

 person-rem).

 Six-hundred 133Ba sources containing 3.7 MBq (100 Ci) each are distributed annually

by a single manufacturer (see Section 4.5.2.2). The annual EDE to a local

express-delivery driver could be 0.02 mSv (2 mrem). Individual doses are less to other

truck drivers, terminal workers, and members of the public. The annual collective EDE

to all drivers, terminal workers, and members of the public is estimated to be

0.001 person-Sv (0.1 person-rem).

 Twenty-five 152Eu sources containing 37 kBq (10 Ci) each are distributed annually by a

single manufacturer (see Section 4.5.2.2). The annual EDE to a local express-delivery

driver could be 3×104

 mSv (0.03 mrem). Individual doses are less to other truck

drivers, terminal workers, and members of the public. The annual collective EDE to all

drivers, terminal workers, and members of the public is estimated to be

2×105

 person-Sv (0.002 person-rem).

4.5.4.2 Routine Use

Devices such as thermoluminescent dosimeter readers and liquid scintillation counters

distributed for use under the general license provided in 10 CFR 31.5 are used primarily by

technicians, educators, researchers, and students. The sources of byproduct material in these

devices are kept in a normally shielded storage position, reducing exposure to an estimated 1%

of the unshielded values, except when specifically employed for calibration or to provide a
4–59

reference light, which is assumed to be the case about 10% of the time. These devices are

relatively large and the sources of byproduct material typically are found near the center, well

away from the cabinet enclosure, even when in use. Thus, the shielding by the cabinet

enclosure and other parts of the device are assumed to reduce the radiation dose from the

internal source in its calibration or reference-light position to 10% of that from an unshielded

source at the same distance.

Operation is typically automatic and initiated from a computer console, so that the operator’s

closest proximity to the radioactive sources is during sample loading and unloading. Other

duties may usually be performed after loading samples and initiating the analyses as the device

performs without operator intervention. This evaluation assumes that 20% of the operator’s

time is spent loading and unloading samples with both the whole body and hands at 0.5 meter

from the shielded reference source, and that an additional 20% of the operator’s time is spent

1 meter from the source at the computer console, with the source in the calibration or

reference-light position half of the time. The remainder of the operator’s time is spent

performing other tasks at an average distance of 2 meters from the shielded source. Two other

room occupants are also assumed to perform tasks at an average distance of 2 meters from

the source in either its storage and calibration or reference-light positions.

If the radiation doses at various distances from an unshielded source are calculated with 

CONDOS (Computer Codes, O’Donnell et al., 1975), so that bremsstrahlung is taken into

account in the calculations and both the operator and other two room occupants spend

1000 h/yr working close to the device, then the following results are obtained:

 For a device with an internal 37-MBq (1000-Ci) source of 14C, the annual EDE to the

operator and to each of the other two occupants of the room would be less than

1×105

 mSv (<0.001 mrem). For the yearly distribution of 80 such devices, it is

estimated that the collective EDE is 2×107

 person-Sv (2×105 person-rem) over the first

year of routine use and 3×106

 person-Sv (3×104 person-rem) over the estimated

15-year lifetime of these devices.

 For a device with an internal 3.7-MBq (100-Ci) source of 133Ba, the annual EDE could

be 0.003 mSv (0.3 mrem) to the operator and 6×104

 mSv (0.06 mrem) to each of the

other two occupants of the room. For the yearly distribution of 600 such devices, it is

estimated that the collective EDE is 0.003 person-Sv (0.3 person-rem) over the first year

of routine use and 0.02 person-Sv (2 person-rem) over the estimated 10-year lifetime of

these devices.

 For a device with an internal 0.37-MBq (10-Ci) source of 152Eu, the annual EDE could

be 9×104

 mSv (0.09 mrem) to the operator and 2×104 mSv (0.02 mrem) to each of the

other two occupants of the room. For the yearly distribution of 25 such devices, it is

estimated that the collective EDE is 3×105

 person-Sv (0.003 person-rem) over the first

year of routine use and 2×10-4 person-Sv (0.02 person-rem) over the estimated 10-year

lifetime of these devices.

The above results give a potential collective dose of about 0.02 person-Sv (2 person-rem) from

routine use of 1 year’s distribution of devices under this potential exemption (see Table 4.5.3). 

To assess the potential individual dose from routine use of a single device, a device containing

a generic source with a dose rate of 0.1 mSv/h (10 mrem/h) at a distance of 10 cm (see Section
4–60

4.5.1) was considered. If the dose rate varies inversely with the square of the distance from the

source, then the potential individual dose to an operator of a device containing such a source

could be 0.02 mSv/yr (2 mrem/yr) and the potential dose to each of the other two occupants in

the room could be 0.005 mSv/yr (0.5 mrem/yr) (see Table 4.5.3).

4.5.4.3 Disposal

Although these calibration and reference sources would be discarded as radioactive waste

under the current regulatory scheme, for purposes of evaluating a possible exemption, all

sources distributed are assumed to be disposed as ordinary waste, as there are usually no

controls over disposal under an exemption.

To estimate the potential doses from the disposal of these sources as ordinary waste in landfills

and incinerators, the generic disposal methodology in Appendix A.2 is used along with the

following assumptions: (1) each 14C source is discarded at the end of 15 years and contains

essentially its full initial activity of 37 MBq (1000 Ci), (2) each 133Ba source is discarded at the

end of 10 years and contains 50% of its initial activity of 3.7 MBq (100 Ci), and (3) each 152Eu

source is discarded at the end of 10 years and contains 60% of its initial activity of 0.37 MBq

(10 Ci). Thus, the 80 sources of 14C would contain 3 GBq (80 mCi) at the time of disposal, the

600 sources of 133Ba would contain 1.1 GBq (30 mCi), and the 25 sources of 152Eu would

contain 56 MBq (1.5 mCi). It is assumed that 80% of the sources are disposed in a landfill and

20% are incinerated. Since the number of sources disposed of annually is less than 3500, the

assumed number of landfills, the applicable dose-to-source ratio (DSR), for individual dose in

Appendix A.2 are multiplied by the ratio of 3500 to the number of items annually disposed. A

similar correction is made for incineration.

4.5.4.3.1 Landfill Disposal

In applying the methodology of Appendix A.2 to disposal at landfills, it is further assumed that

(1) the byproduct material in the calibration and reference sources is in a form that is not readily

dispersible and (2) the sources are not handled directly by waste collectors or by workers during

landfill operations. Thus, the following adjustments are made to the dose-to-source ratios in

Appendix A.2: (1) there is no exposure from inhalation or ingestion by waste collectors or landfill

workers, (2) there is no exposure to off-site members of the public during landfill operations,

(3) there is a reduction by a factor of 10 in the exposure to off-site members of the public from

groundwater releases, and (4) there is a reduction by a factor of 10 in the exposure to future

on-site residents by inhalation and ingestion.

Estimates of potential individual and collective doses from landfill disposal of 14C can be

summarized as follows:

 The annual EDE to off-site members from groundwater releases, waste collectors,

workers at landfills, off-site members of the public exposed to airborne releases during

landfill operations, and future on-site residents, would be less than 1×105

 mSv

(<0.001 mrem).

 The total collective EDE could be 4×106

 person-Sv (4×104 person-rem), due almost

entirely to exposure to off-site members of the public from groundwater releases.
4–61

Estimates of potential individual and collective doses from landfill disposal of 133Ba can be

summarized as follows:

 The annual EDE to waste collectors could be 6×104

 mSv (0.06 mrem). For workers at

landfills, off-site members of the public near landfills, and future on-site residents, the

individual doses would be less.

 The total collective EDE could be 5×104 person-Sv (0.05 person-rem), due about

equally to exposure to waste collectors and landfill workers.

Estimates of potential individual and collective doses from landfill disposal of 152Eu can be

summarized as follows:

 The annual EDE to waste collectors could be 2×104

 mSv (0.02 mrem). For workers at

landfills, off-site members of the public near landfills, and future on-site residents, the

individual doses would be less.

 The total collective EDE could be 1×105

 person-Sv (0.001 person-rem), due about

equally to exposure to waste collectors, landfill workers, and future on-site residents.

4.5.4.3.2 Incineration

In applying the methodology of Appendix A.2 to disposal by incineration, it is also assumed that

there is no exposure to waste collectors by either inhalation or ingestion. However, it is

assumed that all of the various pathways of exposure to workers and off-site members of the

public are fully operative.

Estimates of potential individual and collective doses from incineration of 14C can be

summarized as follows:

 The annual EDE to workers at incinerators, collectors at incinerators and off-site

members of the public near landfills could be less than 1×105 mSv (<0.001 mrem).

 The total collective EDE could be 2×109

 person-Sv (2×107 person-rem), due almost

entirely to exposure to off-site members of the public from airborne releases during

incinerator operations.

Estimates of potential individual and collective doses from incineration of 133Ba can be

summarized as follows:

 The annual EDE to waste collectors could 6×104

 mSv (0.06 mrem). For workers at

incinerators and off-site members of the public near incinerators, the individual doses

would be less.

 The total collective EDE could be 6×10-5 person-Sv (0.006 person-rem), due almost

entirely to exposure to waste collectors at incinerators.
4–62

Estimates of potential individual and collective doses from incineration of 152Eu can be

summarized as follows:

 The annual EDE to waste collectors could be 2×104

 mSv (0.02 mrem). For workers at

incinerators and off-site members of the public near incinerators, the individual doses

would be less.

 The total collective EDE could be 1×10-6 person-Sv (1×104 person-rem), due almost

entirely to exposure to waste collectors at incinerators.

4.5.4.4 Accidents and Misuse

Devices containing byproduct material in either calibration or reference sources are used

primarily in industry and education, rather than in homes or small businesses. Thus, the

following exposure scenarios are considered: (1) transportation accidents involving fires,

warehouse fires, and laboratory fires, and (2) misuse of a calibration or reference source during

repair or attempted modification of a device by an unqualified individual.

Doses for transportation accidents involving fires, warehouse fires, and residential fires can be

estimated by using the generic accident methodology in Appendix A.1. Doses from a laboratory

fire can be estimated by using dose-to-source ratios for a residence and correcting for different

volumes and air exchange rates. Inhalation and submersion doses for a laboratory fire are

essentially equal to those for a residential fire, whereas resuspension doses are approximately

three times greater for the laboratory. It is assumed here that only a single device is involved

and that the release fraction is 0.01%, since the byproduct material is enclosed within the

device as a sealed source and the sealed source may be further encased within another

subassembly of the device. Based on these assumptions and the generic accident

methodology in Appendix A.1, estimates of individual dose are summarized as follows:

 For a firefighter wearing a respirator at a transportation fire, the individual EDEs would

be less than 1×105

 mSv (<0.001 mrem) from a source containing 37 MBq (1000 Ci) of

14C, 3.7 MBq (100 Ci) of 133Ba, or 0.37 MBq (10 Ci) of 152Eu. For a worker who is

involved in the cleanup following the fire and who does not wear a respirator, the

individual EDEs would be less than 1×105

 mSv (<0.001 mrem) for all sources.

 For a firefighter wearing a respirator at a laboratory fire, the individual EDEs would be

less than 1×105

 mSv (<0.001 mrem) from a source containing 37 MBq (1000 Ci) of

14C, 3.7 MBq (100 Ci) of 133Ba, or 0.37 MBq (10 Ci) of 152Eu. For a worker who is

involved in the cleanup following the fire and who does not wear a respirator, the

individual EDEs would be less than 1×105

 mSv (<0.001) mrem for all sources.

Misuse of the calibration or reference source contained in a device might entail removal of a

source or subassembly, followed by close hand work for repair or modification by an unqualified

individual. If it is assumed that the person spends 1 week (40 hours) attempting the repair or

modification, with the trunk of the body located at an average distance of about 50 cm and the

hands located at an average distance of about 10 cm from the source, then the following results

are obtained:
4–63

 For a 37-MBq (1000-Ci) source of 14C, the EDE from irradiation of the whole body

could be about 2×105

 mSv (0.002 mrem) and the dose equivalent to the hands could be

about 4×105

 mSv (0.004 mrem).

 For a 3.7-MBq (100-Ci) source of 133Ba, the EDE from irradiation of the whole body

could be about 0.02 mSv (2 mrem) and the dose equivalent to the hands could be about

0.4 mSv (40 mrem).

 For a 0.37-MBq (10-Ci) source of 152Eu, the EDE from irradiation of the whole body

could be about 0.007 mSv (0.7 mrem) and the dose equivalent to the hands could be

about 0.2 mSv (20 mrem).

 For a generic source with a dose rate of 0.1 mSv/h (10 mrem/h) at a distance of 10 cm,

the EDE from irradiation of the whole body could be about 0.2 mSv (20 mrem) and the

dose equivalent to the hands could be 4 mSv (400 mrem).

4.5.5 Present Analysis for Loose Calibration and Reference Sources

Table 4.5.4 shows the results of the present assessment of potential radiation doses to the

public from loose calibration and reference sources. Results are based on the annual

distribution data in Table 4.5.2 and following effective lifetimes for these various sources. 

Lifetimes are assumed to be 5 years for the 60Co and 204Tl sources and 15 years for the 90Sr

and 137Cs sources.

The amount of activity per source is set equal to 10 times a quantity of a byproduct material as

defined in 10 CFR 30.71, Schedule B. These data are used to estimate potential doses to

individuals exposed to multiple sources during transport and disposal and to estimate collective

doses for all exposure pathways. Dose rates from photons at a distance of 10 cm from the

sources in Table 4.5.2 are less than the potential dose rates allowed by an exemption. Hence,

we also consider a generic source with a dose rate of 0.1 mSv/h (10 mrem/h) at 10 cm to

assess the doses from routine use of a single source under this potential exemption (see

Section 4.5.1).

The sources will probably be shipped and stored in shielded containers that provide some

protection against photons from the sources. No credit is taken, however, for any shielding

against photons from the sources. As a result, dose estimates for distribution and transport

and for routine use of these loose calibration and reference sources will be conservative.

4.5.5.1 Distribution and Transport

A relatively small number of loose calibration and reference sources is expected to be

distributed under an exemption. The same assumptions are applied here as applied in Section

4.5.4.1 for reference or calibration sources shipped internal to a device.

For loose calibration and reference sources, the shipment of sources, individual doses, and the

collective doses are summarized as follows:
4–64

 Two hundred 137Cs sources containing 3.7 MBq (100 Ci) each are distributed annually

by a single manufacturer (see Section 4.5.2.3), and the total distributed annually by all

manufacturers is 400 sources (see Table 4.5.2). The annual EDE to a local

express-delivery driver could be 0.01 mSv (1 mrem). Individual doses are less to other

truck drivers, terminal workers, and members of the public. The annual collective EDE

to all truck drivers, terminal workers, and members of the public is estimated to be

0.001 person-Sv (0.1 person-rem).

 Forty-five 60Co sources containing 0.37 MBq (10 Ci) each are distributed annually by a

single manufacturer (see Section 4.5.2.3), and the total distributed annually by all

manufacturers is 90 sources (see Table 4.5.2). The annual EDE to a local

express-delivery driver could be 0.001 mSv (0.1 mrem). Individual doses are less to

other truck drivers, terminal workers, and members of the public. The annual collective

EDE to all truck drivers, terminal workers, and members of the public is estimated to be

1×10-4 person-Sv (1×10-2 person-rem).

 Ten 204Tl sources containing 3.7 MBq (100 Ci) each are distributed annually by a single

manufacturer (see Section 4.5.2.3), and the total distributed annually by all

manufacturers is 20 sources (see Table 4.5.2). The annual EDE to all delivery drivers,

other truck drivers, terminal workers, and members of the public would be less than

1×105

 mSv (<0.001 mrem). The annual collective EDE to all truck drivers, terminal

workers, and members of the public is estimated to be 2×10-7 person-Sv

(2×10-5 person-rem).

 Ten 90Sr sources containing 37 kBq (1 Ci) each are distributed annually by a single

manufacturer (see Section 4.5.2.3), and the total distributed annually by all

manufacturers is 20 sources (see Table 4.5.2). The annual EDE to all delivery drivers,

other truck drivers, terminal workers, and members of the public would be less than

1×105

 mSv (<0.001 mrem). The annual collective EDE to all truck drivers, terminal

workers, and members of the public is estimated to be 2×109

 person-Sv

(2×107

 person-rem).

4.5.5.2 Routine Use

While the dose rate from photons at 10 cm from an unshielded reference or calibration source

is typically less than 0.1 mSv/h (<10 mrem/h), the dose rate from beta particles may be much

greater and result in a significant dose to the hands, even if the hands are only in close

proximity to the sources for very brief periods of time. Thus, potential doses from external

irradiation of the hands, by beta particles are considered first, then potential doses from

external irradiation of the whole body by photons (i.e., X-rays, gamma rays, and

bremsstrahlung) are considered.

To assess the potential beta-particle doses to the hands during routine use of loose calibration

or reference sources, the following exposure scenario was adopted: (1) an individual uses a

pair of forceps to handle the source for 1 min/day for 150 day/yr (i.e., 2.5 h/yr), and (2) the

hands are located at a distance of 10 cm from the source while using the forceps. For a source

containing 10 times the quantity of 204Tl, the dose rate from beta particles could be 4.5 mSv/h

(450 mrem/h) at 10 cm from the source (see Section 4.5.2.3), and the annual dose equivalent

to the hands could be 10 mSv (1 rem). The annual dose equivalent to the hands could easily
4–65

be 100 times greater, or 1 Sv (100 rem), if the source was handled routinely using fingers

instead of forceps.

To assess the potential photon doses to the public from routine use of loose calibration or

reference sources containing 10 times the quantity of a byproduct material, the following

exposure scenario was adopted: (1) an individual is located in the same room as the source for

1000 h/yr, and (2) the average distance between the source and exposed individual is 2 meters

(see Section 2.13.4.1). Exposure time is based on the assumption that the individual spends

half of his or her working hours during the year in the room with the source, and the assumed

distance from the source would be representative of the average distance in a typical

laboratory. Based on these assumptions and on calculations with CONDOS (Computer Codes,

O’Donnell et al., 1975), the following results were obtained:

 For a 60Co source containing 0.37 MBq (10 Ci), the annual EDE to a user could be

0.002 mSv (2 mrem). For a yearly distribution of 90 such sources, the collective EDE

could be 0.002 person-Sv (0.2 person-rem) over the first year of use or 0.01 person-Sv

(1 person-rem) over the estimated 5-year lifetime of these sources.

 For a 90Sr/90Y source containing 37 kBq (1 Ci), the annual EDE to a user could be

2×105

 mSv (0.002 mrem). For a yearly distribution of 20 such sources, the collective

EDE could be 5×107

 person-Sv (5×105 person-rem) over the first year of use or

7×106

 person-Sv (7×104 person-rem) over the estimated 15-year lifetime of these

sources.

 For a 137Cs source containing 3.7 MBq (100 Ci), the annual EDE to a user could be

0.06 mSv (6 mrem). For a yearly distribution of 400 such sources, the collective EDE

could be 0.02 person-Sv (2 person-rem) over the first year of use or 0.3 person-Sv

(30 person-rem) over the estimated 15-year lifetime of these sources.

 For a 204Tl source containing 3.7 MBq (100 Ci), the annual EDE to a user could be

2 Sv (0.2 mrem). For a yearly distribution of 20 such sources, the collective EDE could

be 4×105

 person-Sv (0.004 person-rem) over the first year of use or 1×104 person-Sv

(0.01 person-rem) over the estimated 5-year lifetime of these sources.

The above results suggest a total collective dose of about 0.3 person-Sv (30 person-rem) from

routine use of 1 year’s distribution of loose calibration and reference sources under an

exemption (see Table 4.5.4). To assess the potential individual dose from photons during

routine use of a single loose calibration or reference source, a generic source of high-energy

photons with a dose rate of 0.1 mSv/h (10 mrem/h) at a distance of 10 cm was considered. If

dose rates vary inversely with the square of the distance from the source, then the estimated

potential individual dose to a user could be 0.3 mSv/yr (30 mrem/yr).

The photon dose estimates given above could be somewhat conservative, because it does not

take into account any shielding between the source and the user. The common practice of

storing calibration sources in a shielded container could result in some reduction to dose. In

addition, the assumed exposure time could be a very conservative overestimate for a realistic

exposure situation, because the source could be stored in rooms such as teaching laboratories

that any individual would occupy infrequently.
4–66

In contrast, rooms in which calibration or reference sources are stored could be occupied on a

continuous basis for up to twice as long as the time assumed here, and the average distance

from the source to the individual could be less than 2 meters. For a maximum exposure to

2000 h/yr and an average distance from the source of 1 meter, estimated individual doses could

be increased by a factor of 8, although such doses should be less likely to occur. In addition,

multiple calibration and reference sources of byproduct material could be located in the same

room, in which case the external dose would increase in proportion to the number of sources.

Considering all of the above factors, the following conclusions seem warranted about potential

individual dose from calibration and reference sources containing 10 times a quantity of a

byproduct material from Schedule B. First, by invoking reasonable assumptions about

exposure conditions, the annual EDE to an individual from photons could be as much as

0.3 mSv (30 mrem). Second, the annual photon dose could be higher if multiple sources were

stored essentially without shielding in locations occupied by an individual during a substantial

portion of a year. Third, by invoking very pessimistic assumptions about exposure conditions

that could occur only in unusual circumstances, the annual photon dose from exposure to either

single or multiple sources could approach or exceed 1 mSv (100 mrem). Finally, annual betaparticle doses to the hands of an individual could be as much as 10 mSv (1 rem) if an individual

handles a source with forceps and as much as 1 Sv (100 rem) or more if an individual routinely

handles a source using fingers instead of forceps.

4.5.5.3 Disposal

Although these loose calibration and reference sources would be discarded as radioactive

waste under the current regulatory scheme, for purposes of evaluating a possible exemption, all

sources distributed are assumed to be disposed as ordinary waste, as there are usually no

controls over disposal under an exemption.

To estimate potential doses from the disposal of these sources as ordinary waste in landfills

and incinerators, the generic disposal methodology in Appendix A.2 is applied with the following

assumptions: (1) each 60Co source is discarded at the end of 5 years and contains 50% of its

initial activity of 0.37 MBq (10 Ci), (2) each 90Sr source is discarded at the end of 15 years and

contains 70% of its initial activity of 37 kBq (1 Ci), (3) each 137Cs source is discarded at the

end of 15 years and contains 70% of its initial activity of 3.7 MBq (100 Ci), and (4) each 204Tl

source is discarded at the end of 5 years and contains 40% of its initial activity of 3.7 MBq

(100 Ci). Thus, at the time of disposal, the 90 sources of 60Co would contain 17 MBq

(450 Ci), the 20 sources of 90Sr would contain 0.52 MBq (14 Ci), the 400 sources of 137Cs

would contain 1.04 GBq (28 mCi), and the 20 sources of 204Tl would contain 30 MBq (800 Ci). 

It is assumed that 80% of the sources are landfill disposed and 20% are incinerated. Since the

number of sources disposed of annually is less than 3500, the assumed number of landfills, the

applicable DSRs for individual dose in Appendix A.2 are multiplied by the ratio of 3500 to the

number of items annually disposed. A similar correction is made for incineration.

4.5.5.3.1 Landfill Disposal

In applying the methodology of Appendix A.2 to disposal at landfills, it is further assumed that

(1) the byproduct material in the calibration and reference sources is in a form that is not readily

dispersible and (2) the sources are not handled directly by waste collectors or by workers during

landfill operations. Thus, the following adjustments to the dose-to-source ratios in
4–67

Appendix A.2: (1) there is no exposure from inhalation or ingestion by waste collectors or landfill

workers, (2) there is no exposure to off-site members of the public during landfill operations,

(3) there is a reduction by a factor of 10 in the exposure to off-site members of the public from

groundwater releases, and (4) there is a reduction by a factor of 10 in the exposure to future

on-site residents by inhalation and ingestion.

Estimates of potential individual and collective doses from landfill disposal of 60Co can be

summarized as follows:

 The annual individual EDE to waste collectors could be 2×104

 mSv (0.02 mrem). For

workers at landfills, off-site members of the public near landfills, and future on-site

residents, the individual doses would be less.

 The total collective EDE could be 2×10-5 person-Sv (0.002 person-rem), due almost

entirely to exposure to waste collectors and landfill workers.

Estimates of potential individual and collective doses from landfill disposal of 90Sr can be

summarized as follows:

 The annual EDE to future on-site residents, collectors and workers at landfills and

off-site members of the public exposed to airborne releases during landfill operations or

to groundwater releases from the landfill, would be less than 1×105

 mSv

(<0.001 mrem).

 The total collective EDE could be 1×1011 person-Sv (1×109

 person-rem), due almost

entirely to exposure to future on-site residents.

Estimates of potential individual and collective doses from landfill disposal of 137Cs can be

summarized as follows:

 The annual EDE to waste collectors could be 6×105

 mSv (0.06 mrem). For workers at

landfills, off-site members of the public near landfills, and future on-site residents, the

individual doses would be less.

 The total collective EDE could be 3×104

 person-Sv (0.03 person-rem), due almost

entirely to exposure to waste collectors and landfill workers. 

Estimates of potential individual and collective doses from landfill disposal of 204Tl can be

summarized as follows:

 The annual EDE to waste collectors, workers at landfills, off-site members of the public

near landfills, and future on-site residents, would be less than 1×105

 mSv

(<0.001 mrem).

 The total collective EDE could be 2×108

 person-Sv (2×106 person-rem), due almost

entirely to exposure to waste collectors and landfill workers.
4–68

4.5.5.3.2 Incineration

In applying the methodology of Appendix A.2 to disposal by incineration, it is also assumed that

there is no exposure to waste collectors by either inhalation or ingestion. However, it is

assumed that all of the various pathways of exposure to workers and off-site members of the

public are fully operative.

Estimates of potential individual and collective doses from incineration of 60Co can be

summarized as follows:

 The annual EDE to waste collectors could 3×104

 mSv (0.03 mrem). For workers at

incinerators and off-site members of the public near incinerators, the individual doses

would be less.

 The total collective EDE could be 4×106 person-Sv (4×104 person-rem), due almost

entirely to exposure to waste collectors at incinerators.

Estimates of potential individual and collective doses from incineration of 90Sr can be

summarized as follows:

 The annual individual EDE to workers at incinerators, waste collectors at incinerators

and off-site members of the public near landfills, would be less than 1×105

 mSv

(<0.001 mrem).

 The total collective EDE could be 1×1011 person-Sv (1×109

 person-rem), due almost

entirely to exposure to off-site members of the public from airborne releases during

incinerator operations.

Estimates of potential individual and collective doses from incineration of 137Cs can be

summarized as follows:

 The annual EDE to waste collectors could be 8×104

 mSv (0.08 mrem). For workers at

incinerators and off-site members of the public near incinerators, the individual doses

would be less.

 The total collective EDE could be 6×105

 person-Sv (0.006 person-rem), due almost

entirely to exposure to waste collectors at incinerators.

Estimates of potential individual and collective doses from incineration of 204Tl can be

summarized as follows:

 The annual EDE to waste collectors, workers at incinerators and off-site members of the

public near incinerators, would be less than 1×105

 mSv (<0.001 mrem).

 The total collective EDE could be 6×109

 person-Sv (6×107 person-rem), due almost

entirely to exposure to waste collectors at incinerators.
4–69

4.5.5.4 Accidents and Misuse

Exposure scenarios for accidents or misuse involving loose calibration and reference sources

based on actual experience are almost unlimited. The ones considered here are: (1) a

laboratory fire involving the release of byproduct material from a source, (2) accidents or

misuse involving the crushing or rupture of a source followed by subsequent ingestion of some

of the released byproduct material, and (3) a misplaced or lost source in the folds of a desk

chair.

In the case of a laboratory fire, the generic accident methodology developed in Appendix A.1 is

applied. Potential doses from a laboratory fire can be estimated using dose-to-source ratios for

a residence and correcting for different volumes and air exchange rates. Inhalation and

submersion doses for a laboratory fire are essentially equal to those for a residential fire,

whereas resuspension doses are approximately three times greater for the laboratory. It is

assumed here that only a single source is involved and that the release fraction is 0.1%. Based

on these assumptions and the generic accident methodology in Appendix A.1, the estimates of

individual dose are summarized as follows:

 For a firefighter who wears a respirator at a laboratory fire, the EDE would be less than

1×105

 mSv (<0.001 mrem) from a single 0.37 MBq (10 Ci) 60Co source, a single

37 kBq (1 Ci) 90Sr source, a single 3.7 MBq (100 Ci) 137Cs source, or a single 3.7 MBq

(100 Ci) 204TI source.

 For a worker who is involved in the cleanup following the fire and who does not wear a

respirator, the EDE could be less than 1×105

 mSv (<0.001 mrem) for all sources.

In the case of accidents or misuse involving the crushing or rupture of a loose calibration or

reference source, the potential doses to the user of the source and to a waste collector were

considered. To estimate the potential dose to the user of the source, the generic accident

methodology developed in Appendix A.1 is applied for ingestion of radioactivity following a spill

of a radioactive material in the form of a powder. First, it is assumed that 10% of the material is

deposited on the skin of an individual and, second, that 0.1% of this deposited material would

be ingested before bathing removed the material from the body. Based on these assumptions

and the generic accident methodology of Appendix A.1 (see Table A.1.8), the individual dose to

a user could be about 3×104

 mSv (0.03 mrem) for a 0.37-MBq (10-Ci) source of 60Co,

1×104

 mSv (0.001 mrem) for a 37-kBq (1-Ci) source of 90Sr/90Y, 0.005 mSv (0.5 mrem) for a

3.7-MBq (100-Ci) source of 137Cs/137mBa, and 3×104

 mSv (0.03 mrem) for a 3.7-MBq

(100-Ci) source of 204Tl. 

To estimate the radiation dose to a waste collector, the generic disposal methodology in

Appendix A.2 (see Table A.2.1) is used. Because the dose-to-source ratios are divided by the

number of landfills in the United States, the first thing to do is multiply by 3500 (i.e., the

estimated number of U.S. landfills), then by the amount of activity in the various sources. Thus,

the potential individual dose to a waste collector from both ingestion and inhalation of byproduct

material from a crushed source would be less than 1×105

 mSv (<0.001 mrem) for a 0.37-MBq

(10-Ci) source of 60Co, a 37-kBq (1-Ci) source of 90Sr/90Y, a 3.7-MBq (100-Ci) source of

137Cs/137mBa, or a 3.7-MBq (100-Ci) source of 204Tl.
4–70

In the case of a misplaced or lost source in the folds of a desk chair, it is assumed that an

individual sits in the desk chair for 20 hours before the source is retrieved from the chair, and

that the distance from the source to the surface of the body is about 1 cm during this time. 

Based on calculations with MicroShield (Computer Codes, Grove Engineering, 1996), the EDE

rates from photon irradiation of the whole body are estimated to be 9×105

 mSv/h

(0.009 mrem/h) and 0.02 mSv (2 mrem/h) for the 3.7-MBq (100-Ci) sources of 204Tl and

137Cs/137mBa, respectively. Dose equivalent rates from photon irradiation of the skin are

estimated to be 0.005 mSv/h (0.5 mrem/h) and 3 mSv/h (300 mrem/h) for the 3.7-MBq

(100-Ci) source of 204Tl and 137Cs/137mBa, respectively. In addition, the dose equivalent rate

from beta-particle irradiation of the skin by a 3.7-MBq (100-Ci) source of 204Tl over an area of

10 cm2

 is estimated to be 0.26 Sv/h (26 rem/h), based on calculations using VARSKIN MOD2

(Computer Codes, Durham, 1992).

The dose equivalent rates to skin are from calculations for a separation distance of 1 cm

between the source and skin and a 0.7 mm cloth cover with a density of 0.4 g/cm3. The EDE

rates are from calculations at a body depth of 10 cm, which is considered a reasonable

approximation for the average depth of the body organs relative to a small source on the

surface of the body. (Refer to modeling in Appendix A.4). Because of the small area of skin

irradiated by a small source on the body’s surface, the contribution of the skin dose to the EDE

is quite small. Assuming an exposed area of 10 cm2

 and a skin weighting factor of 0.01, the

EDE is estimated to be 0.03 mSv (3 mrem) and 0.4 mSv (40 mrem) for the 204Tl and

137Cs/137mBa sources, respectively. However, the dose equivalent to a small area of skin on the

body’s surface could be as much as 60 mSv (6 rem) for the 137Cs/137mBa source and 5 Sv

(500 rem) for the 204Tl source, assuming minimal shielding by articles of clothing or other

materials between the source and skin surface.

4.5.6 Summary

Results of the assessment of potential doses for an exemption involving calibration and

reference sources either internal to a device or loose are presented in Tables 4.5.3 and 4.5.4. 

It is assumed that an exemption would allow some calibration and reference sources (either

loose or internal to a device) to contain as much as 10 times a quantity of a byproduct material

as defined in 10 CFR 30.71, Schedule B. For devices with internal sources, the limit of 10

times a quantity is assumed to apply to both the individual sources and the total within a device.

In the case of internal sources in devices, estimated doses are based on typical designs for

devices distributed under the requirements applicable to distributors and applicants for license

to distribute such devices for use under 10 CFR 31.5. The details of the designs are important

in ensuring control of exposure. For the radiation doses resulting under an exemption of these

devices to be comparable to those estimated, similar controls over the distributors would be

necessary to ensure that the designs are comparable in minimizing exposures to users and

members of the public. Removal of some requirements of a general license would not be likely

to significantly increase the number of thermoluminescent dosimetry readers and liquid

scintillation counters that are distributed annually because the specialized application of the

devices is assumed to be the limiting factor. However, the amount of byproduct material

distributed annually under this proposed exemption for sources internal to a device may

increase because some devices may incorporate sources containing 10 times the quantity of a

byproduct material such as 14C or 133Ba.
4–71

The annual EDEs to individuals from internal sources in devices could be 0.02 mSv (2 mrem)

for distribution and transport, 0.02 mSv (2 mrem) for routine use, and 6×105 mSv (0.006 mrem)

for disposal in landfills or by incineration. For all of these activities combined, the collective

EDE to all users and members of the public could be 0.02 person-Sv (2 person-rem). This

collective dose estimate assumes the annual distribution data in Table 4.5.1 and an effective

lifetime of 10 years for devices containing 133Ba or 152Eu and 15 years for devices containing

14C. For accidents involving fire, the individual EDE could be less than 1×105

 mSv

(<0.001 mrem). For misuse of an internal source in a device during repair or attempted

modification of a device by an unqualified individual, the EDE could be 0.2 mSv (20 mrem). 

Also, the estimated dose equivalent to the hands from misuse during repair or attempted

modification of a device by an unqualified individual could be 4 mSv (400 mrem).

The assessment suggests that the quantity of a long half-life radionuclide commonly used in

either calibration or reference sources is determined primarily from considerations of the

internal dose due to inhalation (see Section 2.13.1), and the external dose rate from a single

quantity of a long half-life radionuclide is typically less than 0.01 mSv/h (<1 mrem/h) at a

distance of 10 cm and less than 0.1 mSv/h (<10 mrem/h) at 10 cm from a source containing 10

quantities as specified in Schedule B(see Section 4.5.1). For example, the dose equivalent rate

at 10 cm from a source containing 10 quantities of 152Eu is about 0.004 mSv/h (0.4 mrem/h). 

Sources are already being distributed under the general license granted in 10 CFR 31.5 that

contain 20 times the quantity of 37 kBq (1 Ci) for 152Eu (see Section 4.5.2.2).

In the case of loose calibration or reference sources, the sources will probably be shipped and

stored in containers that provide some protection against photons from the sources. No credit

is taken, however, for any shielding against photons from the sources in the current dose

estimates. As a result, individual and collective dose estimates for distribution and transport

and for routine use of these loose calibration and reference sources will be conservative.

 

The annual EDEs to individuals from loose calibration and reference sources could be

0.01 mSv (1 mrem) for distribution and transport, 0.3 mSv (30 mrem) for routine use of a single

source, and 8×104

 mSv (0.08 mrem) for disposal in landfills or by incineration. For all of these

activities combined, the collective EDE to all users and members of the public could be

0.3 person-Sv (30 person-rem). This collective dose estimate assumes the annual distribution

data in Table 4.5.2 and an effective lifetime of 5 years for sources containing either 60Co or 204Tl

and 15 years for sources containing either 90Sr/90Y or 137Cs/137mBa. For accidents and misuse, it

is estimated that the individual EDE from a lost or misplaced source could be 0.4 mSv

(40 mrem) and that the dose equivalent to a small area of skin on the individual’s whole body

from the lost or misplaced source could be 5 Sv (500 rem).

The assessment also suggests that such an exemption may not provide enough protection in

limiting exposure to loose calibration or reference sources for a couple of reasons. First, the

EDE during routine use could be greater than 0.3 mSv (30 mrem) if multiple sources were

stored essentially without shielding in locations occupied by an individual during a substantial

part of the year. By invoking conservative assumptions about exposure conditions that could

occur only in unusual circumstances, the annual EDE from exposure to either single or multiple

sources during routine use could approach or exceed 1 mSv (100 mrem) (see Section 4.5.5.2). 

An annual EDE of 1 mSv (100 mrem) is equivalent to the annual dose limit for a member of the

public under the requirements of 10 CFR 20.1301. Second, such an exemption fails to control

the dose equivalent rates from sources of beta particles during routine use (see Section
4–72

4.5.5.2) or misuse (see Section 4.5.5.4), and a lost or misplaced source could deliver a dose

equivalent of several Sv (several hundred rem) to a small area of the skin on the whole body

(see Section 4.5.5.4). Such a skin dose could cause minor radiation burns to the skin if

delivered over a short time (Potten, 1985).
4–73

Table 4.5.1 Estimated Annual Distribution of Calibration and Reference Sources

Contained Internal to Devices

Estimated Annual Distributiona

Radionuclide Devices/yr Ci/deviceb Ci/yrb

14C

133Ba

152Eu

 80

600

 25

1,000

 100

 10

80,000

60,000

 2,500

a

 See Section 4.5.2.

b

 1 Ci = 0.037 MBq.

Table 4.5.2 Estimated Annual Distribution of Loose Calibration and Reference Sources

Estimated Annual Distributiona

Radionuclide Sources/yr Ci/sourceb Ci/yrb

60Co

90Sr

137Cs

204Tl

 90

 20

400

 20

 10

 1

100

100

 900

 20

40,000

 2,000

a

 See Section 4.5.2.

b

 1 Ci = 0.037 MBq.
4–74

Table 4.5.3 Potential Radiation Doses From Calibration and Reference Sources

Distributed Internal to a Device

Exposure Pathway

Individual Annual Effective

Dose Equivalent 

(mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport 2c 0.1

Routine use 2d 2

Disposal as ordinary waste

 Landfills

 Incinerators

0.06e

0.06f

0.05

0.006

Accidents and misuse

 Accidents involving fire

 Repair by unqualified person

 <0.001g

 20h

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Collective doses over effective lifetime of product for 1 year’s distribution. Refer to text for

time period of collective dose assessment.

c

 Dose estimate applies to an express-delivery driver; dose estimates would be less for other

local parcel-delivery drivers, terminal workers, and members of the public (see Section 4.5.4.1).

d

 Dose estimate applies to an operator of a device containing a generic source with a dose rate

of 0.1 mSv/h (10 mrem/h) at a distance of 10 cm; dose estimates are for operators using

devices with 14C, 133Ba, or 152Eu sources and for other persons casually exposed to these

devices (see Section 4.5.4.2).

e

 Dose estimate applies to waste collectors at landfills; dose estimates are less for workers at

landfills, off-site members of the public, and future on-site residents at landfills (see

Section 4.5.4.3.1).

f

 Dose estimate applies to waste collectors at incinerators; dose estimates are less for workers

at incinerators and off-site members of the public near incinerators (see Section 4.5.4.3.2).

g

 Dose estimate applies to all individuals for a laboratory or transportation fire. (See Section

4.5.4.4).

h

 Dose estimate applies to a generic source with a dose rate of 0.1 mSv/h (10 mrem/h) at a

distance of 10 cm and to whole-body irradiation of an unqualified person who removes source

from a device in an attempt to repair or modify device; dose estimate for irradiation of hands

during repair or modification of devices by an unqualified person is 4 mSv (400 mrem) (see

Section 4.5.4.4).
4–75

Table 4.5.4 Potential Radiation Doses From Loose Calibration and Reference Sources

Exposure Pathway

Individual Annual Effective

Dose Equivalent

(mrem)a

Collective Effective

Dose Equivalentb

(person-rem)a

Distribution and transport 1c 0.1

Routine use 30d 30

Disposal as ordinary waste

 Landfills

 Incinerators

 0.06e

 0.08f

0.2

0.006

Accidents and misuse

 Accidents involving fire

 Crushing of a source

 Misplaced or lost source

 <0.001g

0.5h

40i

a

 1 mrem = 0.01 mSv; 1 person-rem = 0.01 person-Sv.

b

 Collective doses over effective lifetime of source for 1 year’s distribution. Refer to text for time

period of collective dose assessment. 

c

 Dose estimate applies to an express-delivery driver; dose estimates are for other local

parcel-delivery drivers, terminal workers, and members of the public (see Section 4.5.4.1).

d

 Dose estimate for external irradiation of whole body by photons from a single source;

depending upon particular byproduct material, annual individual dose could be as much as 

0.3 mSv (30 mrem). Higher annual photon doses could result from exposure to multiple

sources, and more pessimistic, but relatively unlikely, assumptions about exposure conditions

to either single or multiple sources could result in annual photon doses approaching or

exceeding 1 mSv (100 mrem). Annual beta-particle doses to hands of an individual could be as

much as 10 mSv (1 rem) if an individual handles a source with forceps and as much as 1 Sv

(100 rem) or more if a person handles a source without forceps (see Section 4.5.5.2).

e

 Dose estimate applies to waste collectors at landfills; dose estimates are less for workers at

landfills, off-site members of the public, and future on-site residents at landfills (see

Section 4.5.4.3.1).

f

 Dose estimate applies to all individuals for a laboratory fire (see Section 4.5.4.3.2).

g

 Dose estimate applies to workers involved in cleanup following a laboratory fire; dose

estimates for firefighters at a laboratory fire would be less (see Section 4.5.5.4).

h

 Dose estimate applies to a user who ingests byproduct material from a crushed or ruptured

source; dose estimates for ingestion or inhalation of byproduct material from a crushed or

ruptured source by a waste collector are substantially less (see Section 4.5.5.4).

i

 Dose estimate applies to whole-body irradiation of a person from a misplaced or lost 137Cs

source in the folds of a desk chair; dose estimates for a small area of skin on the whole body

are 60 mSv (6 rem) from a 137Cs source and 5 Sv (500 rem) from a 204Tl source (see

Section 4.5.5.4).
R–1

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R–24
APPENDIX A

A.1–1

C  Q

Vkt (1 e kt) , (1)

C  K × S , (2)

A.1 GENERIC ACCIDENT METHODOLOGY

A.1.1 Introduction

A generic methodology was developed to estimate radiation doses from accidents involving

Nuclear Regulatory Commission (NRC)-exempted products and materials. The methodology

includes the following accident exposure scenarios: (1) fires involving the release of radioactive

materials from all types of products, (2) mechanical resuspension of radioactive material during

cleanup after a fire, (3) spills of radioactive materials in liquid or powder form, and (4) crushing

of glass tubes containing radioactive gases. 

For these selected accident exposure scenarios, radionuclide-specific dose-to-source ratios

(DSRs) are calculated. The DSRs give the radiation dose associated with a unit quantity of the

radioactive material at risk in an accident. In the derivation of the DSRs, the methods and

parameter values used in prior assessments addressing similar issues were used when

applicable. The methods and parameters used in the development of this generic methodology

are discussed and used in the derivation of the DSRs for a variety of radionuclides in many

exempted products. Because of the broad range of exemptions, it has been necessary to

estimate radiation doses for some exempted products or materials on a case-by-case basis

when the DSR methodology is not applicable. In some instances, radiation doses have also

been estimated on a case-by-case basis using better data when available for a specific

accident exposure scenario and product of interest.

A.1.2 Airborne Concentrations

Airborne concentrations of radioactive materials during an accident and cleanup following an

accident are estimated using two equations. For the instantaneous release of radioactive

materials during an accident, the average airborne concentration (microcurie (Ci)/m3

) is given

by

where Q = amount of radioactive material (Ci) released at t = 0, 

k = ventilation rate (h1

), 

t = time over which C is averaged (h), and 

V = volume of air into which material is dispersed (m3

). 

For cleanup following an accident involving fire, the average airborne concentration (Ci/m3

) is

given by

where S = level of contamination on a surface (Ci/m2

), and 

K = empirically determined factor for mechanical resuspension of respirable size

particles (m1

).
A.1–2

Q  RF × A , (3)

I  C × BR × t , (4)

A.1.3 Resuspension Factors

Particle resuspension from a contaminated surface has been reviewed by Healy (1980), Schmel

(1980), and Nicholson (1988). The factor used here for mechanical resuspension stresses on a

contaminated surface is 1×105

 m1. The use of this value is supported by practices used in

both radiation protection (Carter, 1983) and transportation (Humphries and Dodd, 1989).

A.1.4 Release Fractions

Release fractions are used to estimate the release of airborne radioactive material during a spill

or fire (see Table A.1.1). Hence, the amount of radioactive material (Ci) released into air in

Equation 1 is given by

where A = total amount of radioactive material involved in the spill or fire (Ci), and

RF = fraction of radioactive material released as respirable size particles (unfitness).

In the case of a fire, a release fraction of 0.1% is assumed for most materials; however, for

gases, a value of 100% is assumed, and for solid materials in protective devices, a value of

0.01% is assumed. If better data are available in a particular case (e.g., a lower release of

uranium in glassware or thorium in alloys in a fire), then those better data have been used. The

practices in transportation (International Atomic Energy Agency (IAEA) SS No. 7) and the

release fractions in 10 CFR 30.72 provide general support for the use of 0.1% in many cases.

In the case of a spill, a release fraction of 0.1% is assumed, provided only small amounts of a

simple liquid or solid and temperatures of less than 100C are involved (Environmental

Protection Agency (EPA) 520/1–89–001). If the spill involves higher temperatures and either

volatile solids or flammable liquids, the use of other release factors should be considered 

(EPA 520/1–89–001; DOE–STD–1027–92). If the spill involves a large amount of liquid, the

use of a simple release factor may not be appropriate (Martin Marietta Energy Systems

(MMES), 1992).

A.1.5 Inhalation Intakes

During a spill or fire, an individual’s intake from inhalation of an airborne radioactive material

(Ci) is given by

where t = time of exposure (h),

BR = breathing rate (m3

/h), and

C = average concentration of the airborne radioactive material (Ci/m3

) over 

the time, t.
A.1–3

Some enclosure volumes and ventilation rates used throughout this report are summarized in

Table A.1.2. A ventilation rate of 1 volume change per hour for a house is within the expected

range of 0.25 to 2 from a recent report by Koontz and Rector (EPA Contract No. 68–D9–0166). 

A ventilation rate of 1 volume change per hour for a large warehouse is within the expected

range of 1 to 4 from a report by the American Society of Heating, Refrigeration, and Air

Condition Engineers (ASHRAE) (1991). Table A.1.2 also includes data for a laboratory-type

room that is thought to be typical of those found at many industrial facilities and educational

institutions such as colleges and high schools. The enclosure volume and ventilation rate

are based on the approximate size of a 20-student laboratory or 30-student classroom

(ASHRAE, 1993). A ventilation rate of 6 volume changes per hour for a laboratory-type room is

within the expected range of 6 to 12 for a chemical laboratory at an industrial or commercial

facility (ASHRAE, 1995; National Research Council, 1995).

The breathing rate assumed for general use throughout this report is the daily average rate of

approximately 0.9 m3

/h (i.e., 22 m3/day or 8000 m3/yr). However, a breathing rate for light

exercise of 1.2 m3

/h has been used for transportation accidents involving fire (IAEA SS No. 7),

and this value has been applied in all accident scenarios. In the case of a fire inside an

enclosure, it is assumed that a firefighter wears a supplied-air respirator that is operated in a

pressure-demand or other positive-pressure mode and that provides a protection factor of 1000

(NIOSH, 1990).

A.1.6 Ingestion Intakes

For ingestion of materials in powder or liquid form, the materials presumably would not be

deliberately ingested during normal use but could be transferred to the hands during a spill and

cleanup following a spill. Once the material is transferred to an individual’s hands or skin, it can

be ingested directly or absorbed through wetted skin in the case of tritiated water.

For direct ingestion of liquids or powders, assumptions made in this report are consistent with

those used by the IAEA in assessing transportation accidents (IAEA SS No. 7). It is assumed,

first, that 10% of the available material would be deposited somewhere on the skin of an

individual and, second, that 0.1% of this deposited material would be ingested before bathing

removed the material from the body. Thus, the direct ingestion intake, I, is estimated to be

1×104

 times A, where A is the total amount of available material at risk in the accident.

For absorption of tritiated water through wetted skin, the equations of Osborne (1966) could be

applied. However, it is assumed that 10% of the tritiated water is spilled somewhere on the skin

of the individual and is completely absorbed through the skin before bathing. Thus, the intake

of tritiated water through the wetted skin is estimated to be 10% times A or 10×101 A, where A

is the total amount of tritiated water at risk in the accident.

A.1.7 Radiation Dose Estimation

Radiation doses are estimated using the effective dose equivalent (EDE) based on the

International Commission on Radiological Protection (ICRP) 26 approach that radiation risk

should be the same, whether the whole body is irradiated uniformly or several organs receive all

of the radiation dose. The ICRP 26 approach replaces the critical organ concept that was used

for many years. The EDE is the sum of the radiation doses to each organ, after the dose
A.1–4

H  DCF × I , (5)

H  DCF × C × t , (6)

equivalent for each organ is multiplied by an organ weighting factor based on somatic health

risk estimates from many studies (ICRP 26).

In addition to the radiation doses resulting from inhalation or ingestion, it is also possible to

receive a radiation dose from submersion in airborne radioactive materials and from

resuspension of any radioactive materials during cleanup following a fire. These potential

exposure pathways are all considered in the following discussions. In the case of submersion

that may result in a significant radiation dose being delivered to the skin, a skin weighting factor

of 0.01 is also used in estimating the EDE. This recommendation was made by the ICRP

following ICRP 26 (Kocher and Eckerman, 1988) and incorporated in the later

recommendations of ICRP Publication 60.

The EDE (rem) to an individual from inhalation and ingestion of a radioactive material is given

by

where I = intake of an individual by inhalation or ingestion (Ci), and

DCF = dose conversion factor for the 50-year committed EDE from inhalation or

ingestion (rem/Ci) (see Sections 2.1 and 3.1, and EPA–520/1–88–020).

The EDE (rem) to an individual from submersion in airborne radioactive material is given by

where t = time of an individual’s exposure (h),

C = average concentration of the airborne material (Ci/m3

) over the time, t, and

DCF = dose conversion factor for air submersion (rem-m3

/Ci-h) (see Sections 2.1

and 3.1, and EPA–402–R–93–081). The skin dose component may be

included using a 0.01 weighting factor.

The DSRs in the following sections of this report are the EDE (for inhalation, ingestion and

submersion), calculated per Ci of a parent byproduct material (e.g., 60Co or 137Cs) or mg of a

parent source material (e.g., natural thorium). However, when calculating the DSRs, it is

necessary to convert the mass of a parent source material (mg) to activity (Ci). For example,

the mass of natural thorium is due almost entirely to 232Th, which has a specific activity of 232Th

of 0.0403 terabecquerel (TBq/g (1.09×107

 Ci/g) (see Table 3.1.3 of Section 3.1), and the

activity of the 232Th in 1 mg of natural thorium is 4.03 Bq (1.09×104 Ci). If 20 years have

elapsed since the natural thorium was chemically separated (see Table 3.1.5 of Section 3.1),

the total activity in the 1-mg source is as follows: 4.03 Bq (1.09×104 Ci) of 232Th; 3.67 Bq

(9.92×105 Ci) of 228Ra; 3.51 Bq (9.48×105 Ci) of 228Th and 224Ra; and some additional

activity from a number of short half-life products of decay (see Table 3.1.3 of Section 3.1) that

do not contribute significantly in most of the dose calculations.
A.1–5

DSR  8.6×107



i

DCFi × RFi × Ai , (7)

A.1.7.1 Radiation Dose From Inhalation During an Accident Involving Fire

For completeness, three types of fires are considered: (1) warehouses belonging to

manufacturers or distributors that may contain large numbers of exempted products,

(2) transportation accidents in which a few cartons or pallets of exempted products are

involved, and (3) residences in which only a few exempted products are involved (see

Tables A.1.4 through A.1.6). In many of the previous assessments, residential and warehouse

fires have been considered, but not transportation fires. Transportation fires were included

here because many exempted products may be shipped to vendors or consumers without being

stored in large numbers.

For transportation accidents occurring indoors (i.e., storeroom or cargo-handling bay) or

outdoors (i.e., transportation vehicles) and involving fire, the IAEA (SS No. 7) estimates that the

inhalation intake during a 30-minute period following the start of the fire ranges from about 0.01

to 0.1%. The IAEA recommends the use of an intake factor, I, of 1.0×103 Q for a firefighter or

bystander in the plume of smoke from the fire. It is assumed that a bystander would not stand

in the plume of smoke from a fire and that a firefighter who is in the plume from the fire would

wear a supplied-air respirator. Hence, an inhalation intake factor, I, of 8.6×107 Q is used,

based on the average concentration in the air of a storeroom or cargo-handling bay for

30 minutes following the start of a fire and the use of a supplied-air respirator with a protection

factor of 1000 (see Table A.1.3).

The first column of Table A.1.4 presents the DSRs used in this report for a firefighter at a

transportation accident involving a fire. The equation for calculating the DSRs (rem/Ci or

rem/mg) is obtained using Equations (3) and (4) and is given by

where DCFi = dose conversion factor for inhalation of a radionuclide i (rem/Ci), 

RFi = release factor for a radionuclide i (unfitness), and 

A i = activity of each radionuclide i (Ci) per 1 Ci of a parent byproduct material

or 1 mg of a parent source material. 

The DSRs developed here are based on a release factor, RF, of 100% for gases (e.g., tritium

(

3

H) and 85Kr) and 0.1% for solids, powders, or liquids (see Table A.1.1). If a solid is contained

in a protective device (e.g., 241Am in a smoke detector), the DSRs for inhalation of solids in the

first column of Table A.1.4 should be reduced by a factor of 10.

For warehouse and residential fires, the same equation as above is used, except for the

numerical constant. The numerical constant used for a residential fire was 1.0×106

 and that

used for a warehouse fire was 1.6×107

. These constants are based on the inhalation intakes,

I, given in Table A.1.3 for a ventilation rate of 1 volume change per hour in both a residence

and a warehouse. It should be noted from the data in Table A.1.3 that a ventilation rate of 1 air

change per hour will predict the inhalation intake, I, to within a factor of 2 over the wide range of

ventilation rates expected for residences (EPA Contract No. 68–D9–0166) and warehouses

(ASHRAE, 1991). Warehouses are usually not air conditioned, but they are often heated and

ventilated sufficiently to provide a tolerable working situation. The DSRs developed here for
A.1–6

DSR  7.21×104 hr/m 3



i

DCFi × RFi × Ai , (8)

firefighters at warehouse and residential fires are provided in the first column of Tables A.1.5

and A.1.6, respectively. A blank space in a table means that the exposure pathway is not

operative for that particular radionuclide (e.g., 85Kr is an inert gas that is not absorbed into the

body via the inhalation pathway).

A.1.7.2 Radiation Dose From Submersion During an Accident Involving Fire

For transportation accidents involving fire, the IAEA (SS No. 7) uses an average concentration,

C, of 1.44×103 Q per m3

 for the radioactivity in air during the 30-minute exposure time following

the start of the fire (see Table A.1.3). The equation for calculating the DSRs (rem/Ci or

rem/mg) from submersion in contaminated air is obtained using Equations (3) and (6) and is

given by

where DCFi

 = dose conversion factor for submersion in a radionuclide i (rem-m3/Ci-h), 

RFi = release factor for a radionuclide i (unfitness), and

A i = activity of each radionuclide i (Ci) per 1 Ci of a parent byproduct material

or 1 mg of a parent source material.

The DSRs developed here are based on a release factor, RF, of 100% for gases (e.g., 3

H and

85Kr) and 0.1% for solids, powders, or liquids (see Table A.1.1). If a solid is contained in a

protective device (e.g., 241Am in a smoke detector), then the DSRs for submersion in air

contaminated by solids in Tables A.1.4, A.1.5, and A.1.6 should be reduced by a factor of 10.

For warehouse and residential fires, the same equation as above is used, except for the

numerical constant. The numerical constant used for a residential fire was 8.75×104

 h/m3 and

that used for a warehouse fire was 1.31×104

 h/m3. These constants are based on the average

concentration, C, given in Table A.1.3 for a ventilation rate of 1 volume change per hour in both

a residence and a warehouse. It should be noted from the data in Tables A.1.4 through A.1.6

that submersion may be an important exposure pathway for a firefighter who is wearing a

supplied-air respirator, which reduces the EDEs for inhalation by a factor of 1000 or more.

For residential fires, consideration is given to doses to a person escaping the fire or a neighbor

trying to rescue a person from a fire. Inhalation doses for a nonfirefighter will be greater than

for a firefighter who wears a supplied-air respirator and has an individual intake of 1.0×106 Q

(see value for a residence with a ventilation rate of 1 volume change per hour in Table A.1.3). 

If an escape or rescue time of 10 minutes is assumed, then individual intakes for nonfirefighters

are calculated to be 4.1×104 Q/m3

, and their inhalation doses would be 410 times greater than

that estimated for firefighters (i.e., 4.1×104 Q/1.0×106 Q). Thus, the DSRs in the first column

of Table A.1.6 can be multiplied by 410 to estimate the inhalation dose and the DSRs in the

second column of Table A.1.6 can be multiplied by 0.39 to estimate the submersion dose to a

nonfirefighter (i.e., a person escaping from a residential fire or a neighborhood hero trying to

rescue a person from a residential fire).
A.1–7

DSR  1.0×104 m 2



i

DCFi × Ai × K × BR × t , (9)

A.1.7.3 Radiation Dose From Mechanical Resuspension Following a Fire

For the cleanup following a fire, it is assumed, first, that the mechanically resuspendable activity

is 1% of the total activity involved in the fire and, second, that it is spread uniformly over the

total floor area of the enclosure involved in the fire (NUREG/CP–0001, Wrixon and Freke). 

Thus, the level of contamination, S, of a storeroom or cargo-handling bay with a floor area

of 100 m2

 (see Table A.1.2) is estimated to be 1% times A divided by 100 m2 or 1×104 A m2,

where A is the total activity involved in the fire. The equation for calculating the DSRs (rem/Ci

or rem/mg) from mechanical resuspension following a fire is obtained using Equations (2), (4),

and (5) and is given by

where DCFi

 = dose conversion factor for inhalation of a radionuclide i (rem/Ci), 

A i = activity of each radionuclide i (Ci) per 1 Ci of a parent byproduct material

or 1 mg of a parent source material, 

K = mechanical resuspension factor of 1×105

 m1,

BR = breathing rate of 1.2 m3

/h, and 

t = exposure time, which is assumed to be an 8-hour work day.

For mechanical resuspension of radioactive material following a warehouse or residential fire,

the same equation is used except for the constant value. The constant values used for a

residence and warehouse are 5.4×105

 m2 and 1.1×105 m2, respectively.

As an example of the use of the DSRs in Tables A.1.4 through A.1.6, consider a residential fire

that destroys two smoke detectors containing 37 kBq (1 Ci) each of 241Am (see Section 2.15). 

The total amount of 241Am material involved in the fire is 74 kBq (2 Ci). Because 241Am in

smoke detectors has a release rate of approximately 104

 from actual measurements

(NUREG/CR–0403), the smoke detector acts as a protective device, and the DSRs for

inhalation and submersion during the fire should be reduced by a factor of 10 (see Footnote a”

to Table A.1.6). The individual EDEs to a firefighter are estimated to be (4.4×107

 rem/Ci) ×

2 Ci ÷ by 10 or 0.009 nanosievert (nSv) (0.9 nrem) for inhalation and (9.8×1012 rem/Ci) ×

2 Ci ÷ by 10 or 2×105

 nSv (0.002 nrem) for submersion. Thus, the estimated individual EDE

to a firefighter is due almost entirely to inhalation of 241Am from the smoke detectors.

For the cleanup following a residential fire under the same conditions, the committed EDE to an

individual from inhalation of resuspended material during a work day of 8 hours is estimated to

be (2.3×106

 rem/Ci) × 2 Ci, or 0.05 Sv (5 rem). Because the spread of contamination on

the floor is considered to be independent of the release of material into the air, the factor of 10

reduction is not applied to the EDEs for the inhalation from resuspension of material on the floor

of the house. For fire inspectors who spend about 25% of their time inspecting mostly house

fires during a year (i.e., 62.5 working days), annual individual EDEs are estimated to be

0.05 Sv (5 rem) per day (8 hours) times 62.5 working days (500 hours), or 0.003 mSv

(0.3 mrem).

Finally, consider the radiation dose to a person escaping from a fire or a neighbor trying to

rescue a person from a fire under the same conditions. Their inhalation dose is estimated to be
A.1–8

DSR  1.0×103



i

DCFi × Ai × RF , (10)

DSR  1.0×104



i

DCFi × Ai , (11)

410 times that to a firefighter (i.e., 410 × 0.9 nrem), or 0.004 Sv (0.4 rem), and their

submersion dose is estimated to be 0.39 times that to a firefighter (i.e., 0.39 × 2 prem), or

0.01 pSv (1 prem).

It should be noted that for the above examples, all dose values less than 1×105

 mSv

(<0.001 mrem) would have been reported as less than values” (i.e., less than 1×105

 mSv

(<0.001 mrem)) if they had been included in a specific assessment presented in this report. 

This is consistent with the notation that was used for this report.

A.1.7.4 Radiation Dose From Spilled Liquids and Powders

For inhalation following a spill, an individual intake, I, of 1.0×103 Q is used, based on a

30-minute exposure time in a laboratory-type room with an enclosure volume of 180 m3

 and a

ventilation rate of 6 volume changes per hour (see Table A.1.7). The equation used to

calculate the DSRs (rem/Ci or rem/mg) for inhalation is obtained using Equations (4) and (5)

and is given by 

where DCFi = dose conversion factor for inhalation of a radionuclide i (rem/Ci),

A i = activity of each radionuclide i (Ci) per 1 Ci of a parent byproduct material

or 1 mg of a parent source material, and

RF = release factor of 0.1% for spills of liquids or powders.

It should be noted that the individual intake, I, of 1.0×103 Q used in the above equation also

provides an estimate of the maximum inhalation dose to an individual for longer exposure times

following the spill (i.e., 1 hour or more) at the expected ventilation rate of 6 volume changes per

hour or more in a laboratory-type room (see Table A.1.7 and Section A.1.5).

For ingestion following a spill with the exception of 3

H, an individual intake factor, I, of

1.0×104 A is assumed, where A is the total amount of activity involved in the spill (see

Section A.1.6). The equation for calculating the DSRs (rem/Ci or rem/mg) for ingestion of a

material other than 3

H is obtained using Equation (5) and is given by

where DCFi

 = dose conversion factor for ingestion (rem/Ci), and

A i = activity of each radionuclide i (Ci) per 1 Ci of a parent byproduct material

or 1 mg of a parent source material.

For 3

H, the same equation as above is used, except the numerical constant is 1.0×101 (see

Section A.1.6). Table A.1.8 presents the DSRs for both ingestion and inhalation following a spill

of a liquid or powder.
A.1–9

As an example of a dose calculation for a spill, consider a quantity of 100 Ci (0.37 MBq) of 14C

obtained as a liquid in the form of a labeled organic compound (see Section 2.13). By the use

of Equation (10), the DSR for inhalation is (1.0×103

) × (2.1×103 rem/Ci) × 1 Ci × (1.0×103),

or 5.7×1010 mSv/kBq (2.1×109

 rem/Ci) of 14C, and by the use of Equation (11), the DSR for

ingestion is (1.0×104

) × (2.1×103 rem/Ci) × 1 Ci, or 5.7×105 Sv/GBq (2.1×107 rem/Ci) of

14C. Thus, the individual EDEs for the inhalation exposure pathway, the ingestion exposure

pathway, and both pathways combined for the spill of the 0.37 MBq (100 Ci) of 14C are

estimated to be 0.002 Sv (0.2 rem), 0.2 mSv (20 rem), and 0.2 mSv (20 rem),

respectively.

(It should be noted that since these doses are less than 1×105

 mSv (<0.001 mrem), they would

have been reported as less than values” (i.e., less than 1×105

 mSv (<0.001 mrem)) if they had

been part of an assessment in this report. This is consistent with the notation that was used for

this report).

A.1.7.5 Radiation Dose From Crushing of Glass Tubes Containing Radioactive Gases

Table A.1.9 presents DSRs for crushing of glass tubes containing 3

H in the form of tritiated

water vapor (HTO) or the noble gases 85Kr and 220Rn. The 220Rn is assumed to come from the

radioactive decay of natural thorium in a glass tube such as a lamp or fluorescent lamp starter

(see Section 3.7). The external doses from air submersion in 85Kr were calculated using

Equations (1), (3), and (6), and the external doses from inhalation of 3

H in the form of HTO or

220Rn and its short-lived decay products were calculated using Equations (1), (3), (4), and (5). 

The exposure times needed to approach the maximum radiation doses are shorter in

enclosures with high ventilation rates than in enclosures with low ventilation rates. Hence, the

exposure times were varied in the calculations to give an indication of how long it would take to

approach the maximum radiation dose for the various types of enclosures considered in

Table A.1.9.

As an example of the use of the DSRs in Table A.1.9, consider a night sight for a gun with three

gaseous tritium light sources (GTLSs) containing approximately 66.6 MBq (18 mCi) of 3

H each

(see Section 2.14). The tritium in the GTLSs is assumed to be 99% elemental 3

H and only 1%

HTO, so that the amount of HTO in the GTLS is 6.66 MBq (180 Ci). If one of the GTLSs is

crushed in a home with a volume of 450 m3

 and a ventilation rate of 1 volume change per hour,

then the committed EDE to an individual in the house from inhalation of HTO over the next

8 hours would be 2.6×107

 rem/Ci times 180 Ci, or approximately 5×104 mSv (0.05 mrem). 

The elemental 3

H can be ignored because it contributes insignificantly to the radiation dose

received by the individual (see ICRP 68, Table C.1).

A.1.8 Summary

A generic methodology was developed to estimate radiation doses from accidents involving

NRC-exempted products and materials. The methodology considers the following four accident

exposure scenarios: (1) fires involving the release of radioactive materials from all types of

exempted products, (2) mechanical resuspension of radioactive materials during cleanup after a

fire, (3) spills of radioactive materials in liquid or powder form, and (4) crushing of glass tubes

containing radioactive gases. For these selected accident exposure scenarios, DSRs are

provided that give the radiation dose per unit quantity of radioactive material at risk in an
A.1–10

accident. Because of the broad range of exemptions, it has been necessary to estimate

radiation doses on a case-by-case basis when the DSR methodology is not applicable, and

radiation doses have been estimated on a case-by-case basis in some assessments using

better available data for a specific accident exposure scenario and product of interest.
A.1–11

Table A.1.1 Factors for Estimating Release of Respirable Size Particles 

During a Spill or Fire

Type of Factor Value

Release factor for materials involved in a fire

 Gases

 Solids/powders/liquids

 Solids in protective devices

Release factor for spills of liquids or powders

100%

0.1%

0.01%

0.1%
A.1–12

Table A.1.2 Enclosure Volumes and Ventilation Rates

Type of Enclosure

Enclosure Volume

(m3

)

Ventilation Rate

(volume/h)

Warehousea

Storeroom or 

cargo-handling bayb

3,000

300

1

4

Residencec

Bedroomd

450

27

1

1

Watch repair shope

 Small

 Large

Laboratoryf

18

34

180

1

1

6

Tractor trailerg

Large delivery truckh

Small delivery trucki

Automobilej

87

25

7.1

6.2

1

3

5

5

a

 See NUREG/CR–1775 and O’Donnell et al. (1981). Volume corresponds to a warehouse

measuring approximately 30.5 m × 30.5 m × 3.66 m.

b

 See IAEA SS No. 7. Volume corresponds to a storeroom or cargo-handling bay measuring

10 m × 10 m × 3 m.

c

 See O’Donnell et al. (1981). Volume corresponds to a residence with a living area of 186 m2

and a ceiling height of 2.44 meters.

d See NUREG/CR–0216. Volume corresponds to a room measuring approximately

3.66 m × 3.05 m × 2.44 m.

e

 See NUREG/CR–0215, NUREG/CR–0216. These reports also provide estimates of the

enclosure volumes for a jewelry store, catalog store, department store, etc.

f

 The volume corresponds to a room with a floor area of 60.4 m2 and a ceiling height of

3.05 meters.

g

 See Etnier and O’Donnell (1979). Cargo area in trailer is assumed to be 13.7 meters long,

2.35 meters wide, and 2.7 meters high. Volume is approximately 87 m3

 and ventilation rate is

likely very low in the trailer.

h

 See Etnier and O’Donnell (1979). Cargo area of truck is assumed to be 5 meters long,

2.2 meters wide, and 2.3 meters high. Volume is approximately 25 m3

 and the ventilation rate

is assumed to be less than a small delivery truck or automobile.

i

 See Etnier and O’Donnell (1979). Cargo area of truck is assumed to be 3.05 meters long,

1.8 meters wide, and 1.3 meters high. Volume is approximately 7.1 m3

 and the ventilation rate

is assumed to be similar to that of an automobile.

j

 See Etnier and O’Donnell (1979). Passenger area is assumed to be 3.05 meters long,

1.7 meters wide, and 1.2 meters high. Volume is approximately 6.2 m3 and the ventilation rate

is 5 volume changes per hour.
A.1–13

Table A.1.3 Average Concentrations in Air and Inhalation Intakes of Radioactive

Material During the 30-Minute Period Following a Fire

Type of Fire

Ventilation Rate 

(k)

Average

Concentration 

(C)

Individual Intake 

(I)

Transportation firea

House fireb

Warehouse firec

4 h1

0.25 h1

0.5 h1

1 h1

2 h1

1 h1

2 h1

3 h1

4 h1

1.4×103 Q m3

2.1×103 Q m3

2.0×103 Q m3

1.7×103 Q m3

1.4×103 Q m3

2.6×104 Q m3

2.1×104 Q m3

1.7×104 Q m3

1.4×104 Q m3

8.6×107 Q

1.3×106 Q

1.2×106 Q

1.0×106 Q

8.4×107 Q

1.6×107 Q

1.3×107 Q

1.0×107 Q

8.4×108 Q

a

 Based on an enclosure volume of 300 m3, a breathing rate of 1.2 m3/h, and use of a suppliedair respirator with a protection factor of 1,000.

b

 Based on an enclosure volume of 450 m3, a breathing rate of 1.2 m3/h, and use of a suppliedair respirator with a protection factor of 1,000.

c

 Based on an enclosure volume of 3,000 m3, a breathing rate of 1.2 m3/h, and use of a

supplied-air respirator with a protection factor of 1,000.
A.1–14

Table A.1.4 Radiation Dose-to-Source Ratios for a Transportation

Accident Involving Fire

Dose-to-Source Ratios (DAR)

Radionuclide

Effective Dose

Equivalent for

 Inhalationa,b

Effective Dose

Equivalent for

Submersiona,b

Effective Dose

Equivalent for

Resuspensiona

3

H (vapor)c

14C

46Sc

55Fe

60Co

63Ni

85Kr (gas)

90Sr + 90Y

106Ru + 106Rh

109Cd + 109mAg

137Cs + 137mBa

147Pm

152Eu

204Tl

210mBi + 206Tl

210Po

241Am

Th (natural)d

U (natural)e

U (depleted)f

8.3×1011 rem/Ci

1.8×1012 rem/Ci

2.6×1011 rem/Ci

2.3×1012 rem/Ci

1.9×1010 rem/Ci

2.7×1012 rem/Ci

1.1×109

 rem/Ci

4.1×1010 rem/Ci

9.5×1011 rem/Ci

2.8×1011 rem/Ci

3.4×1011 rem/Ci

1.9×1010 rem/Ci

2.1×1012 rem/Ci

6.5×1010 rem/Ci

8.1×109

 rem/Ci

3.8×107

 rem/Ci

1.8×1010 rem/mg 

7.5×1011 rem/mg 

3.8×1011 rem/mg 

2.5×1014 rem/Ci

9.3×1010 rem/Ci

1.2×109

 rem/Ci

2.4×109

 rem/Ci

8.8×1012 rem/Ci

1.1×1010 rem/Ci

4.5×1012 rem/Ci

2.5×1010 rem/Ci

8.4×1014 rem/Ci

5.3×1010 rem/Ci

1.7×1012 rem/Ci

1.2×1010 rem/Ci

3.9×1015 rem/Ci

7.6×1012 rem/Ci

7.2×1011 rem/mg 

6.8×1015 rem/mg 

6.1×1015 rem/mg 

2.0×1011 rem/Ci

2.9×1010 rem/Ci

2.6×1011 rem/Ci

2.1×109

 rem/Ci

2.9×1011 rem/Ci

1.3×108

 rem/Ci

4.6×109

 rem/Ci

1.1×109

 rem/Ci

3.1×1010 rem/Ci

3.7×1010 rem/Ci

2.1×109

 rem/Ci

2.3×1011 rem/Ci

7.3×108

 rem/Ci

9.0×108

 rem/Ci

4.2×106

 rem/Ci

1.9×109

 rem/mg 

8.3×1010 rem/mg 

4.2×1010 rem/mg 

See following page for footnotes.
A.1–15

Footnotes to Table A.1.4

a

 1 rem/Ci = 0.27 Sv/Bq; 1 rem/mg = 0.01 Sv/mg.

b

 Values should be reduced by a factor of 10 for fires involving solid radioactive materials

contained in protective devices (e.g., 241Am in smoke detectors). c

 Values apply to tritiated water and are increased by a factor of 1.5 to account for absorption

through the skin (ICRP 30).

d

 Values assume that natural thorium is 20 years old so that decay chain is approximately 87%

of equilibrium (see Table 3.1.5 of Section 3.1).

e

 Values assume that the mass abundances in natural uranium are 99.2745% by weight 238U,

0.720% by weight 235U, and 0.0055% by weight 234U (Parrington et al., 1996).

f

 Values assume that the mass abundances in depleted uranium are 99.7495% by weight 238U,

0.25% by weight 235U, and 0.0005% by weight 234U (Rich et al., 1988).
A.1–16

Table A.1.5 Radiation Dose-to-Source Ratios for a Warehouse Fire

Dose-to-Source Ratios (DSR)

Radionuclide

Effective Dose

Equivalent for

Inhalationa,b

Effective Dose

Equivalent for

Submersiona,b

Effective Dose

Equivalent for

Resuspensiona

3

H (vapor)c

14C

46Sc

55Fe

60Co

63Ni

85Kr (gas)

90Sr + 90Y

106Ru + 106Rh

109Cd + 109mAg

137Cs + 137mBa

147Pm

152Eu

204Tl

210mBi + 206Tl

210Po

241Am

Th (natural)d

U (natural)e

U (depleted)f

1.5×1011 rem/Ci

3.3×1013 rem/Ci

4.8×1012 rem/Ci

4.3×1013 rem/Ci

3.5×1011 rem/Ci

5.0×1013 rem/Ci

2.1×1010 rem/Ci

7.6×1011 rem/Ci

1.8×1011 rem/Ci

5.1×1012 rem/Ci

6.2×1012 rem/Ci

3.5×1011 rem/Ci

3.8×1013 rem/Ci

1.2×109

 rem/Ci

1.5×109

 rem/Ci

7.0×108

 rem/Ci

3.3×1011 rem/mg 

1.4×1011 rem/mg 

7.1×1012 rem/mg 

4.6×1015 rem/Ci

1.7×1010 rem/Ci

2.2×1010 rem/Ci

4.4×1010 rem/Ci

1.6×1012 rem/Ci

2.0×1011 rem/Ci

8.4×1013 rem/Ci

4.7×1011 rem/Ci

1.5×1014 rem/Ci

9.9×1011 rem/Ci

3.1×1013 rem/Ci

2.2×1011 rem/Ci

7.2×1016 rem/Ci

1.4×1012 rem/Ci

1.3×1011 rem/mg 

1.2×1015 rem/mg 

1.1×1015 rem/mg 

2.2×1012 rem/Ci

3.2×1011 rem/Ci

2.9×1012 rem/Ci

2.3×1010 rem/Ci

3.3×1012 rem/Ci

1.4×109

 rem/Ci

5.0×1010 rem/Ci

1.2×1010 rem/Ci

3.4×1011 rem/Ci

4.1×1011 rem/Ci

2.3×1010 rem/Ci

2.5×1012 rem/Ci

7.8×109

 rem/Ci

9.9×109

 rem/Ci

4.6×107

 rem/Ci

2.1×1010 rem/mg 

9.2×1011 rem/mg 

4.7×1011 rem/mg 

See following page for footnotes.
A.1–17

Footnotes to Table A.1.5

a

 1 rem/Ci = 0.27 Sv/Bq; 1 rem/mg = 0.01 Sv/mg.

b

 Values should be reduced by a factor of 10 for fires involving solid radioactive materials

contained in protective devices (e.g., 241Am in smoke detectors). c

 Values apply to tritiated water and are increased by a factor of 1.5 to account for absorption

through the skin (ICRP 30).

d

 Values assume that natural thorium is 20 years old so that the decay chain is approximately

87% of equilibrium (see Table 3.1.5 of Section 3.1).

e

 Values assume that the mass abundances in natural uranium are 99.2745% by weight 238U,

0.720% by weight 235U, and 0.0055% by weight 234U (Parrington et al., 1996).

f

 Values assume that the mass abundances in depleted uranium are 99.7495% by weight 238U,

0.25% by weight 235U, and 0.0005% by weight 234U (Rich et al., 1988).
A.1–18

Table A.1.6 Radiation Dose-to-Source Ratios for a Residential Fire

Dose-to-Source Ratios (DAR)

Radionuclide

Effective Dose

Equivalent for

Inhalationa,b,c

Effective Dose

Equivalent for

Submersiona,b,c

Effective Dose

Equivalent for

Resuspensiona

3

H (vapor)d

14C

46Sc

55Fe

60Co

63Ni

85Kr (gas)

90Sr + 90Y

106Ru + 106Rh

109Cd + 109mAg

137Cs + 137mBa

147Pm

152Eu

204Tl

210mBi + 206Tl

210Po

241Am

Th (natural)e

U (natural)f

U (depleted)g

1.0×1010 rem/Ci

2.1×1012 rem/Ci

3.0×1011 rem/Ci

2.7×1012 rem/Ci

2.2×1010 rem/Ci

3.1×1012 rem/Ci

1.3×109

 rem/Ci

4.8×1010 rem/Ci

1.1×1010 rem/Ci

3.2×1011 rem/Ci

3.9×1011 rem/Ci

2.2×1010 rem/Ci

2.4×1012 rem/Ci

8.0×109

 rem/Ci

9.4×109

 rem/Ci

4.4×107

 rem/Ci

2.2×1010 rem/mg 

8.7×1011 rem/mg 

4.4×1011 rem/mg 

3.1×1014 rem/Ci

1.2×109

 rem/Ci

1.5×109

 rem/Ci

3.0×109

 rem/Ci

1.1×1011 rem/Ci

1.3×1010 rem/Ci

5.8×1012 rem/Ci

3.3×1010 rem/Ci

1.0×1013 rem/Ci

6.5×1010 rem/Ci

2.1×1012 rem/Ci

1.5×1010 rem/Ci

5.0×1015 rem/Ci

9.8×1012 rem/Ci

1.5×1013 rem/mg 

8.3×1015 rem/mg 

7.4×1015 rem/mg 

1.1×1011 rem/Ci

1.6×1010 rem/Ci

1.4×1011 rem/Ci

1.1×109

 rem/Ci

1.6×1011 rem/Ci

6.7×109

 rem/Ci

2.5×109

 rem/Ci

5.7×1010 rem/Ci

1.7×1010 rem/Ci

2.0×1010 rem/Ci

1.2×109

 rem/Ci

1.2×1011 rem/Ci

3.9×108

 rem/Ci

4.9×108

 rem/Ci

2.3×106

 rem/Ci

1.1×109

 rem/mg 

4.5×1010 rem/mg 

2.3×1010 rem/mg 

See following page for footnotes.
A.1–19

Footnotes to Table A.1.6

a

 1 rem/Ci = 0.27 Sv/Bq; 1 rem/mg = 0.01 Sv/mg.

b

 Values should be reduced by a factor of 10 for fires involving solid radioactive materials

contained in protective devices (e.g., 241Am in smoke detectors). c

 Values apply to a firefighter. To estimate radiation doses to a person escaping from a fire or a

neighborhood hero trying to rescue a person from a fire, multiply DSRs for inhalation by 410

and DSRs for submersion by 0.39 (see Section A.1.7.2).

d

 Values apply to tritiated water and are increased by a factor of 1.5 to account for absorption

through the skin (ICRP 30).

e

 Values assume that natural thorium is 20 years old so that decay chain is approximately 87%

of equilibrium (see Table 3.1.5 of Section 3.1).

f

 Values assume that the mass abundances in natural uranium are 99.2745% by weight 238U,

0.720% by weight 235U, and 0.0055% by weight 234U (Parrington et al., 1996).

g

 Values assume that the mass abundances in depleted uranium are 99.7495% by weight 238U,

0.25% by weight 235U, and 0.0005% by weight 234U (Rich et al., 1988).
A.1–20

Table A.1.7 Average Concentrations in Air and Individual Intakes From Inhalation 

During Various Exposure Times Following a Spill in a Laboratory-Type Room a

Ventilation Rate 

(k)

Exposure Time 

(t)

Average Concentration 

(C)

Individual Intake 

(I)

6 h1

9 h1

12 h1

15 min

30 min

1 h

15 min

30 min

1 h

15 min

30 min

1 h

2.9×103 Q m3

1.8×103 Q m3

9.2×104 Q m3

2.2×103 Q m3

1.2×103 Q m3

6.2×103 Q m3

1.8×103 Q m3

9.2×104 Q m3

4.6×104 Q m3

8.6×104 Q

1.0×103 Q

1.1×103 Q

6.6×104 Q

7.3×104 Q

7.4×103 Q

5.3×104 Q

5.5×104 Q

5.5×104 Q

a

 Assumes an enclosure volume of 180 m3 and a breathing rate of 1.2 m3/h without a respirator.
A.1–21

Table A.1.8 Radiation Dose-to-Source Ratios for a Spill of a Liquid 

or Powder in a Laboratory-Type Room

Dose-to-Source Ratios (DSRs)

Radionuclide

Effective Dose Equivalent 

for Inhalationa

Effective Dose Equivalent

for Ingestiona

3

H (liquid)

14C

46Sc

55Fe

60Co

63Ni

90Sr + 90Y

106Ru + 106Rh

109Cd + 109mAg

137Cs + 137mBa

147Pm

152Eu

204Tl

210mBi + 206Tl

210Po

241Am

Th (natural)d

U (natural)e

U (depleted)f

9.6×1011 rem/Cib

2.1×109 rem/Ci

3.0×108

 rem/Ci

2.7×109

 rem/Ci

2.2×107

 rem/Ci

3.1×109

 rem/Ci

1.3×106

 rem/Ci

4.8×107

 rem/Ci

1.1×107

 rem/Ci

3.2×108

 rem/Ci

3.9×108

 rem/Ci

2.2×107

 rem/Ci

2.4×109

 rem/Ci

7.6×106

 rem/Ci

9.4×106

 rem/Ci

4.4×104

 rem/Ci

2.1×107

 rem/mg 

8.7×108

 rem/mg 

4.4×108

 rem/mg 

6.4×106

 rem/Cic

2.1×107 rem/Ci

6.4×107

 rem/Ci

6.1×108

 rem/Ci

1.0×106

 rem/Ci

5.8×108

 rem/Ci

1.4×105

 rem/Ci

2.7×106

 rem/Ci

1.3×106

 rem/Ci

5.0×106

 rem/Ci

1.1×107

 rem/Ci

6.5×107

 rem/Ci

3.4×107

 rem/Ci

9.6×106

 rem/Ci

1.9×104

 rem/Ci

3.6×104

 rem/Ci

5.1×108

 rem/mg 

1.9×108

 rem/mg 

1.0×108

 rem/mg 

See following page for footnotes.
A.1–22

Footnotes to Table A.1.8

a

 1 rem/Ci = 0.27 Sv/Bq; 1 rem/mg = 0.01 Sv/mg.

b

 Value applies to tritiated water and is increased by a factor of 1.5 to account for absorption

through the skin (ICRP 30).

c

 Value applies to intake of tritiated water through skin rather than direct ingestion of tritiated

water spilled on the skin (see Section A.1.6).

d

 Values assume that natural thorium is 20 years old so that decay chain is approximately 87%

of equilibrium (see Table 3.1.5 of Section 3.1).

e

 Values assume that mass abundances in natural uranium are 99.2745% by weight 238U,

0.720% by weight 235U, and 0.0055% by weight 234U (Parrington et al., 1996).

f

 Values assume that mass abundances in depleted uranium are 99.7495% by weight 238U,

0.25% by weight 235U, and 0.0005% by weight 234U (Rich et al., 1988).
A.1–23

Table A.1.9 Radiation Dose-to-Source Ratios for Crushing of Glass Tubes

Dose-to-Source Ratios (DSRs)

(rem/Ci)a

Enclosureb Exposure Time 3Hc 85Krd 220Rne

Warehouse

Storeroom or

cargo-handling bay

Residence

Bedroom

Large watch repair

 shop

Small watch repair

 shop

Laboratory

30 min

1 h

2 h

4 h

8 h

15 min

30 min

1 h

2 h

30 min

1 h

2 h

4 h

8 h

30 min

1 h

2 h

4 h

8 h

30 min

1 h

2 h

4 h

8 h

30 min

1 h

2 h

4 h

8 h

15 min

30 min

1 h

1.5×108

2.4×108

3.3×108

3.8×108

3.8×108

6.1×108

8.3×108

9.4×108

9.6×108

1.0×107

1.6×107

2.2×107

2.5×107

2.6×107

1.7×106

2.7×106

3.7×106

4.2×106

4.3×106

1.3×106

2.1×106

2.9×106

3.3×106

3.4×106

2.5×106

4.0×106

5.5×106

6.3×106

6.4×106

8.3×108

1.0×107

1.1×107

4.4×1010

7.1×1010

9.7×1010

1.1×109

1.1×109

1.8×109

2.4×109

2.7×109

2.8×109

2.9×109

4.7×109

6.4×109

7.3×109

7.5×109

4.9×108

7.8×108

1.1×107

1.2×107

1.2×107

3.9×108

6.2×108

8.5×108

9.7×108

9.9×108

7.3×108

1.2×107

1.6×107

1.8×107

1.9×107

2.4×109

2.9×109

3.1×109

3.4×108

5.4×108

7.3×108

8.2×108

8.3×108

1.4×107

1.9×107

2.1×107

2.2×107

2.3×107

3.6×107

4.9×107

5.4×107

5.5×107

3.8×106

6.0×106

8.1×106

9.1×106

9.2×106

3.0×106

4.8×106

6.5×106

7.2×106

7.3×106

5.7×106

9.1×106

1.2×105

1.4×105

1.4×105

1.9×107

2.3×107

2.4×107

See following page for footnotes.
A.1–24

Footnotes to Table A.1.9

a

 1 rem/Ci = 0.27 Sv/Bq.

b

 See Table A.1.2 for enclosure volumes and ventilation rates.

c

 Values for internal dose from inhalation and absorption of tritiated water vapor (HTO) through

the skin are based on an effective dose equivalent (EDE) of 9.6×105

 rem/Ci of HTO in air

(see Table 2.1.2 of Section 2.1, and EPA–520/1–88–020).

d

 Values for external dose from submersion in 85Kr are based on an EDE rate of 0.14 mSv/yr

(1.4×102

 rem/yr) plus 1% of the skin dose-equivalent rate of 1.5 rem/yr per Ci/m3 of 85Kr in air

(see Table 2.1.2 of Section 2.1, and EPA–402–R–93–081).

e

 Values for internal dose from inhalation of 220Rn and its progeny are based on an (1) EDE rate

of 1.6×103

 rem/yr per Ci/m3 of 220Rn in radioactive equilibrium with its progeny in indoor air and

(2) equilibrium factors calculated as ratios of time-averaged concentrations of 212Pb to 220Rn in

indoor air of various structures (see Section 3.1 and ICRP 50).
A.2–1

A.2 GENERIC DISPOSAL METHODOLOGY

A.2.1 Introduction

This appendix presents a generic methodology for estimating radiation doses from disposal of

exempted amounts of source or byproduct materials. Doses are estimated for the following

three disposal options: municipal landfills, municipal waste incinerators, and recycling in metals. 

For each disposal option, groups of exposed individuals and populations are defined, including

workers associated with operations at landfills, incinerators, and metal smelters and members

of the public who could be exposed in a variety of ways, depending upon the particular option. 

For disposal in landfills and incinerators, both individual and collective doses are calculated. 

However, because recycling of metals should be an unusual occurrence for most exempted

materials, only individual doses are estimated for this option.

For each group of exposed individuals and populations for an assumed disposal option,

assumed exposure pathways, which generally include external exposure, inhalation, and

ingestion, are defined. Then, for each exposure pathway, radionuclide-specific dose-to-source

ratios (DSRs) are calculated for the exposed individuals and populations. The DSRs give

effective dose equivalents (EDEs) per unit activity of radionuclides disposed for the assumed

disposal option and exposure pathway. Doses then are estimated from the relationship

H i j =DSR i j × A i , (1)

where H i j = EDE from exposure to radionuclide i for exposure pathway j ,

DSR = dose-to-source ratio for the particular radionuclide and exposure pathway, and

A i = assumed activity of the particular radionuclide disposed for the assumed

option. 

The DSRs are the quantities calculated in the generic disposal methodology. The DSRs then

are applied to assumed activities of radionuclides to estimate doses to individuals and

populations.

The following section describes the three disposal options assumed in developing the generic

disposal methodology, including the groups of exposed individuals and populations for each

option and the exposure pathways assumed for each group. The next three sections present

the models and parameter values for calculating the DSRs for each of the three disposal

options and the calculated DSRs for each radionuclide and exposure pathway. The last section

illustrates the application of the calculated DSRs to the estimation of individual and collective

doses from disposal of exempted amounts of radionuclides.

A.2.2 Description of Disposal Options

This section describes the exposure scenarios and associated exposure pathways for landfill

disposal, incineration, and metal recycling assumed in the generic disposal methodology for the

purpose of estimating doses to exposed individuals and populations.
A.2–2

A.2.2.1 Disposal in Municipal Landfills

More than half of all municipal solid waste generated in the United States is sent to landfills for

disposal (Environmental Protection Agency (EPA) 530–R–9–042). Therefore, except in unusual

cases in which particular items containing exempted amounts of radionuclides are not expected

to enter municipal waste streams, disposal in municipal landfills should be a common

occurrence.

For disposal of exempted amounts of radioactive materials in municipal landfills, the following

four groups of individuals are assumed to be exposed: (1) waste collectors, (2) workers at the

landfills, (3) off-site members of the public residing near the landfills, and (4) future on-site

residents at the landfills. The assumed exposure pathways for these groups are described in

the following paragraphs.

A.2.2.1.1 Waste Collectors

Waste collectors are individuals who collect waste from the generating site, haul the waste to

garbage trucks, and transport the waste to landfills. Exposure to waste collectors are assumed

to occur primarily during hauling of waste to garbage trucks in small containers. Exposure to

waste collectors during transport to landfills should be considerably less than during waste

collection because (1) the exposure time during transport should be much less than during

collection, (2) the distance between a waste collector and the sources should be greater during

transport than while hauling waste containers, and (3) garbage trucks should provide greater

shielding from external exposure than waste containers. Waste collectors are assumed to

receive exposures from the following three pathways: (1) external exposure to radionuclides in

the waste containers, (2) inhalation of radionuclides emitted from the waste containers into the

air, and (3) ingestion of radionuclides in the waste.

A.2.2.1.2 Landfill Workers

Workers at landfills are individuals who are located on top of the waste pile during operations

and who perform tasks such as dumping of waste, grading of waste following dumping, and

covering of the waste at periodic intervals. Exposure to other workers at a landfill should be

considerably less than exposure to workers on the waste pile, primarily because other workers

would be located at much larger distances from any sources. Workers at landfills are assumed

to receive exposures from the following three pathways: (1) external exposure to radionuclides

in the waste pile, (2) inhalation of radionuclides suspended from the waste pile into the air, and

(3) ingestion of radionuclides in the waste pile.

A.2.2.1.3 Off-Site Members of the Public

For off-site members of the public who reside near landfills, two different exposure scenarios

are considered. The first scenario, which would occur only during landfill operations, involves

releases of radionuclides into the air and subsequent atmospheric transport to off-site locations. 

For atmospheric releases during landfill operations, off-site residents are assumed to receive

exposures from the following four pathways: (1) inhalation of airborne radionuclides, (2)

external exposure to airborne radionuclides, (3) external exposure to radionuclides deposited

on the ground surface, and (4) ingestion of food products contaminated by deposition onto the

ground surface.
A.2–3

The second exposure scenario for off-site members of the public who reside near landfills,

which is assumed to occur only after landfills are closed, involves releases of radionuclides into

groundwater and subsequent transport to a nearby municipal well. Exposures for this scenario

would occur at times considerably later than the exposures from atmospheric releases during

landfill operations described above. Therefore, the exposed individuals and populations in the

two scenarios would not be the same. For releases to groundwater, off-site residents are

assumed to receive exposures from the pathway that involves ingestion of radionuclides in

drinking water obtained from a well. Based on a previous generic assessment for water

releases (Cook and Hunt, 1994), other potential exposure pathways are assumed to be

insignificant.

A.2.2.1.4 Future On-Site Residents

At some time after closure of a landfill, members of the public are assumed to establish

permanent residency on the landfill site. A suburban housing development is assumed, in

which no on-site sources of drinking water are established. Exposure to on-site residents are

assumed to result from the uncovering of waste during excavation at the site. Residents are

assumed to receive exposures from the following three pathways: (1) external exposure to

radionuclides in the waste during indoor and outdoor residence on the site, (2) inhalation of

radionuclides suspended from the waste into the air during indoor and outdoor residence on the

site, and (3) ingestion of radionuclides in the waste. Based on the assumption that a suburban

housing development is established at the landfill site, no other exposure pathways are

assumed to occur.

A.2.2.2 Disposal in Municipal Incinerators

A substantial fraction of all municipal solid waste generated in the United States is sent to

incinerators for disposal (EPA–530–R–94–042). Therefore, except in unusual cases in which

particular items containing exempted amounts of radionuclides are not expected to enter

municipal waste streams, disposal by incineration should be a common occurrence.

For disposal of exempted amounts of radioactive materials by incineration, the following three

groups of individuals are assumed to be exposed: (1) waste collectors, (2) workers at the

incinerators, and (3) off-site members of the public residing near the incinerators. The

assumed exposure pathways for these groups are described in the following paragraphs.

A.2.2.2.1 Waste Collectors

For waste collectors at incinerators, the assumed exposure scenario and exposure pathways

are the same as those described for waste collectors at landfills in Appendix A.2.2.1.1. 

Therefore, waste collectors are assumed to receive exposures from the following three

pathways: (1) external exposure to radionuclides in the waste containers, (2) inhalation of

radionuclides emitted from the waste containers into the air, and (3) ingestion of radionuclides

in the waste.

A.2.2.2.2 Incinerator Workers

Workers at incinerators are individuals who engage in sweeping or other cleanup activities

while located at the edge of a partially enclosed tipping area where garbage trucks unload
A.2–4

waste at the facility. Exposure to workers during waste unloading in the tipping area are

assumed to be substantially higher than exposure to workers during other operations at the

incinerators. Workers at incinerators are assumed to receive exposures from the following

three pathways: (1) external exposure to radionuclides in the waste pit, (2) inhalation of

radionuclides suspended from the waste pit into the air, and (3) ingestion of radionuclides in the

waste.

A.2.2.2.3 Off-Site Members of the Public

Off-site members of the public who reside near incinerators are assumed to receive exposures

from stack releases of radionuclides into the air following waste incineration and subsequent

atmospheric transport to off-site locations. The assumed exposure scenario and exposure

pathways for airborne releases from an incinerator are the same as those described for

airborne releases during landfill operations in Appendix A.2.2.1.3. Therefore, off-site residents

are assumed to receive exposures from the following four pathways: (1) inhalation of airborne

radionuclides, (2) external exposure to airborne radionuclides, (3) external exposure to

radionuclides deposited on the ground surface, and (4) ingestion of food products contaminated

by deposition onto the ground surface.

Off-site releases of airborne radionuclides from incineration facilities also could occur during

unloading of waste from garbage trucks into the tipping area. These releases would result in

the exposure pathways for off-site residents listed above. However, because of the general

concern about airborne releases of hazardous substances at waste incinerators, the partially

enclosed tipping area normally is ventilated so that a negative pressure, compared with the

outdoor air pressure, is maintained (Phone call, S. J. Levy, Office of Solid Waste, U.S.

Environmental Protection Agency, Washington, DC, February 1997). This precludes

substantial releases into the atmosphere during dumping operations and any other activities in

the tipping area. Therefore, releases during waste dumping should be unimportant compared

with stack releases during incineration.

Following incineration of waste, ash is removed for final disposal. However, doses from

disposal of incinerator ash are not considered in this assessment, primarily because incinerator

ash normally must be managed separately from municipal waste and would not be sent to

municipal landfills (EPA–530–R–94–042). Incinerator ash normally is managed as hazardous

waste due, for example, to the presumed presence of toxic heavy metals, and disposal in a

permitted facility for hazardous waste is required. Because of the stringent requirements for

treatment and disposal of hazardous waste under the Resource Conservation and Recovery

Act (RCRA) specified in 40 CFR 264, disposal of incinerator ash as hazardous waste should

result in doses substantially less than the doses from disposal of nonhazardous waste in

municipal landfills.

A.2.2.3 Metal Recycling

A substantial fraction of all municipal solid waste generated in the United States is recovered for

recycling (EPA–530–R–94–042). However, most of the recovered and recycled materials

include items such as paper and paper products, plastic, glassware, and aluminum and other

metal containers that would not contain radioactive materials, and recycling of most items

containing exempted amounts of radioactive materials is not expected to be a common

occurrence.
A.2–5

In this assessment, recovery and recycling of items containing exempted amounts of

radioactive materials are assumed to occur only for the purpose of recovering ferrous metals

and alloys (e.g., steel). This assessment is not concerned with recycling of exempted items

when the intent is to recover and reuse the radioactive material itself, because (1) this activity is

not a form of disposal, (2) it normally would not result in the introduction of radioactive material

into some other product as incidental contamination, and (3) it would be carried out by

licensees of the Nuclear Regulatory Commission (NRC) or an Agreement State.

For exempted items that could be assumed to be recycled for the purpose of recovering ferrous

metals, the items are assumed to be sent to a metal (i.e., steel) smelter. The following three

groups of individuals are assumed to be exposed: (1) workers at smelters, (2) off-site members

of the public residing near smelters, and (3) members of the public who use recycled products

containing radioactive material. Assumed exposure pathways for these groups are described in

the following paragraphs.

A.2.2.3.1 Smelter Workers

Based on a previous assessment (Hill et al., 1995), the individuals at metal smelters who are

assumed to receive the highest doses are slag workers. These workers are assumed to

receive exposures from the following three pathways: (1) external exposure to radionuclides in

slag, (2) inhalation of radionuclides emitted from slag into the air, and (3) ingestion of

radionuclides in slag.

A.2.2.3.2 Off-Site Members of the Public

Off-site members of the public who reside near smelters are assumed to receive exposures

from stack releases of radionuclides into the air following smelting and subsequent atmospheric

transport to off-site locations. The assumed exposure scenario and exposure pathways for

airborne releases from a smelter are the same as those described in Appendix A.2.2.1.3 for

airborne releases during landfill operations. Therefore, off-site residents are assumed to

receive exposures from the following four pathways: (1) inhalation of airborne radionuclides,

(2) external exposure to airborne radionuclides, (3) external exposure to radionuclides

deposited on the ground surface, and (4) ingestion of food products contaminated by deposition

onto the ground surface.

A.2.2.3.3 Users of Recycled Products

Members of the public are assumed to receive exposures during use of products containing

recycled metal. During use of contaminated products, members of the public are assumed to

receive exposures from the pathway that involves external exposure to radionuclides in the

product. Inhalation and ingestion of radionuclides in recycled metal products would not

normally occur (Hill et al., 1995) and, thus, is not considered in this assessment.

A.2.3 Dose Assessment for Disposal in Landfills

This section presents the models and parameter values for estimating individual and collective

doses from disposal of radioactive material in municipal landfills, and the results of the dose

assessment in the form of doses per unit activity of radionuclides disposed in all landfills
A.2–6

(i.e., the DSRs in Equation (1) are tabulated). The groups of individuals considered in the dose

assessment and their associated exposure pathways are described in Appendix A.2.2.1.

A.2.3.1 Waste Collectors

Waste collectors at municipal landfills are assumed to receive external, inhalation, and

ingestion exposures while hauling waste in containers from collection sites to a garbage truck. 

The dose assessment for waste collectors is described in the following paragraphs.

A.2.3.1.1 External Exposure to Individuals

The EDE to individual waste collectors from external exposure is calculated by assuming that a

unit activity of each photon-emitting radionuclide is uniformly distributed in a waste container,

which is assumed to be a cylinder with a height of 0.9 meter and a radius of 0.38 meter. The

uncompacted waste is assumed to have an average density over the container volume of

0.4 g/cm3

, and the self-shielding provided by the waste is taken into account by assuming that

the waste resembles water in its shielding properties. The shielding provided by the walls of the

container is taken into account by assuming that the wall thickness is 0.32 cm and that the wall

material, which normally is plastic, also resembles water in its shielding properties. Based on

these assumptions, the EDE rate near a waste container for a unit activity of 1 microcurie (Ci)

(37 kilobecquerel (kBq)) of each radionuclide in the waste was calculated using MicroShield

(Computer Codes, Grove Engineering, 1996).

In estimating the annual EDE to an individual waste collector, the following exposure conditions

were assumed. First, the collector was assumed to be located at a distance of 0.3 meter from

the surface of the waste container. Second, exposure to a single waste container was

assumed to occur for 4 hours, based on information that waste collectors typically work this

long before unloading the contents of a garbage truck (Phone call, J. Bailey, Browning-Ferris

Industries, Knoxville, TN, July 1994). Thus, the annual individual dose was calculated by

assuming that all of the annual disposals of radionuclides in landfills occur during a single waste

collection. Finally, in calculating the dose to a waste collector at a single landfill, the annual

disposal of a unit activity of each radionuclide in all landfills was assumed to be distributed

equally among 3500 operating landfills (EPA–530–R–96–006).

A.2.3.1.2 Inhalation Exposure to Individuals

The annual EDE to an individual waste collector from inhalation exposure is estimated by

assuming that the radioactive material is in a readily dispersible form and could be released into

the air during waste collection. The annual individual dose, H i , from an annual disposal of a

unit activity, A i , of 1 Ci (37 kBq) of each radionuclide i in all landfills is given by

H i /A i (rem/Ci) = (1/N L ) × (1/M c ) × L a × f a × U a × T × D inh, i , (2)

where N L = number of operating landfills,

M c = mass of waste per waste container (g),

L a = atmospheric mass loading of waste emitted from waste container (g/m3

),

f a = respirable fraction of airborne material,
A.2–7

U a = breathing rate for waste collector (m3

/h),

T = exposure time for waste collector (h), and

D inh, i = inhalation dose coefficient for radionuclide i (rem/Ci).

As in the analysis for external exposure described in the previous section, the number of

operating landfills, N L , in which disposals of radionuclides are assumed to be distributed

equally is 3500. The annual inhalation dose to an individual waste collector can be calculated

by assuming that all exposures occur during a single collection trip containing 1 year’s disposals

of a radionuclide. The assumed values of the other parameters in Equation (2) are described

as follows:

 The mass of waste per container, M c, was 1.4×105

 g, based on the assumed waste

density of 0.4 g/cm3

 and the dimensions of a waste container given in

Appendix A.2.3.1.1.

 The atmospheric mass loading of waste, L a, was 4×105

 g/m3, as described below.

 The respirable fraction of the airborne material, f a , was 0.7 (EPA, RAE–9232/1–2).

 The breathing rate, U a , was 1.2 m3

/h, which is a value appropriate for light activity (see

Appendix A.1).

 The exposure time for a single collection trip, T, was 4 hours (see Appendix A.2.3.1.1).

 The inhalation dose coefficient (D inh ) for each radionuclide was obtained from current

Federal guidance (EPA–520/1–88–020).

 For tritium (3

H) the inhalation dose factor has been increased by a factor of 1.5 to

account for absorption through the skin.

No data are available for estimating the atmospheric mass loading of waste emitted from a

container during waste collection. The value assumed in this assessment corresponds to an

average background dust loading (Anspaugh et al., 1975). The release of readily dispersible

material to the atmosphere during waste collection could be greater than the average release of

naturally occurring material on the ground surface. However, much of the waste could be

contained, for example, in plastic bags, and releases from a small source should be dispersed

away from the waste collector by prevailing winds for some fraction of the time, even when the

collector is close to the container.

A.2.3.1.3 Ingestion Exposure to Individuals

The annual EDE to an individual waste collector from ingestion exposure is estimated by

assuming that the radioactive material is in a readily dispersible form and could be transferred

to the hands of the individual during waste collection. The annual individual dose, H i , from an

annual disposal of a unit activity, A i , of 1 Ci (37 kBq) of each radionuclide i in all landfills is

given by

H i /A i (rem/Ci) = (1/N L ) × (1/M c ) × U ing × T × D ing, i , (3)
A.2–8

where the factor 1/N L again represents the assumption that the annual disposals of

radionuclides are distributed equally among all operating landfills, the factors M c and T are the

mass of waste per waste container and exposure time for the waste collector described with

Equation (2) in the previous section, and the other factors are described as follows:

U ing = ingestion rate of waste for waste collector (g/h), and

D ing, i = ingestion dose coefficient for radionuclide i (rem/Ci).

As in the analyses for external and inhalation exposure described previously, the annual

ingestion dose to a waste collector can be calculated by assuming that all exposures occur

during a single collection trip containing 1 year’s disposals of a radionuclide. The assumed

values of the parameters in Equation (3) that were not presented with Equation (2) are the

ingestion rate of waste, U ing (6×103

 g/h), which is a value appropriate for commercial or

industrial activities (EPA, OSWER Directive 9285.6–03), and the ingestion dose coefficient

(D ing ) for each radionuclide which was obtained from current Federal guidance

(EPA–520/1–88–020).

A.2.3.1.4 Collective Dose for Waste Collectors

The calculations of annual individual dose described in the previous three sections assume that

only one waste collector at each landfill collects waste containing radioactive material. This

assumption would overestimate the dose to an average waste collector when more than one

collector is involved at a landfill and the radioactive materials are distributed randomly in all

waste collections. However, given that a typical (i.e., median) landfill receives about 2.5×106

 kg

of waste per year (EPA/530–SW88–034) and that a normal garbage truck with a capacity of

about 20 m3

 (Phone call, J. Bailey, Browning-Ferris Industries, Knoxville, TN, July 1994) can

haul about 2×104

 kg of waste per shipment, assuming a density of compacted waste of about

1 g/cm3

, a typical landfill evidently could be serviced by very few trucks and, thus, very few

waste collectors. Therefore, it is not unreasonably pessimistic to assume only a single exposed

waste collector per typical landfill, even though this would not be the case at unusually large

landfills. Based on the assumption that only a single waste collector is exposed per landfill, the

annual collective dose to waste collectors from 1 year’s disposals of a radionuclide in all landfills

is obtained by multiplying the annual individual dose by the total number of operating landfills,

which again is assumed to be 3500 (EPA–530–R–96–006).

A.2.3.1.5 Results of Dose Calculations

The annual individual and collective EDEs to waste collectors at municipal landfills from 1 year’s

disposals of a unit quantity of 1 Ci (37 kBq) of each radionuclide in all landfills estimated in this

assessment (i.e., the DSRs in Equation (1)) are presented in Tables A.2.1 and A.2.2. In

applying the DSRs to disposal of particular items containing radioactive material, the following

points should be noted.

First, if a particular item could be sent to either landfills or incinerators for disposal, which is

often assumed to be the case, the DSRs should be reduced by the fraction of the total annual

disposals sent to landfills. Recent data indicate that the amount of waste sent to landfills is

about four times the amount sent to incinerators (EPA–530–R–94–042). Therefore, if the input

to the dose assessment is an assumed total activity of a radionuclide disposed per year in all
A.2–9

landfills and incinerators, the DSRs for all exposure pathways should be reduced by a factor of

0.8 to take into account the fraction of the disposed activity sent to landfills.

Second, the DSRs for inhalation and ingestion assume that the radioactive materials are in a

readily dispersible form, but this would not be the case for many items. Therefore, depending

on the physical form of the particular items of concern, the assessor could reduce the DSRs for

inhalation and ingestion whenever the radioactive materials should be significantly less

dispersible than loose materials in trash. The following guidelines for reducing the DSRs for

inhalation and ingestion in these cases are suggested:

 For some exempt items, distribution and use may be limited, and it may not be

conservative to assume that the annual disposal occurs uniformly over all 3500 landfills. 

For this situation, an adjustment should be applied to reflect localized use and disposal. 

As a simple adjustment, if the assumed number of items to be disposed of annually is

less than 3500, which is the assumed number of disposal sites, the DSRs should be

increased by the ratio of 3500 divided by the number of items.

 For radioactive materials that should be considerably less dispersible than loose

materials in trash, but nonetheless could be dispersed to some extent, the DSRs for

inhalation and ingestion may be reduced by a factor of 10. Examples of this case might

include radioactive materials in the form of plated foils or solid items that are easily

breakable or crushable into small pieces.

 For items that should be nondispersible during normal waste collection activities, the

DSRs for inhalation and ingestion may be assumed to be zero (0). Examples of this

case might include large solid metal forms or radioactive materials dispersed in

substantial glass or ceramic forms.

If no correction for dispersibility is applied, then, for any radionuclide, the DSR for either

individual or collective dose from all exposure pathways is the sum of the DSRs for external

exposure, inhalation, and ingestion.

A.2.3.2 Workers at Landfill

Workers at municipal landfills are assumed to receive external, inhalation, and ingestion

exposures while located on top of a waste pile. The dose assessment for landfill workers is

described in the following paragraphs.

A.2.3.2.1 External Exposure to Individuals

The annual EDE to an individual landfill worker from external exposure is estimated by

assuming that the source is an infinitely thick, uniformly contaminated volume of soil-equivalent

material and that the worker is operating heavy equipment on top of exposed waste at the

working face of the landfill. The annual individual dose, H i , from an annual disposal of a unit

activity, A i , of 1 Ci (37 kBq) of each photon-emitting radionuclide i in all landfills is given by

H i /A i (rem/Ci) = (1/N L ) × (1/M w ) × f ex × f sh × D ext, i , (4)
A.2–10

where the factor 1/N L represents the assumption described in Appendix A.2.3.1.1 that the

annual disposals of radionuclides are distributed equally among all operating landfills, and the

other factors are described as follows:

M w = mass of waste disposed in landfill annually (g/yr),

f ex = fraction of the year during which exposure occurs,

f sh = shielding factor while operating heavy equipment, and

D ext, i = external dose coefficient for radionuclide i (rem/yr per Ci/g).

The assumed values of the parameters in Equation (4), except for N L, which again is assumed

to be 3500, are described as follows:

 The mass of waste disposed in a landfill annually, M w , was 2.5×109

 g, as described

below.

 The fraction of the year during which exposure occurs, f ex , was 0.18, based on an

assumed time spent working on the waste pile of 1600 h/yr.

 The shielding factor while operating heavy equipment, f sh , was 0.75.

 The external dose coefficient (D ext ) for each radionuclide for an infinitely thick volume

source in soil was obtained from current Federal guidance (EPA 402–R–93–081).

The mass of waste disposed in a landfill was assumed to be the median value for all landfills

(EPA–520/1–88–020). The median value was used, rather than the average, because the

distribution of disposals in landfills is highly skewed and about 84% of all landfills receive less

than the average amount of waste (EPA–520/1–88–020). Use of the median value also gives

higher estimates of dose.

For 85Kr, external exposure was estimated by assuming that half of the activity would be

released into the air during landfill operations, due to breakage of half of the containers for this

radionuclide, and half would be retained in the waste in intact containers. Therefore, the

external dose from 85Kr retained in the waste would be half of the value calculated as described

above. The 85Kr released into the air also would result in external exposure to workers from

submersion in an atmospheric cloud. However, if the release occurs over the working face of

the landfill with an assumed area of 2100 m2

 (i.e., an area of about 46 m × 46 m), the mixing

height for the release is assumed to be 10 meters, and the average wind speed is assumed to

be 2 m/s, the exposure time for any release would be less than 25 seconds. Using the dose

coefficient for air submersion from current Federal guidance (EPA–402–R–93–081), the

external dose from submersion in the atmospheric cloud would be more than three orders of

magnitude lower than the dose from external exposure to 85Kr remaining in the waste pile. 

Therefore, external exposure to landfill workers to 85Kr released from the waste pile can be

neglected.

A.2.3.2.2 Inhalation Exposure to Individuals

The annual EDE to an individual landfill worker from inhalation exposure is estimated using a

model of the form given by Equation (2) in Appendix A.2.3.1.2. The mass of waste disposed

annually in a landfill, M w , and the exposure time, T, were the values given above for external
A.2–11

exposure. As described with Equation (2), the respirable fraction of airborne material, f a , was

0.7; the breathing rate, U a , was 1.2 m3

/h; and the inhalation dose coefficients for radionuclides,

D inh , were obtained from current Federal guidance. The assumed atmospheric mass loading

of waste, L a , was 2×104

 g/m3, which is a recommended value for the dust loading due to

mechanical disturbance (Healy, 1980).

A.2.3.2.3 Ingestion Exposure to Individuals

The annual EDE to an individual landfill worker from ingestion exposure is estimated using a

model of the form given by Equation (3) in Appendix A.2.3.1.3. The mass of waste disposed

annually in a landfill, M w , and the exposure time, T, were the values given above for external

and inhalation exposure. The ingestion dose coefficients for radionuclides, D ing , were obtained

from current Federal guidance. The assumed ingestion rate of waste, U ing , was 0.06 g/h,

which is a value appropriate for construction activities (EPA, OSWER Directive 9285.6–03).

A.2.3.2.4 Collective Dose for Landfill Workers

Based on discussions with a landfill operator (Phone call, J. Bailey, Browning-Ferris Industries,

Knoxville, TN, July 1994), a total of five workers are assumed to be located at the open working

face of a landfill during the working year of 1600 hours. Therefore, for any exposure pathway,

the annual collective EDE to landfill workers is given by the annual individual dose multiplied by

the factor 5 × 3,500 = 17,500, where 3,500 again is the assumed number of operating landfills

(EPA–530–R–96–006).

A.2.3.2.5 Results of Dose Calculations

Tables A.2.3 and A.2.4 present the annual individual and collective EDEs to workers at

municipal landfills from 1 year’s disposals of a unit quantity of 1 Ci (37 kBq) of each

radionuclide in all landfills estimated in this assessment (i.e., the DSRs in Equation (1)). In

applying the DSRs to disposal of particular items containing radioactive material, the three

corrections described in Appendix A.2.3.1.5 normally could be applied. The first is a reduction

in all DSRs by a factor of 0.8 to take into account the fraction of all disposals sent to landfills

when disposal by incineration also could occur. The second is a reduction in the DSRs for

inhalation and ingestion that could be applied when the radioactive materials should be

significantly less dispersible than loose materials in the waste. The correction factors for

dispersibility discussed in Appendix A.2.3.1.5 for inhalation and ingestion exposure to waste

collectors should be applicable to exposure to landfill workers. The third correction addresses

the situation where there is limited distribution or use of the exempt material. As discussed in

Section A.2.3.1.5, an adjustment should be applied to reflect localized use and disposal. If the

assumed number of items to be disposed of annually is less than 3500, which is the assumed

number of disposal sites, the DSRs should be increased by the ratio of 3500 divided by the

number of items.

A.2.3.3 Off-Site Members of the Public During Landfill Operations

During operations at municipal landfills, off-site individuals and populations are assumed to be

exposed to radionuclides released into the air and transported to off-site locations. The

following four exposure pathways are assumed to occur: (1) inhalation of airborne

radionuclides, (2) external exposure to airborne radionuclides, (3) external exposure to
A.2–12

radionuclides deposited on the ground surface, and (4) ingestion of food products contaminated

by deposition onto the ground surface.

For radionuclides in particulate form, the fraction of the material disposed released into the air

during landfill operations is assumed to be 1×103

. This release fraction is obtained from the

following factors. First, the emission rate of particulate material during dumping and grading

operations is assumed to be 4.3 kg/h (EPA–450/1–89–003). Second, this emission rate is

reduced by a factor of 3 to account for the normal wetting of waste at a landfill to control

airborne dust levels. Third, the emissions are assumed to occur for 2000 h/yr of operations. 

Finally, as described in Appendix A.2.3.2.1, the mass of waste disposed in a landfill annually is

2.5×109

 g.

For 3

H, which would not be in particulate form but is assumed to be present in soil water, the

amount of disposed material that becomes airborne per year is estimated by multiplying the

amount of 3

H per unit volume of waste after disposal by an assumed evapotranspiration rate of

0.4 m/yr (Computer Codes, Yu et al., 1993) and by the assumed 2100 m2

 area of the working

face of the landfill from which evapotranspiration occurs. As discussed in Appendix A.2.3.2.1,

half of the amount of 85Kr disposed per year is assumed to become airborne during landfill

operations.

For the assumed releases of radionuclides into the air described above, annual individual and

collective doses to nearby residents were calculated using CAP–88 (Computer Codes, Beres,

1990). Assumptions used in the calculations are described as follows:

 The releases occur at ground level and uniformly over an assumed area of the working

face at a landfill of 2100 m2

. The meteorological data (i.e., the distribution of wind speed

and direction, annual rainfall, and average temperature) used in the calculations were

for a site in Oak Ridge, TN.

 The rural agricultural data set contained in CAP–88 (Computer Codes, Beres, 1990)

was used to evaluate the dose from ingestion of contaminated food products. For the

vegetable, milk, and beef pathways, this data set specifies (1) the fraction of the

ingested foodstuff produced at home, produced in the remainder of the assessment

area, or imported, and (2) the density of beef and milk cattle and the fraction of the land

surface cultivated for vegetable crops.

 In calculating collective dose, a population of 500,000 was assumed to be uniformly

distributed within a distance of 80 km of each landfill. The assumed population was

based on the average population density in the United States (Bureau of Census, 1990),

and is intended to represent the variety of rural and semi-urban locations of landfills.

Table A.2.5 presents the annual individual and collective EDEs to off-site residents due to

airborne releases during operations at municipal landfills from 1 year’s disposals of a unit

quantity of 1 Ci (37 kBq) of each radionuclide in all landfills estimated in this assessment (i.e.,

the DSRs in Equation (1)). In applying the DSRs to disposal of particular items containing

radioactive material, the three corrections described in Appendix A.2.3.1.5 normally could be

applied. The first is a reduction in the DSRs by a factor of 0.8 to take into account the fraction

of all disposals sent to landfills when disposal by incineration also could occur. The second is a

reduction in the DSRs that could be applied when the radioactive materials should be
A.2–13

significantly less dispersible than loose materials in the waste. The latter reduction would be

applied to all exposure pathways, because all of the pathways result from airborne releases. 

The correction factors for dispersibility discussed in Appendix A.2.3.1.5 for inhalation and

ingestion exposure to waste collectors should be applicable to exposure to off-site residents.

The third correction addresses the situation in which there is limited distribution or use of the

exempt material. As discussed in Section A.2.3.1.5, an adjustment should be applied to reflect

localized use and disposal. If the assumed number of items to be disposed of annually is less

than 3500, which is the assumed number of disposal sites, the DSRs should be increased by

the ratio of 3500 divided by the number of items.

A.2.3.4 Off-Site Members of the Public Following Landfill Closure

Following closure of a landfill, off-site individuals and populations are assumed to be exposed to

radionuclides released into groundwater and transported to a nearby municipal well. Ingestion

of drinking water obtained from the well is the only exposure pathway considered.

For this scenario, the following distinction is made in estimating individual and collective doses. 

The estimated individual dose is the highest dose that would occur in any year from all

disposals over the operating lifetime of a landfill, and the intent is to estimate the individual dose

from actual disposal practices at a landfill. However, the estimated collective dose is the dose

that would occur in an exposed population, over an assumed period of time after disposal, from

only 1 year’s disposals. The intent is to estimate the collective dose from the disposal of

1 year’s distribution of exempted items, based on an assumption that the number of items

disposed per year would equal the annual distribution, for comparison with the annual collective

dose during distribution and transport and routine use. The assessments of individual and

collective dose for this scenario are described in the following paragraphs.

A.2.3.4.1 Individual Dose From Well Water Use

The annual EDE to an off-site individual who ingests water obtained from a well located near

the landfill is estimated using a simple model for release of radionuclides into groundwater and

transport to the well. Conceptually, a first-order leaching model is used to estimate the annual

release of radionuclides into groundwater. The resulting concentration of radionuclides in

groundwater is estimated by diluting the annual release in an assumed annual volume of

groundwater flowing underneath the landfill, and the radionuclides are assumed to be

transported in groundwater to the well without dispersion or further dilution. Therefore, the

concentrations of radionuclides at the well are reduced relative to the concentrations in

groundwater beneath the landfill only by radioactive decay during the travel time from the

landfill to the well.

Based on the simple conceptual model for release from a landfill and transport to a well

described above, the annual EDE to an individual, H i , from consumption of radionuclide i in

drinking water can be written as

H i (rem/yr) = (A T, i × λ L, i /q w) × U w × D ing, i × exp(λ R, i t i ) , (5)
A.2–14

where A T, i = total activity of radionuclide i in landfill at time leaching begins (Ci),

λ L, i = leaching constant from landfill into groundwater for radionuclide i (1/yr),

q w = annual dilution volume of water beneath landfill (m3

/yr),

U w = ingestion rate of water from well by individual (m3

/yr),

D ing, i = ingestion dose coefficient for radionuclide i (rem/Ci),

λ R, i = decay constant for radionuclide i (1/yr), and

t i = travel time in groundwater from landfill to well for radionuclide i (yr).

Leaching of radionuclides into groundwater is assumed to begin when disposal operations at

the landfill cease, which maximizes the activity of radionuclides available for leaching. Then, if

A i is the annual disposal of a unit activity of 1 Ci (37 kBq) of radionuclide i in all landfills, the

total activity of radionuclide i in a single landfill at the time leaching begins is given by

A T, i = (1/N L ) × (A i /λ R, i ) [1 exp(λ R, i T L )] , (6)

where the factor 1/N L again represents the assumption described in Appendix A.2.3.1.1 that the

annual disposals of radionuclides are distributed equally among all operating landfills, and the

rest of this equation gives the activity at the end of the operating lifetime of the landfill, T L ,

taking into account the annual disposals, A i , and radioactive decay during the operating

period. By combining Equations (5) and (6), the annual individual EDE from an annual disposal

of the unit activity of radionuclide i is given by

H i /A i (rem/Ci) = (1/N L) × (1/λ R, i ) [1 exp(λ R, iT L)] × (λ L, i /q w) × U w (7)

× D ing, i × exp(λ R, i t i ) .

The leaching constant, λ L, i , and the travel time in groundwater from the landfill to the well, t i ,

for radionuclide i in Equation (7) are described in the following paragraphs.

The leaching constant for radionuclide i, λ L, i , is based on the model of Baes and Sharp (1983)

for a saturated medium, corrected for leaching in an unsaturated medium (NUREG/CR–4370;

Computer Codes, Rogers and Hung, 1987). The leaching model for a saturated medium has

previously been used, for example, in a generic assessment for releases from contaminated

soil (NUREG/CR–5512). The leaching constant is written as

λ L, i (1/yr) = (I/θd w R i ) × (I/K s ) , (8)

where I = infiltration rate of water through landfill (m/yr),

θ = volumetric water content of material in landfill (dimensionless),

d w = thickness of waste in landfill (m),

R i = retardation factor for transport of radionuclide i in water, and

K s = saturated hydraulic conductivity of material in landfill (m/yr).
A.2–15

The factor I/θd w R i is the leaching constant for a saturated medium (Baes and Sharp, 1983),

and the factor I/K s , called the contact time fraction (NUREG/CR–4370), is the correction for

leaching in an unsaturated medium. The contact time fraction takes into account that leaching

of radionuclides in an unsaturated medium occurs only during the fraction of the time that water

is infiltrating through the medium. The retardation factor, R i , is given by (Baes and Sharp,

1983)

R i = 1 + ρK d, i /θ , (9)

where the parameter θ is listed with Equation (8) above and

ρ = bulk density of material in landfill (g/cm3

), and

K d, i = solid/solution distribution coefficient for radionuclide i (mL/g).

The travel time in groundwater from the landfill to the well, t i , for radionuclide i is given

by

t i = R i X/V w , (10)

where R i is the retardation factor for transport of radionuclide i in water in Equation (9) and

X = distance of groundwater flow from edge of landfill to well (m), and

V w = groundwater velocity (m/yr).

The model for estimating the annual individual dose from an annual disposal of a unit activity of

a radionuclide in all landfills is given by Equations (7) to (10). The assumed values of

parameters in Equation (7) that are not contained in other equations are described as follows:

 The operating lifetime of a landfill, T L , was 30 years, based on data for operating and

closed facilities (EPA/530–SW88–034).

 The annual dilution volume of water beneath the landfill, q w , was 7×104

 m3/yr, obtained

as described below.

 The ingestion rate of water from the well by an individual, U w , was 0.73 m3

/yr, based on

a consumption rate of drinking water of 2 L/day (EPA/600/P–95/002Fa).

 The ingestion dose coefficient, D ing , for each radionuclide was obtained from current

Federal guidance (EPA–520/1–88–020).

The dilution volume for radionuclides leached from the landfill into groundwater was assumed to

be the annual precipitation multiplied by the area of the landfill (EPA RAE–9232/1–2). As

described later in this section, the annual infiltration of water through the landfill was assumed

to be one-half of the annual precipitation. Therefore, the assumed dilution volume of water is

twice the volume of water infiltrating through the landfill. This assumption is intended to be
A.2–16

representative of sites where only local recharge of groundwater occurs and the distance from

the landfill to the well is comparable to or less than the dimension of the landfill parallel to the

groundwater flow path. The assumed dilution volume would be quite conservative at sites

where the amount of groundwater flowing beneath the landfill is much greater than the local

recharge. In this assessment, the assumed precipitation was 0.9 m/yr, which is representative

of sites with relatively abundant rainfall, and the assumed area of the landfill is 7.6×104

 m2,

based on a typical size of planned facilities (EPA/530–SW88–034). The product of the annual

precipitation and the area of the landfill gives the assumed dilution volume.

The assumed values of the parameters in Equations (8) to (10) are described as follows:

 The infiltration rate of water through the landfill, I, was 0.45 m/yr, based on an

assumption that infiltration is one-half of total precipitation (EPA RAE–9232/1–2), which

is appropriate for sites with relatively abundant rainfall, and an assumed annual

precipitation of 0.9 m/yr.

 The volumetric water content of material in the landfill, θ, was 0.3, which is a

representative upper bound for different types of soil (Baes and Sharp, 1983).

 The thickness of waste in the landfill, d w , was 10 meters, based on the assumed area

of the landfill given above and typical waste volumes for a landfill

(EPA/530–SW88–034).

 The bulk density, ρ, of material in the landfill was 1.4 g/cm3

 (Baes and Sharp, 1983).

 The solid and solution distribution coefficient, K d , for each radionuclide was the value

adopted in NUREG/CR–5512, and is given in Table A.2.6.

 The contact time fraction, I/K s , for waste in the unsaturated zone was 5×103

, as

described below.

 The distance of groundwater flow from the edge of the landfill to the well, X, was

100 meters, based on an assumption that the well is located at the boundary of the

buffer zone.

 The groundwater velocity, V w , was 10 m/yr, which is representative of sites with

relatively fast groundwater flow (NUREG/CR–4370).

The assumed contact time fraction for waste in the unsaturated zone was based on values

developed by the NRC for reference sites in different regions of the United States

(NUREG/CR–4370), and is intended to represent an average value at sites with relatively

abundant rainfall. For an assumed annual infiltration of water of 0.45 m/yr, the assumed

contact time fraction of 5×103

 corresponds to a saturated hydraulic conductivity, K s , in the

landfill of about 1×102

 m/yr. This is toward the low end of representative values for different soil

types (Computer Codes, Yu et al., 1993) and, thus, would result in somewhat conservative

values of the contact time fraction at many sites.

In estimating the annual individual dose using Equations (7) to (10), an additional constraint is

applied in regard to the time period of concern. In particular, an individual dose is calculated for
A.2–17

a radionuclide only if the travel time from the landfill to the well obtained from Equation (10) is

1000 years or less. The assumed time period for the calculations is based on the NRC’s stated

intention that its radiological criteria for decontamination and decommissioning of contaminated

sites would be applied only for 1000 years (NRC, 62 FR 39058). Therefore, for the assumed

distance from the landfill to the well of 100 meters and the assumed groundwater velocity of

10 m/yr, an individual dose is calculated for a radionuclide only if the retardation factor is 100 or

less (i.e., if the distribution coefficient in Table A.2.6 is about 21 mL/g or less).

A.2.3.4.2 Collective Dose from Well Water Use

As discussed at the beginning of Appendix A.2.3.4, the collective dose for releases from a

landfill to groundwater is calculated for 1 year’s disposal of a unit activity of each radionuclide,

rather than the annual disposals over the operating lifetime of the landfill. In this assessment,

the disposals are assumed to occur in the last year of operations, which maximizes the

collective dose. Therefore, based on Equations (5) and (7) for the annual individual dose from

disposals over the operating lifetime of landfills, the annual individual dose from 1 year’s

disposals used in the calculation of collective dose is given by

H i (rem/yr) = (1 Ci) × (1/N L ) × (λ L, i /q w) × U w × D ing, i × exp(λ R, i t i ) . (11)

The calculation of collective dose from use of well water near landfills is based on the annual

individual dose in Equation (11) and the following assumptions. First, the population served by

all municipal wells located near landfills is 700,000 (EPA/530–SW88–034); i.e., the average

population at each of the 3,500 landfills is 200. Second, the collective dose is calculated for the

time period between the arrival of the contaminant plume, as obtained from Equation (10), and

1000 years, which is the time period of concern for the calculations discussed in the previous

section. Thus, the collective dose over 1,000 years from 1 year’s disposals of a unit quantity of

a radionuclide in all landfills is obtained by integrating the collective dose for the first year of

exposure, as obtained from the annual individual dose in Equation (11) and the assumed

population of 700,000, from time t i in Equation (10) to 1,000 years, taking into account

radioactive decay over that time.

A.2.3.4.3 Results of Dose Calculations

For releases to groundwater and transport to a nearby well, the individual and collective EDEs

to off-site residents from disposals of a unit quantity of 1 Ci (37 kBq) of each radionuclide in all

landfills estimated in this assessment (i.e., the DSRs in Equation (1)) are presented in

Table A.2.7. As discussed previously, the individual doses represent the annual dose from

disposal of the unit quantity of each radionuclide during each year over the assumed 30-year

operating lifetime of landfills, but the collective doses represent the dose over 1000 years from

disposal of the unit quantity of each radionuclide during the last year of operations only.

In applying the DSRs to disposal of particular items containing radioactive material, the

following points should be noted. First, as in the other exposure scenarios for disposal in

landfills, all DSRs normally should be reduced by a factor of 0.8 to take into account the fraction

of all disposals sent to landfills when disposal by incineration also could occur.
A.2–18

Second, the DSRs for releases to groundwater assume the radioactive materials are dispersed

in the landfill in a form that would be readily accessible to infiltrating water, but this would not be

the case for many items. Therefore, depending on the physical form of the particular item of

concern, the assessor could reduce the DSRs whenever the radioactive materials should be

significantly less accessible to infiltrating water than finely dispersed materials. Based on the

accessibility index for disposal of low-level radioactive waste developed by the NRC

(NUREG–0782), the following guidelines for reducing the DSRs for releases to groundwater are

suggested:

 For materials in the form of small bulk solids that should be significantly less accessible

to infiltrating water than loose materials in waste or very small items, the DSRs may be

reduced by a factor of 10.

 For large solid items, such as solid metal forms, that should be accessible to infiltrating

water only at the surface of the waste form and should have a low leachability, the DSRs

may be reduced by a factor of 100.

The reduction factor for leaching of relatively inaccessible materials in water is similar in

concept to the reduction factor applied to releases of less dispersible materials during waste

collection and landfill operations that is discussed in Appendix A.2.3.1.5. However, during

waste collection and landfill operations, releases from some types of items may be assumed to

be zero (0), but an assumption of zero (0) release would not be reasonable for disposal,

because even large metal, glass, or ceramic waste forms would be subject to leaching at the

surface.

An additional correction is needed for addressing the situation in which there is limited

distribution or use of the exempt material. As discussed in Section A.2.3.1.5, an adjustment

should be applied to reflect localized use and disposal. If the assumed number of items to be

disposed of annually is less than 3500, which is the assumed number of disposal sites, the

DSRs should be increased by the ratio of 3500 divided by the number of items. 

A.2.3.5 Future On-Site Residents at Landfills

In the course of developing a model for exposure to future on-site residents at municipal landfill

sites for use in the generic disposal methodology in Appendix A.2, two issues required

consideration. The first was the types of exposure scenarios involving access to municipal

landfill sites by members of the public that would be reasonable to assume following closure of

a site and its release for public use. The second issue was the particular exposure pathways

that should be assumed for the chosen exposure scenario.

Based on available information, it appeared that the most common uses of municipal landfill

sites following closure and release to the public would be as golf courses, public parks, or other

recreational areas, or perhaps as an industrial park. This is reasonable when one considers

that municipal landfills now are constructed, operated, and closed under Subtitle D of the

Resource Conservation and Recovery Act (RCRA) in much the same way as hazardous waste

disposal facilities and it is considered desirable not to unduly disturb disposed waste after

closure. None of the expected uses of closed landfill sites involve permanent occupancy by

members of the public. However, construction of housing at the sites is a credible, albeit
A.2–19

somewhat unlikely, occurrence and permanent residence in housing should result in higher

doses to members of the public than the other credible uses noted above.

Since the assumption of permanent on-site residence in housing already is expected to be

conservative compared with more likely exposure scenarios at landfill sites, it was decided that

exposure pathways should be included in the scenario only if they would be reasonably likely to

occur. Based on this consideration, it was decided not to include a food ingestion (vegetable)

pathway in the dose assessment for future on-site residents, because it is obvious from

observation that most home owners do not have a vegetable garden, especially a garden at the

same location as their home. In contrast, external exposure and intakes by inhalation of

suspended activity or ingestion of waste materials either are unavoidable or are reasonably

likely occurrences at any site and for any living habits, and these scenarios were included in the

dose assessment for future on-site residents.

The decision not to include a vegetable pathway in the dose assessment for future on-site

residents is in accordance with recommendations of the International Commission on

Radiological Protection (ICRP) in regard to the definition of critical groups to be used in

radiation protection, as described, for example, in ICRP Publication 26 (see references). The

critical group essentially is the population group expected to receive the highest doses, but the

ICRP intends that doses should be assessed for average exposures within the critical group,

rather than the maximum possible exposure to any individual. Thus, in the case of interest

here, the critical population group consists of individuals who are assumed to reside on a

municipal landfill site, and average (i.e., expected) doses to these individuals should be

calculated. Since exposure to individuals due to consumption of vegetables grown on a landfill

site is not expected to be a normal occurrence, the ICRP does not intend that this pathway

should be included in a dose assessment for this critical group. Based on these considerations,

the following modeling assumptions were used for estimating doses for future on-site residents

at landfills. 

At some time following closure of a municipal landfill, members of the public are assumed to

establish permanent residency in a suburban housing development on the landfill site. The

following three exposure pathways are assumed to occur: (1) external exposure to

radionuclides in the landfill during indoor and outdoor residence on the site, (2) inhalation of

radionuclides suspended from the landfill into the air during indoor and outdoor residence on

the site, and (3) ingestion of radionuclides in the waste. The existence of these exposure

pathways is based on an assumption that waste in the landfill is uncovered during excavation of

the site and remains uncovered during site occupancy.

For this exposure scenario, the same distinction is made in estimating individual and collective

doses as is made in the scenario for releases to groundwater described at the beginning of

Appendix A.2.3.4. That is, the estimated individual dose is the highest dose that would occur in

any year, due to all disposals over the operating lifetime of a landfill. For any radionuclide, this

dose would occur at the time residence on the landfill site first occurs. However, the estimated

collective dose is the dose that would occur in an assumed population, over an assumed time

period after disposal, due only to 1 year’s disposals during the last year of landfill operations. 

The dose assessment for future on-site residents is described in the following paragraphs.
A.2–20

A.2.3.5.1 External Exposure to Individuals

The annual EDE to an individual on-site resident from external exposure is estimated by

assuming that the source is an infinitely thick, uniformly contaminated volume of soil. Using

Equation (4) in Appendix A.2.3.2.1 and taking into account the buildup and decay of

radionuclides disposed over the operating lifetime of the landfill, T L , as in Equation (6), the

annual individual dose, H i , from an annual disposal of a unit activity, A i , of 1 Ci (37 kBq) of

each photon-emitting radionuclide i in all landfills is given by

H i /A i (rem/Ci) = (1/N L ) × (1/λ R, i ) [1 exp(λ R, i T L )] × (1/M w T L ) (12)

× [(f in × f sh) + f out ] × D ext, i × exp(λ R, i T C ) ,

where the various parameters are defined with Equations (4) to (6), except the fraction of the

year during which exposure occurs is separated into the fraction of the time indoors, f in , and

outdoors, f out , the shielding factor, f sh , applies only during indoor residence, and the last term

in this equation represents radioactive decay during the time, T C , between closure of the

facility and the establishment of permanent residency on the landfill site.

The assumed number of operating landfills, N L, the values of the mass of waste disposed in a

landfill annually, M w , and the external dose coefficient, D ext , for each radionuclide are

described with Equation (4), and the operating lifetime of the landfill, T L , again is assumed to

be 30 years. The assumed values of the other parameters in Equation (12) are described as

follows:

 The fraction of the year during which indoor exposure occurs, f in , was 0.65

(EPA/600/P–95/002Fa).

 The shielding factor during indoor residence, f sh , was 0.7 (NRC, Regulatory Guide.

1.109).

 The fraction of the year during which outdoor exposure occurs, f out , was 0.05

(EPA/600/P–95/002Fa).

 The time period between closure of the landfill and the establishment of permanent

residency on the site, T C , was 30 years.

Given the assumptions about indoor and outdoor exposure times and the shielding factor during

indoor residence, the contribution to external dose while outdoors is only about 10% and, thus,

can be neglected. The assumed value of T C is based on the presumption that, given current

requirements in 40 CFR 258 for post-closure activities at landfills under RCRA, the sites will not

be released for unrestricted use by the public immediately upon closure.

For 85Kr, the assessment of external dose also assumes that only half of the disposed activity

remains in the waste following landfill operations (see Appendix A.2.3.2.1), and the dose

obtained from Equation (12) is reduced by a factor of 2.
A.2–21

A.2.3.5.2 Inhalation Exposure to Individuals

The annual EDE to an individual on-site resident from inhalation exposure is estimated using a

model of the form given by Equation (2) in Appendix A.2.3.1.2. Therefore, taking into account

the buildup and decay of radionuclides disposed over the operating lifetime of the landfill, T L, as

in Equation (6), and the time between landfill closure and the onset of permanent residency, T C,

as in Equation (12), the annual individual dose, H i , from an annual disposal of a unit activity, A i

of 1 Ci (37 kBq) of each radionuclide i in all landfills is given by

H i /A i (rem/Ci) = (1/N L ) × (1/λ R, i ) [1 exp(λ R, i T L )] × (1/M w T L ) × f a × U a (13)

× [(f in × L a, in ) + (f out × L a, out )] × D inh, i × exp(λ R, i T C) ,

where the various parameters are defined with Equations (2), (4), (5), (6), and (12), except U a

is the annual breathing rate and the atmospheric mass loading is separated into values indoors, 

L a, in , and outdoors, L a, out , which are applied to the corresponding indoor and outdoor exposure

times.

The assumed number of operating landfills, N L , and the values of the mass of waste disposed

in a landfill annually, M w , the respirable fraction of airborne material, f a , and the inhalation

dose coefficient, D inh , for each radionuclide are described with Equations (2) and (4). The

operating lifetime of the landfill, T L , and the time delay before the onset of permanent

residency, T C , each are again assumed to be 30 years. The fraction of the year during which

exposure occurs indoors, f in , and outdoors, f out , again are assumed to be 0.65 and 0.05,

respectively. Assumed values of the other parameters in Equation (13) are described as

follows:

 The annual breathing rate, U a , was 8400 m3

/yr, based on an assumed breathing rate

for resting and light activity of 23 m3

/day (EPA/600/P–95/002Fa).

 The atmospheric mass loading of waste outdoors, L a, out , was 4×105

 g/m3, which is an

average background dust loading (Anspaugh et al., 1975).

 The atmospheric mass loading of waste indoors, L a, in , was one-third of the value

outdoors (EPA, RAE–9232/1–2).

For thorium, the dose from inhalation exposure considers the contribution from 220Rn during

indoor residence. The dose from exposure to 220Rn during indoor residence is obtained from a

natural analog model (Cook and Hunt, 1994), which is based on the known average dose from

indoor radon per unit concentration of the parent radionuclide in surface soil. For an indoor

residence time of 0.5, the natural analog model gives an EDE from exposure to 220Rn of

2.7 sievert (Sv)/yr per GBq/m3

 (1.0×102 rem/yr per Ci/m3) of 232Th in soil. Thus, for the indoor

residence time of 0.65 assumed in this assessment, the EDE from inhalation of 220Rn is

3.5 Sv/yr per GBq/m3

 (1.3×102 rem/yr per Ci/m3) of 232Th in soil. The dose from exposure to

220Rn during outdoor residence is only a few percent of the dose during indoor residence (Cook

and Hunt, 1994) and, thus, can be neglected.
A.2–22

For uranium, only the short-lived decay products that would be in activity equilibrium with the

parent uranium isotopes at times shortly after chemical separation are considered in this

assessment. Therefore, the dose from inhalation of 222Rn is not considered, because the

parent radionuclide 226Ra builds up in the waste only at times long after chemical separation.

A.2.3.5.3 Ingestion Exposure to Individuals

The annual EDE to an individual on-site resident from ingestion exposure is estimated using a

model of the form given by Equation (3) in Appendix A.2.3.1.3. Therefore, taking into account

the buildup and decay of radionuclides disposed over the operating lifetime of the landfill, T L ,

as in Equation (6), and the time between landfill closure and the onset of permanent residency,

T C , as in Equation (12), the annual individual dose, H i , from an annual disposal of a unit

activity, A i , of 1 Ci (37 kBq) of each radionuclide i in all landfills is given by

H i /A i (rem/Ci) = (1/N L ) × (1/λ R, i ) [1 exp(λ R, i T L )] × (1/M w T L ) × U ing × T (14)

× D ing, i × exp(λ R, i T C ) , 

where the various parameters are defined with Equations (2) to (6) and (12).

The assumed number of operating landfills, N L , and the values of the mass of waste disposed

in a landfill annually, M w , and the ingestion dose coefficient, D ing , for each radionuclide are

described with Equations (2) to (4). The operating lifetime of the landfill, T L , and the time delay

before onset of permanent residence, T C , each are again assumed to be 30 years. Assumed

values of the other parameters in Equation (14) are described as follows:

 The ingestion rate of waste, U ing , was 4×103

 g/h, which is a value appropriate for

residential ingestion of soil and dust (EPA, OSWER Directive 9285.6–03).

 The exposure time was 440 h/yr, based on an assumption that ingestion exposure

occurs mainly during outdoor residence on the site and that the fraction of the year

during which exposure occurs outdoors is 0.05 (EPA/600/P–95/002Fa).

A.2.3.5.4 Collective Dose for Future On-Site Residents

As discussed at the beginning of Appendix A.2.3.5, the collective dose for future on-site

residents at a landfill is calculated for 1 year’s disposals of a unit activity of each radionuclide,

rather than the annual disposals over the operating lifetime of the landfill. In this assessment,

the disposals are assumed to occur in the last year of operations. Therefore, the annual

individual dose from 1 year’s disposals used in the calculation of collective dose is obtained

from Equations (12) to (14) by omitting the term describing the buildup and decay of activity

during the operating lifetime of the landfill, T L. This approach essentially distributes the year’s

disposals over the entire landfill.

The calculation of collective dose for future on-site residents at landfills is based on the

individual dose calculated as described above and the following assumptions. First, the number

of residents at each landfill site, based on the average density of suburban populations in the
A.2–23

United States (Bureau of Census, 1990) and the size of a typical landfill (EPA/530–SW88–034),

is 10 (i.e., the total number of residents at all 3,500 landfill sites is 35,000). Second, as

described in Appendix A.2.3.4.2, the collective dose is calculated by integrating the individual

dose over 1000 years, taking into account the exposed population and radioactive decay over

that time.

A.2.3.5.5 Results of Dose Calculations

Tables A.2.8 and A.2.9 present the annual individual and collective EDEs to future on-site

residents from disposals of a unit quantity of 1 Ci (37 kBq) of each radionuclide in all landfills

estimated in this assessment (i.e., the DSRs in Equation (1)). As discussed previously, the

individual doses in Table A.2.8 represent the annual dose from disposal of the unit quantity of

each radionuclide during each year over the assumed 30-year operating lifetime of landfills, but

the collective doses in Table A.2.9 represent the dose over 1000 years from disposal of the unit

quantity of each radionuclide during the last year of operations only.

In applying the DSRs to disposal of particular items containing radioactive material, the

following points should be noted. First, as in the other exposure scenarios for disposal in

landfills, all DSRs normally should be reduced by a factor of 0.8 to take into account the fraction

of all disposals sent to landfills when disposal by incineration also could occur.

Second, the DSRs for inhalation and ingestion assume that the radioactive materials are

dispersed in the landfill in a form that would be readily suspended into the air or ingested, but

this would not be the case for many items. Therefore, depending on the physical form of the

particular item of concern, the assessor could reduce the DSRs for these exposure pathways

whenever the radioactive materials should be significantly less dispersible than loose materials

in the waste. Because exposures are assumed to occur well after landfill closure, the correction

factors for leachability in water described in Appendix A.2.3.4.3 should be appropriate for

inhalation and ingestion exposure to future on-site residents, rather than the correction factors

for dispersibility during landfill operations described in Appendix A.2.3.1.5.

An additional correction is needed for addressing the situation where there is limited distribution

or use of the exempt material. As discussed in Section A.2.3.1.5, an adjustment should be

applied to reflect localized use and disposal. If the assumed number of items to be disposed of

annually is less than 3500, which is the assumed number of disposal sites, the DSRs should be

increased by the ratio of 3500 divided by the number of items. 

A.2.4 Dose Assessment for Disposal in Incinerators

This section presents the models and parameter values for estimating individual and collective

doses from disposal of radioactive materials in municipal incinerators, and the results of the

dose assessment in the form of doses per unit activity of radionuclides disposed in all

incinerators (i.e., the DSRs in Equation (1)) are tabulated. The groups of individuals considered

in the dose assessment and their associated exposure pathways are described in

Appendix A.2.2.2.
A.2–24

A.2.4.1 Waste Collectors

The exposure scenario and exposure pathways for waste collectors at municipal incinerators

would be the same as for waste collectors at landfills. Therefore, except for the assumption

about the number of operating facilities, the models and parameter values for estimating

individual and collective doses for waste collectors at incinerators would be the same as those

presented in Appendix A.2.3.1.

In this assessment, the number of operating incinerators is assumed to be 150 (Kiser, 1995). 

Therefore, since the dose to individual waste collectors from external, inhalation, and ingestion

exposure per unit activity of radionuclides disposed in all incinerators is inversely proportional to

the number of incinerators (see Appendix A.2.3.1), the individual dose would be a factor of

3500/150 = 23 higher than the corresponding individual dose for waste collectors for the same

unit activity disposed at all municipal landfills, where 3500 is the assumed number of operating

landfills. However, the collective dose to all waste collectors at incinerators per unit activity

disposed at all incinerators would be the same as the collective dose to all waste collectors at

landfills for the same unit activity disposed at all landfills.

Tables A.2.10 and A.2.11 present the annual individual and collective EDEs to waste collectors

at municipal incinerators from 1 year’s disposals of a unit quantity of 1 Ci (37 kBq) of each

radionuclide in all incinerators estimated in this assessment (i.e., the DSRs in Equation (1)). In

applying the DSRs to disposal of particular items containing radioactive material, the following

points should be noted.

First, as discussed in Appendix A.2.3.1.5, if a particular item could be sent to either landfills or

incinerators for disposal, the DSRs should be reduced by the fraction of the total annual

disposals assumed to be sent to incinerators. Recent data indicate that the amount of waste

sent to incinerators is about one-fourth of the amount sent to landfills (EPA–530–R–94–042). 

Therefore, if the input to the dose assessment is an assumed total activity of a radionuclide

disposed per year in all landfills and incinerators, the DSRs for all exposure pathways should be

reduced by a factor of 0.2 to consider the fraction of the disposed activity sent to incinerators.

Second, for some items, such as large bulk metal forms that are not normally used in consumer

products, it may be reasonable to assume that none of the material would be sent to

incinerators for disposal. In these cases, the exposure scenario could be assumed not to apply. 

Also, if the number of items to be incinerated annually is less than 150, which is the number of

assumed incinerators, the DSRs should be increased by the ratio of 150 divided by the number

of items.

Third, the DSRs for inhalation and ingestion exposure could be reduced whenever the

exempted items of concern should be significantly less dispersible than loose materials in the

waste. Suitable correction factors for dispersibility are discussed in Appendix A.2.3.1.5.

A final correction addresses the situation in which there is limited distribution or use of the

exempt material. As discussed in Section A.2.3.1.5, an adjustment should be applied to reflect

localized use and disposal. If the assumed number of items to be disposed of annually is less

than 3500, which is the assumed number of disposal sites, the DSRs should be increased by

the ratio of 3500 divided by the number of items.
A.2–25

A.2.4.2 Workers at Incinerator

Workers at municipal incinerators are assumed to receive external, inhalation, and ingestion

exposures while located at the edge of a partially enclosed waste pit where garbage trucks

unload waste at the facility. The dose assessment for incinerator workers is described in the

following paragraphs.

A.2.4.2.1 External Exposure to Individuals

The annual EDE to an individual incinerator worker from external exposure is estimated by

assuming that the source is an infinitely thick, uniformly contaminated volume of soil-equivalent

material and that the worker is standing at the edge of the source volume. The individual dose,

H i , from an annual disposal of a unit activity, A i , of 1 Ci (37 kBq) of each photon-emitting

radionuclide i in all incinerators is estimated using a model of the form given by Equation (4) in

Appendix A.2.3.2.1, with the following changes in the model and parameter values:

 The factor 1/N L , where N L is the number of operating landfills, is replaced by the factor

1/N I , where N I is the number of operating incinerators which, as noted in

Appendix A.2.4.1, is assumed to be 150.

 The mass of waste disposed in an incinerator annually, M w , was 2.0×1011 g, as

described below.

 The fraction of the year during which exposure occurs, f ex , was 0.09, based on an

assumption that a worker spends 1600 h/yr in the vicinity of the waste pit but only half of

this time at the edge of the waste pile.

 The shielding factor, f sh , is unity for a worker standing at the edge of the waste pile.

 For a worker standing at the edge of the waste pile, the external dose coefficient (D ext )

for each radionuclide is one-half of the value for a source of infinite lateral extent.

The mass of waste disposed in an incinerator was assumed to be the average value for all

incinerators. It was calculated from the reported amount of municipal waste combusted in 1993

of about 3.0×1010 kg (EPA–530–R–94–042) and the assumed number of incinerators.

As described in Appendix A.2.3.2.1, external exposure to incinerator workers to 85Kr is

estimated by assuming that half of the activity would be released into the air during unloading of

waste, due to breakage of half of the containers for this radionuclide. Therefore, the external

dose from 85Kr retained in the waste would be half of the value calculated as described above.

A.2.4.2.2 Inhalation Exposure to Individuals

The annual EDE to an individual incinerator worker from inhalation exposure is estimated using

a model of the form given by Equation (2) in Appendix A.2.3.1.2. The number of operating

incinerators, N I , and the mass of waste disposed in an incinerator annually, M w , were the

values given above for external exposure. The respirable fraction of airborne material, f a ,

again was 0.7, and the assumed values for the other parameters that differ from the values

given with Equation (2) are described as follows:
A.2–26

 The atmospheric mass loading of waste, L a , was 4×104

 g/m3, based on measurements

at an operating facility (Hahn et al., 1989).

 The exposure time for the worker, T, was 1600 h/yr.

A.2.4.2.3 Ingestion Exposure to Individuals

The annual EDE to an individual incinerator worker from ingestion exposure is estimated using

a model of the form given by Equation (3) in Appendix A.2.3.1.3. The number of operating

incinerators, N I , the mass of waste disposed in an incinerator annually, M w , and the exposure

time for the worker, T, were the values given above for external and inhalation exposure. The

assumed ingestion rate of waste, U ing , was 6×103

 g/h, which is a value appropriate for

commercial or industrial activities (EPA, OSWER Directive 9285.6–03).

A.2.4.2.4 Collective Dose for Incinerator Workers

In this assessment, a total of two workers are assumed to be located near the waste pit at each

incinerator during the working year. Therefore, for any exposure pathway, the annual collective

EDE to incinerator workers is given by the annual individual dose multiplied by the factor

2 × 150 = 300, where 150 again is the assumed number of operating incinerators.

A.2.4.2.5 Results of Dose Calculations

Tables A.2.12 and A.2.13 present the annual individual and collective EDEs to workers at

municipal incinerators from 1 year’s disposals of a unit quantity of 1 Ci (37 kBq) of each

radionuclide in all incinerators estimated in this assessment (i.e., the DSRs in Equation (1)). In

applying the DSRs to disposal of particular items containing radioactive material, the

corrections described in Appendix A.2.4.1 normally could be applied. The first is a reduction in

all DSRs by a factor of 0.2 to take into account the fraction of all disposals sent to incinerators

when disposal in landfills also could occur. Alternatively, when appropriate, it could be

assumed that none of the items would be sent to an incinerator for disposal, and the exposure

scenario could be assumed not to apply. The second correction is a reduction in the DSRs for

inhalation and ingestion that could be applied when the radioactive materials should be

significantly less dispersible than loose materials in the waste. The same correction factors for

dispersibility discussed in Appendix A.2.3.1.5 should be applicable to inhalation and ingestion

exposure to incinerator workers. Additionally, where there is limited distribution or use of the

exempt material as discussed in Section A.2.3.1.5, an adjustment should be applied to reflect

localized use and disposal. Also, if the number of items to be incinerated annually is less than

150, which is the number of assumed incinerators, the DSRs should be increased by the ratio

of 150 divided by the number of items.

A.2.4.3 Off-Site Members of the Public

During operations at incinerators, off-site individuals and populations are assumed to be

exposed to radionuclides released into the air and transported to off-site locations. The

following four exposure pathways are assumed to occur: (1) inhalation of airborne

radionuclides, (2) external exposure to airborne radionuclides, (3) external exposure to

radionuclides deposited on the ground surface, and (4) ingestion of food products contaminated

by deposition onto the ground surface.
A.2–27

Airborne releases at incinerators are assumed to result entirely from waste incineration (see

Appendix A.2.2.2.3). For radionuclides in particulate form, the fraction of material incinerated

that is released into the air from the incinerator stack is assumed to be 1.9×104

(EPA–450–AP–42–5ED). For 3H, which would not be in particulate form, all of the amount of

material disposed is assumed to be released into the air during incineration. Finally, half of the

total amount of 85Kr sent to incinerators is assumed to be released during operations at the

waste pits (see Appendix A.2.4.2.1), and the other half is assumed to be released during

incineration. However, for ease of calculation, doses are calculated by assuming that all of the

85Kr sent to incinerators is released from the stack during incineration.

For the assumed releases of radionuclides into the air described above, annual individual and

collective doses to nearby residents were calculated using CAP–88 (Computer Codes, Beres,

1990). Assumptions used in the calculations are described as follows:

 Releases occur at a height of 43 meters through a stack of diameter 1.5 meters. The

source temperature was 127C and the exit velocity was 7.3 m/s. The meteorological

data used in the calculations were for a site in Islip, NY, which is in an area with a

substantial number of operating incinerators.

 The urban agricultural data set in the CAP–88 computer code (see Appendix A.2.3.3)

was used to evaluate dose from ingestion of contaminated food products. The choice of

this data set reflects the location of many incinerators in urban or suburban areas.

 In calculating collective dose, a population of 2 million was assumed to be uniformly

distributed within a distance of 80 km of each incinerator. The assumed population was

based on the average population density in suburban areas in the United States (Bureau

of Census, 1990), and is intended to represent the urban or suburban locations of many

incinerators.

Table A.2.14 presents the annual individual and collective EDEs to off-site residents due to

airborne releases during waste incineration from 1 year’s disposals of a unit quantity of 1 Ci

(37 kBq) of each radionuclide in all incinerators estimated in this assessment (i.e., the DSRs in

Equation (1)). In applying the DSRs to disposal of particular items containing radioactive

material, the values normally should be reduced by a factor of 0.2 to take into account the

fraction of all disposals sent to incinerators when disposal in landfills also could occur. 

Alternatively, when appropriate, it could be assumed that none of the items would be sent to an

incinerator for disposal, and the exposure scenario could be assumed not to apply. 

Additionally, where there is limited distribution or use of the exempt material as discussed in

Section A.2.3.1.5, an adjustment should be applied to reflect localized use and disposal. Also,

if the number of items to be incinerated annually is less than 150, which is the number of

assumed incinerators, the DSRs should be increased by the ratio of 150 divided by the number

of items.

A.2.5 Dose Assessment for Metal Recycling

This assessment also considers doses to individual members of the public resulting from

recovery of items containing exempted amounts of radioactive materials for use in recycled

ferrous metals (e.g., steel). The groups of individuals considered in the dose assessment and
A.2–28

their associated exposure pathways are described in Appendix A.2.2.3. Only individual doses

are estimated for this disposal option, because recycling in metals is expected to be an unusual

occurrence for most exempted materials.

The estimates of individual dose from recycling of radionuclides in metals obtained in this

assessment are based directly on results of a previous study by Hill et al. (1995). The

assessment methodology used by Hill et al. resembles the methodology developed previously

by the International Atomic Energy Agency (IAEA) to derive exemption levels for recycle and

reuse of materials containing trivial quantities of radionuclides (IAEA SS No. 111–P–1.1).

Estimates of individual dose were obtained by Hill et al. (1995) for slag workers at metal

smelters, users of recycled products, and members of the public residing near smelters. In this

assessment, the previous results, which are in the form of annual EDEs per unit activity

concentration of radionuclides in the feed material sent to a smelter, are converted to annual

EDEs per unit activity of radionuclides sent to all smelters using the following assumptions. 

First, the number of facilities at which most smelting occurs is 100 (Phone call, C. Bechak,

Steel Manufacturing Association, Washington, DC, June 1996), and the unit activity of

radionuclides disposed per year is distributed equally among all smelters. Second, in

estimating doses to slag workers and users of recycled products, about 100 Mg/yr of metal are

sent to each smelter and are incorporated in the finished product, and about 10 Mg/yr of the

feed material are incorporated in the slag (Hill et al., 1995). Third, in estimating doses to

off-site members of the public, the fraction of radioactive material released to the air during

smelting is assumed to be 6.5×103

 (EPA–450–AP–42–5ED), except all 3H is assumed to be

released.

In this assessment, the recycled material is assumed to be steel used in automobiles (Hill et al.,

1995). Tables A.2.15 and A.2.16 present the annual individual EDEs from 1 year’s disposals of

a unit quantity of 1 Ci (37 kBq) of each radionuclide in all smelters estimated in this

assessment (i.e., the DSRs in Equation (1)). The results in Table A.2.15 give the estimated

dose to an individual slag worker or user of an automobile, whichever is greater, and an

identification of the critical exposure pathway. Table A.2.16 gives the estimated dose to an offsite member of the public from airborne emissions. The DSRs in these tables can be applied to

assumed activities of radionuclides sent to all smelters annually to obtain estimates of annual

EDEs to exposed individuals.

A.2.6 Application of Results

This section describes how the results given in this appendix are used to obtain estimates of

individual and collective doses from disposal of exempt items containing radioactive material. 

The generic disposal methodology has been developed with the intention of minimizing the

number of inputs and assumptions that the user must provide in applying the results.

In many of the tables of results, separate DSRs are given for external, inhalation, and ingestion

exposure. In general, the DSR for all exposure pathways is the sum of the DSRs for each

pathway.

Three types of inputs and assumptions must be provided by the user in applying the results in

the tables. The first input, which is always required, is the assumed total quantity of a
A.2–29

radionuclide disposed annually in all disposal facilities. However, in estimating doses from

disposal in landfills, incinerators, or metal smelters, no further assumptions are required about

the total number of facilities receiving the assumed annual disposals, because such

assumptions are incorporated in the model equations for calculating the DSRs given in the

tables.

The second input, which also is generally required, is an assumption about whether disposal in

landfills and incinerators would occur, or whether disposal would occur only in landfills. In the

latter case, the assumed annual disposals of a radionuclide would be applied directly to the

DSRs in the tables for disposal in landfills to obtain estimates of individual and collective doses,

and the DSRs in the tables for disposal in incinerators would be ignored. However, if disposal

in both landfills and incinerators is assumed to occur, the DSRs in the tables for landfills should

be reduced by a factor of 0.8, and the DSRs in the tables for incinerators should be reduced by

a factor of 0.2. The assumed total annual disposals of a radionuclide in all facilities then would

be applied to these results. This type of adjustment is not used in applying the DSRs in the

tables for recycling.

The third type of input consists of assumptions about the extent to which radionuclides could be

dispersed during waste collection and waste operations at landfills or incinerators or released

following disposal in landfills, compared with loose materials in the waste. Specifically, the

DSRs for inhalation and ingestion exposure to waste collectors, workers at landfills or

incinerators, and future on-site residents at landfills, and the DSRs for exposure to off-site

residents at landfills due to airborne releases during landfill operations, can be reduced if the

radioactive materials are considered to be significantly less dispersible than loose materials in

trash. The DSRs for exposure to off-site residents at landfills due to releases to groundwater

can be reduced if the radioactive materials are considered to be significantly less accessible to

infiltrating water than loose materials in the waste. Suggested dose reduction factors for these

exposure scenarios are discussed in Appendixes A.2.3.1.5 and A.2.3.4.3. If no assumptions

about dose reduction factors for dispersibility or accessibility to water are made by the user,

then the default dose estimates would be based only on the first two inputs described above.

A specific example is considered: disposal of 1 million items, each containing 1 g of thorium. 

Disposal in both landfills and incinerators is assumed to occur, and the radioactive material is

assumed to be in a physical form that is 10 times less dispersible and 10 times less accessible

to water than loose materials in the waste. Based on these assumptions, the following

estimates of dose are obtained:

 Waste collectors at landfills (Tables A.2.1 and A.2.2)

Annual individual dose = (106

 g) [(3.1×1010 rem/g) + (0.1)(5.8×1011 rem/g)

+ (0.1)(2.8×1011 rem/g)] (0.8) = 2.6×103

 mSv (0.26 mrem)

Annual collective dose = (106

 g) [(1.1×106 p-rem/g) + (0.1)(2.0×107 p-rem/g)

+ (0.1)(9.9×108

 p-rem/g)] (0.8) = 8.8×103 person-Sv (0.88 person-rem)
A.2–30

 Workers at landfills (Tables A.2.3 and A.2.4)

Annual individual dose = (106

 g) [(2.8×1011 rem/g) + (0.1)(6.3×1012 rem/g)

+ (0.1)(6.0×1012 rem/g)] (0.8) = 2.3×104

 mSv (0.023 mrem)

Annual collective dose = (106

 g) [(4.8×107 p-rem/g) + (0.1)(1.1×107 p-rem/g)

+ (0.1)(1.0×107

 p-rem/g)] (0.8) = 0.004 person-Sv (0.40 person-rem)

 Off-site residents at landfills due to airborne releases (Table A.2.5)

Annual individual dose = (106

 g) [(0.1)(3.3×1013 rem/g)] (0.8) = 2.6×107 mSv

(2.6×105

 mrem)

Annual collective dose = (106

 g) [(0.1)(2.0×107 p-rem/g)] (0.8) = 

1.6×104

 person-Sv (0.016 person-rem)

 Off-site residents at landfills due to releases to groundwater (Table A.2.7)

Annual individual dose = (106

 g) [(0.1)(2.4×1015 rem/g)] (0.8) = 1.9×109 mSv

(1.9×107

 mrem) Collective dose not calculated 

(exposures occur beyond 1000 years)

 Future on-site residents at landfills (Tables A.2.8 and A.2.9)

Annual individual dose = (106

 g) [(9.2×1011 rem/g) + (0.1)(2.3×1010 rem/g)

+ (0.1)(1.1×1013 rem/g)] (0.8) = 9.6×104

 mSv (0.096 mrem)

Collective dose over 1000 years = (106

 g) [(1.1×104 p-rem/g)

+ (0.1)(2.6×104

 p-rem/g) + (0.1)(1.3×107 p-rem/g)] (0.8)

= 1.1 person-Sv (110 person-rem)

 Waste collectors at incinerators (Tables A.2.10 and A.2.11)

Annual individual dose = (106

 g) [(7.2×109 rem/g) + (0.1)(1.4×109 rem/g)

+ (0.1)(6.6×1010 rem/g)] (0.2) = 0.015 mSv (1.5 mrem)

Annual collective dose = (106

 g) [(1.1×106 p-rem/g) + (0.1)(2.0×107 p-rem/g)

+ (0.1)(9.9×108

 p-rem/g)] (0.2) = 2.2×103 person-Sv (0.22 person-rem)

 Workers at incinerators (Tables A.2.12 and A.2.13)

Annual individual dose = (106

 g) [(5.4×1012 rem/g) + (0.1)(3.7×1012 rem/g)

+ (0.1)(1.8×1013 rem/g)] (0.2) = 1.2×105

 mSv (1.2×103 mrem)

Annual collective dose = (106

 g) [(1.6×109 p-rem/g) + (0.1)(1.1×109 p-rem/g)

+ (0.1)(5.4×1011 p-rem/g)] (0.2) = (3.4×106

 person-Sv)

(3.4×104

 person-rem)
A.2–31

 Off-site residents at incinerators due to airborne releases (Table A.2.14)

Annual individual dose = (106

 g)(3.8×1015 rem/g)(0.2) = 7.6×109 mSv

(7.6×107

 mrem)

Annual collective dose = (106

 g)(2.4×108 p-rem/g)(0.2) = (4.8×105 person-Sv)

(4.8×103

 person-rem)

Thus, based on the generic disposal methodology, the highest annual individual EDEs would be

(2.6 Sv (0.26 mrem)) to waste collectors at landfills or 0.015 mSv (1.5 mrem) to waste

collectors at incinerators. The collective EDE from 1 year’s distribution of the exempted items

would be 1.1 person-Sv (110 person-rem), due almost entirely to exposure to future on-site

residents at landfills for 1000 years after loss of institutional controls over the sites. If exposure

to future on-site residents were not considered, the collective EDE would be 0.015 person-Sv

(1.5 person-rem).

If the same quantity of thorium were sent to smelters for metal recycling, the estimated annual

individual EDEs, as obtained from Table A.2.15 and A.2.16, would be 0.48 mSv (48 mrem) to a

slag worker and 6.3×106

 mSv (6.3×104 mrem) to an off-site member of the public. However,

since these dose values are less than 1×105

 mSv (<0.001 mrem) they would have been

reported as less than values” (i.e., less than 1×105

 mSv (<0.001 mrem)) if they had been

included in a specific assessment presented in this report. This is consistent with the notation

that was used for this report.
A.2–32

Table A.2.1 Individual Dose-to-Source Ratios for Exposure to Waste

Collectors at Municipal Landfills a

Individual Effective Dose Equivalent

(rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 2.6×1017 3.3×1015

14C 5.9×1016 1.1×1013

36Cl 6.2×1015 1.6×1013

46Sc 1.5×109 8.3×1015 3.3×1013

55Fe 7.5×1016 3.2×1014

60Co 1.8×109 6.1×1014 1.4×1012

63Ni 8.7×1016 3.0×1014

85Kr 1.9×1012

90Sr + 90Y 3.7×1013 8.0×1012

99Tc 6.6×1016 2.8×1016 7.6×1014

106Ru + 106Rh 1.7×1010 1.3×1013 1.4×1012

109Cd + 109mAg 7.9×1012 3.2×1014 6.8×1013

129I 1.1×1011 4.9×1014 1.4×1011

133Ba 3.4×1010 2.2×1015 1.8×1013

137Cs + 137mBa 4.6×1010 9.0×1015 2.6×1012

147Pm 4.0×1015 1.1×1014 5.5×1014

152Eu 8.9×1010 6.2×1014 3.4×1013

204Tl 1.4×1012 6.7×1016 1.8×1013

210mBi + 206Tl 2.5×1010 2.3×1013 5.0×1012

210Po 6.7×1015 2.6×1012 9.9×1011

Th (natural)d 9.4×1010

(2.1×1010 rem/g)e

2.7×1010

(5.8×1011 rem/g)e

1.3×1010

(2.8×1011 rem/g)e

U (natural)f 1.5×1011

(9.8×1012 rem/g)g

3.5×1011

(2.4×1011 rem/g)g

1.4×1011

(9.9×1012 rem/g)g

U (depleted)h 2.2×1011

(8.0×1012 rem/g)i

3.4×1011

(1.2×1011 rem/g)i

1.4×1011

(5.2×1012 rem/g)i

241Am 2.7×1011 1.2×1010 1.9×1010

See following page for footnotes.
A.2–33

Footnotes to Table A.2.1

a

 Models and parameter values are presented in Appendix A.2.3.1. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all landfills. DSRs for all

exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 Ci (37 kBq) of

radionuclides per year is assumed to be disposed in landfills plus incinerators (see

Appendix A.2.3.1.5).

b

 1 rem/Ci = 0.27 Sv/Bq. c

 Depending on the physical form of the particular items of concern, DSR ratios for the

exposure pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.1.5).

d

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on assumed activity abundances of 232Th and

228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

f

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

h

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–34

Table A.2.2 Collective Dose-to-Source Ratios for Exposure to Waste

Collectors at Municipal Landfills a

Collective Effective Dose Equivalent

(person-rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 9.2×1014 1.2×1011

14C 2.1×1012 3.8×1010

36Cl 2.2×1011 5.5×1010

46Sc 5.4×106 2.9×1011 1.2×109

55Fe 2.6×1012 1.1×1010

60Co 6.4×106 2.2×1010 4.9×109

63Ni 3.1×1012 1.1×1010

85Kr 6.6×109

90Sr + 90Y 1.3×109 2.8×108

99Tc 2.3×1012 9.8×1013 2.7×1010

106Ru + 106Rh 5.9×107 4.7×1010 5.0×109

109Cd + 109mAg 2.8×108 1.1×1010 2.4×109

129I 4.0×108 1.7×1010 5.0×108

133Ba 1.2×106 7.7×1012 6.2×1010

137Cs + 137mBa 1.6×106 3.1×1011 9.1×109

147Pm 1.4×1011 3.9×1011 1.9×1010

152Eu 3.1×106 2.2×1010 1.2×109

204Tl 4.8×109 2.4×1012 6.1×1010

210mBi + 206Tl 8.6×107 8.0×1010 1.8×108

210Po 2.4×1011 9.2×109 3.5×107

Th (natural)d 3.3×106

(7.4×107 p-rem/g)e

9.3×107

(2.0×107 p-rem/g)e

4.5×107

(9.9×108 p-rem/g)e

U (natural)f 5.2×108

(3.4×108 p-rem/g)g

1.2×107

(8.4×108 p-rem/g)g

5.0×108

(3.4×108 p-rem/g)g

U (depleted)h 7.8×108

(2.8×108 p-rem/g)i

1.2×107

(4.4×108 p-rem/g)i

4.9×108

(1.8×108 p-rem/g)i

241Am 9.4×108 4.4×107 6.6×107

See following page for footnotes.
A.2–35

Footnotes to Table A.2.2

a

 Models and parameter values are presented in Appendix A.2.3.1. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all landfills. DSRs for all

exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 Ci (37 kBq) of

radionuclides per year is assumed to be disposed in landfills plus incinerators (see

Appendix A.2.3.1.5). 

b

 1 person-rem/Ci = 0.27 person-Sv/Bq. c

 Depending on the physical form of the particular items of concern, DSRs for the exposure

pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.1.5).

d

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

f

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

h

 Calculated DSRs take into account contributions from the short-lived decay products of 238U

and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–36

Table A.2.3 Individual Dose-to-Source Ratios for Exposure to Workers 

at Municipal Landfills a

Individual Effective Dose Equivalent

(rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 3.0×1018 7.0×1016

14C 6.4×1017 2.3×1014

36Cl 6.7×1016 3.3×1014

46Sc 2.0×1010 9.1×1016 7.0×1014

55Fe 8.3×1017 6.7×1015

60Co 2.5×1010 6.7×1015 3.0×1013

63Ni 9.5×1017 6.3×1015

85Kr 1.1×1013

90Sr + 90Y 4.0×1014 1.7×1012

99Tc 2.0×1015 3.1×1017 1.6×1014

106Ru + 106Rh 2.0×1011 1.5×1014 3.0×1013

109Cd + 109mAg 4.2×1013 3.5×1015 1.4×1013

129I 2.0×1013 5.3×1015 3.0×1012

133Ba 3.1×1011 2.4×1016 3.7×1014

137Cs + 137mBa 5.7×1011 9.8×1016 5.5×1013

147Pm 7.9×1016 1.2×1015 1.2×1014

152Eu 1.1×1010 6.8×1015 7.1×1014

204Tl 6.4×1014 7.4×1017 3.7×1014

210mBi + 206Tl 2.2×1011 2.6×1014 1.1×1012

210Po 8.2×1016 2.9×1013 2.1×1011

Th (natural)d 1.3×1010

(2.8×1011 rem/g)e 2.9×1011(6.3×1012 rem/g)e 2.7×1011(6.0×1012 rem/g)e

U (natural)f 1.3×1012

(8.9×1013 rem/g)g

3.9×1012

(2.7×1012 rem/g)g

3.0×1012

(2.1×1012 rem/g)g

U (depleted)h 2.1×1012

(7.6×1013 rem/g)i

3.7×1012

(1.4×1012 rem/g)i

3.0×1012

(1.1×1012 rem/g)i

241Am 6.9×1013 1.4×1011 4.0×1011

See following page for footnotes.
A.2–37

Footnotes to Table A.2.3

a

 Models and parameter values are presented in Appendix A.2.3.2. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all landfills. DSRs for all

exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 Ci (37 kBq) of

radionuclides per year is assumed to be disposed in landfills plus incinerators (see

Appendix A.2.3.1.5). 

b

 1 rem/Ci = 0.27 Sv/Bq. c

 Depending on the physical form of the particular items of concern, DSRs for the exposure

pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.1.5).

d

 Calculated DSRs take into account contributions from decay products of 232Th, which are

assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and activity

abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

f

 Calculated DSRs take into account contributions from short-lived decay products of 238U and

235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed

mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

h

 Calculated DSRs take into account contributions from the short-lived decay products of 238U

and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–38

Table A.2.4 Collective Dose-to-Source Ratios for Exposure to Workers

at Municipal Landfills a

Collective Effective Dose Equivalent

(person-rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 5.2×1014 1.2×1011

14C 1.1×1012 4.0×1010

36Cl 1.2×1011 5.8×1010

46Sc 3.5×106 1.6×1011 1.2×109

55Fe 1.4×1012 1.2×1010

60Co 4.5×106 1.2×1010 5.2×109

63Ni 1.7×1012 1.1×1010

85Kr 2.0×109

90Sr + 90Y 7.0×1010 2.9×108

99Tc 3.5×1011 5.5×1013 2.8×1010

106Ru + 106Rh 3.5×107 2.6×1010 5.3×109

109Cd + 109mAg 7.3×109 6.2×1011 2.5×109

129I 3.6×109 9.3×1011 5.3×108

133Ba 5.4×107 4.2×1012 6.5×1010

137Cs + 137mBa 9.9×107 1.7×1011 9.6×109

147Pm 1.4×1011 2.1×1011 2.0×1010

152Eu 1.9×106 1.2×1010 1.2×109

204Tl 1.1×109 1.3×1012 6.5×1010

210mBi + 206Tl 3.8×107 4.5×1010 1.8×108

210Po 1.4×1011 5.1×109 3.7×107

Th (natural)d 2.2×106

(4.8×107 p-rem/g)e

5.1×107

(1.1×107 p-rem/g)e

4.8×107

(1.0×107 p-rem/g)e

U (natural)f 2.3×108

(1.6×108 p-rem/g)g

6.8×108

(4.6×108 p-rem/g)g

5.3×108

(3.6×108 p-rem/g)g

U (depleted)h 3.6×108

(1.3×108 p-rem/g)i

6.5×108

(2.4×108 p-rem/g)i

5.2×108

(1.9×108 p-rem/g)i

241Am 1.2×108 2.4×107 7.0×107

See following page for footnotes.
A.2–39

Footnotes to Table A.2.4

a

 Models and parameter values are presented in Appendix A.2.3.2. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all landfills. DSRs for all

exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 Ci (37 kBq) of

radionuclides per year is assumed to be disposed in landfills plus incinerators (see

Appendix A.2.3.1.5). 

b

 1 person-rem/Ci = 0.27 person-Sv/Bq. c

 Depending on the physical form of particular items of concern, DSRs for the exposure

pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.1.5).

d

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

f

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

h

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–40

Table A.2.5 Individual and Collective Dose-to-Source Ratios for Exposure to Off-Site

Residents at Municipal Landfills Due to Airborne Releases During

Landfill Operations a

Radionuclide

Individual Effective Dose Equivalentb,c

(rem/Ci)d

Collective Effective Dose

Equivalentb,c

(person-rem/Ci)d

3

H 1.1×1017 3.3×1012

14C 1.7×1016 3.0×1010

36Cl 9.1×1014 1.5×107

46Sc 1.6×1015 1.6×109

55Fe 4.4×1017 5.5×1011

60Co 3.0×1014 3.1×108

63Ni 4.4×1017 3.5×1011

85Kr 2.9×1018 4.5×1012

90Sr + 90Y 2.2×1015 1.4×109

99Tc 1.9×1015 1.8×109

106Ru + 106Rh 3.7×1015 2.5×109

109Cd + 109mAg 1.2×1015 7.7×1010

129I 4.3×1014 6.0×109

133Ba 9.2×1015 9.4×109

137Cs + 137mBa 4.3×1015 5.2×109

147Pm 2.2×1016 1.4×1010

152Eu 3.0×1014 3.1×108

204Tl 3.4×1016 5.3×1010

210mBi + 206Tl 6.0×1015 e 3.7×109 e

210Po 9.1×1014 5.4×108

Th (natural)f 1.5×1012

(3.3×1013 rem/g)g

9.1×107

(2.0×107 p-rem/g)g

U (natural)h 6.0×1013

(4.1×1013 rem/g)i

3.5×107

(2.4×107 p-rem/g)i

U (depleted)j 5.7×1013

(2.1×1013 rem/g)k

3.3×107

(1.2×107 p-rem/g)k

241Am 2.7×1012 1.6×106

See following page for footnotes.
A.2–41

Footnotes to Table A.2.5 

a

 Models and parameter values are presented in Appendix A.2.3.3. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all landfills.

b

 DSRs should be reduced by a factor of 0.8 if a unit activity of 1 Ci (37 kBq) of radionuclides

per year is assumed to be disposed in landfills plus incinerators (see Appendix A.2.3.1.5).

c

 Depending on the physical form of the particular items of concern, DSRs could be reduced

whenever radioactive materials would be significantly less dispersible than loose materials in

trash (see Appendix A.2.3.1.5).

d

 1 rem/Ci = 0.27 Sv/Bq; 1 person-rem/Ci = 0.27 person-Sv/Bq. e

 Value estimated based on the calculated DSR for Th (natural), which results primarily from

inhalation exposure, and the assumption that the dose from 210mBi also results primarily from

inhalation exposure.

f

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

h

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

j

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

k

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–42

Table A.2.6 Solid and Solution Distribution Coefficients (Kd) for Elements a

Element

Kd

(mL/g) Element

Kd

(mL/g)

 H 0 I 1.0

 C 6.7 Ba 5.2×101

 Cl 1.7 Cs 2.7×102

 Sc 3.1×102 Pm 2.4×102

 Fe 1.6×102 Eu 2.4×102

 Co 6.0×101 Tl 3.9×102

 Ni 4.0×102 Bi 1.2×102

 Sr 1.5×101 Po 1.5×102

 Tc 1.0×101 Th 3.2×103

 Ru 5.5×101 U 1.5×101

 Cd 4.0×101 Am 1.9×103

a

 Values obtained from Table 6.7 of NUREG/CR–5512.
A.2–43

Table A.2.7 Individual and Collective Dose-to-Source Ratios for Exposure to Off-Site

Residents at Municipal Landfills Due to Releases to Groundwater a

Radionuclide

Timeb

(yr)

Individual Effective

Dose Equivalentc,d

(rem/Ci)f

Collective Effective

Dose Equivalentc,d,e

(person-rem/Ci)f

3

H 1.0×101 1.2×1015 1.0×109

14C 3.2×102 4.3×1015 6.5×108

36Cl 8.9×101 2.3×1014 4.9×107

46Sc

55Fe

60Co

63Ni

90Sr + 90Y 7.1×102 3.3×1021 4.5×1015

99Tc 1.5×101 6.8×1014 1.6×106

106Ru + 106Rh

109Cd + 109mAg

129I 5.7×101 3.3×1012 7.3×105

133Ba

137Cs + 137mBa

147Pm

152Eu

204Tl

210mBi + 206Tl 5.6×103 1.2×1014

210Po

Th (natural)g 1.5×105 1.1×1014

(2.4×1015 rem/g)h

U (natural)i 7.1×102 2.7×1013

(1.8×1013 rem/g)j

1.7×106

(1.2×106 p-rem/g)j

U (depleted)k 7.1×102 2.6×1013

(9.6×1014 rem/g)l

1.7×106

(6.2×107 p-rem/g)l

241Am

See following page for footnotes.
A.2–44

Footnotes to Table A.2.7

a

 Models and parameter values are presented in Appendix A.2.3.4. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual individual effective dose equivalents

(EDEs) from the disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all

landfills over 30-year operating lifetime and collective EDEs for 1,000 years after facility closure

from disposal of a unit activity of 1 Ci (37 kBq) of radionuclides in all landfills during the last

year of operations only.

b

 Travel time of radionuclide in groundwater from landfill to off-site municipal well. If no entry is

given, the travel time is much greater than the half-life of the radionuclide, and the resulting

doses are essentially zero (0).

c

 DSRs should be reduced by a factor of 0.8 if a unit activity of 1 Ci (37 kBq) of radionuclides

per year is assumed to be disposed in landfills plus incinerators (see Appendix A.2.3.1.5).

d

 Depending on the physical form of the particular items of concern, DSRs could be reduced

whenever radioactive materials should be significantly less accessible to infiltrating water than

loose materials in waste (see Appendix A.2.3.4.3).

e

 Collective dose is calculated only if the travel time of the radionuclide in groundwater is less

than 1,000 years.

f

 1 rem/Ci = 0.27 Sv/Bq; 1 person-rem/Ci = 0.27 person-Sv/Bq.

g

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1.

h

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

i

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

j

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

k

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

l

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–45

Table A.2.8 Individual Dose-to-Source Ratios for Exposure to Future On-Site

Residents at Municipal Landfills a

Individual Effective Dose Equivalent

(rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 4.2×1020 1.2×1018

14C 1.5×1017 4.4×1016

36Cl 1.6×1016 6.4×1016

46Scd

55Fe 1.2×1021 4.1×1020

60Co 4.1×1012 7.8×1018 2.7×1017

63Ni 1.7×1017 8.9×1017

85Kr 2.4×1014

90Sr + 90Y 3.3×1015 1.1×1014

99Tc 6.5×1015 9.8×1015 3.1×1016

106Ru + 106Rh 3.7×1021 1.9×1025 3.1×1025

109Cd + 109mAg 6.7×1021 4.1×1024 1.3×1023

129I 6.7×1013 1.3×1015 5.8×1014

133Ba 6.6×1012 3.7×1018 4.6×1017

137Cs + 137mBa 6.8×1011 8.6×1017 3.8×1015

147Pm 1.2×1019 1.3×1020 1.0×1020

152Eu 4.1×1011 1.8×1016 1.5×1016

204Tl 1.6×1016 1.3×1020 5.2×1019

210mBi + 206Tl 7.2×1011 5.6×1014 2.0×1014

210Pod

Th (natural)e 4.2×1010

(9.2×1011 rem/g)f

1.0×109 g

(2.3×1010 rem/g)f

2.6×1013

(5.7×1014 rem/g)f

U (natural)h 4.3×1012

(2.9×1012 rem/g)i

9.3×1013

(6.4×1013 rem/g)i

5.8×1014

(4.0×1014 rem/g)i

U (depleted)j 6.8×1012

(2.5×1012 rem/g)k

8.9×1013

(3.3×1013 rem/g)k

5.7×1014

(2.1×1014 rem/g)k

241Am 2.1×1012 3.1×1012 7.1×1013

See following page for footnotes.
A.2–46

Footnotes to Table A.2.8

a

 Models and parameter values are presented in Appendix A.2.3.5. Dose-to-source ratios

(DSRs) give annual individual effective dose equivalents from the disposal of a unit activity of

1 Ci (37 kBq) of radionuclides per year in all landfills over 30-year operating lifetime. DSRs

should be reduced by a factor of 0.8 if a unit activity of 1 Ci (37 kBq) of radionuclides per year

is assumed to be disposed in landfills plus incinerators (see Appendix A.2.3.1.5). 

b

 1 rem/Ci = 0.27 Sv/Bq. c

 Depending on the physical form of the particular items of concern, DSRs for the exposure

pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.5.5).

d

 Time after facility closure for first occurrence of on-site residence is much greater than the

half-life of the radionuclide, and the resulting doses are essentially zero (0).

e

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1.

f

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th and

228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

g

 Value is the dose from exposure to 220Rn and its short-lived decay products during indoor

residence (see Appendix A.2.3.5.2). Contribution from inhalation exposure to 232Th and its

other decay products is 3.5×1012 rem/Ci and, thus, is negligible by comparison.

h

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

j

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U, and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

k

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–47

Table A.2.9 Collective Dose-to-Source Ratios for Exposure to Future On-Site

Residents at Municipal Landfills a

Collective Effective Dose Equivalent

(person-rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 1.8×1015 5.1×1014

14C 1.7×1011 4.8×1010

36Cl 1.9×1010 7.4×1010

46Scd

55Fe 4.1×1017 1.5×1015

60Co 1.5×107 2.8×1013 9.7×1013

63Ni 3.2×1012 1.7×1011

85Kr 9.7×1010

90Sr + 90Y 2.3×1010 7.5×1010

99Tc 7.6×109 1.1×108 3.6×1010

106Ru + 106Rh 1.3×1016 6.7×1021 1.1×1020

109Cd + 109mAg 2.3×1016 1.4×1019 4.7×1019

129I 7.9×107 1.5×109 6.8×108

133Ba 2.7×107 1.5×1013 1.9×1012

137Cs + 137mBa 4.8×106 6.0×1012 2.7×1010

147Pm 4.2×1015 4.6×1016 3.5×1016

152Eu 1.8×106 8.1×1012 6.7×1012

204Tl 5.5×1012 4.6×1016 1.8×1014

210mBi + 206Tl 8.4×105 6.6×108 2.3×108

210Pod

Th (natural)e 4.9×104

(1.1×104 p-rem/g)f

1.2×103 g

(2.6×104 p-rem/g)f

6.1×107

(1.3×107 p-rem/g)f

U (natural)h 4.5×106

(3.1×106 p-rem/g)i

1.1×106

(7.4×107 p-rem/g)i

6.7×108

(4.6×108 p-rem/g)i

U (depleted)j 6.9×106

(2.6×106 p-rem/g)k

1.0×106

(3.8×107 p-rem/g)k

6.6×108

(2.4×108 p-rem/g)k

241Am 1.3×106 1.8×106 4.2×107

See following page for footnotes.
A.2–48

Footnotes to Table A.2.9

a

 Models and parameter values are presented in Appendix A.2.3.5. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give collective effective dose equivalents for

1,000 years after loss of institutional controls over disposal facilities from disposal of a unit

activity of 1 Ci (37 kBq) of radionuclides per year in all landfills during the last year of

operations only. DSRs should be reduced by a factor of 0.8 if a unit activity of 1 Ci (37 kBq) of

radionuclides per year is assumed to be disposed in landfills plus incinerators (see

Appendix A.2.3.1.5). 

b

 1 person-rem/Ci = 0.27 person-Sv/Bq. c

 Depending on the physical form of the particular items of concern, DSRs for the exposure

pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.5.5).

d

 Time after facility closure for first occurrence of on-site residence is much greater than the

half-life of the radionuclide, and resulting doses are essentially zero (0).

e

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1.

f

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th and

228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

g Value is dose from exposure to 220Rn and its short-lived decay products during indoor

residence (see Appendix A.2.3.5.2). Contribution from inhalation exposure to 232Th and its

other decay products is 8.2×106

 person-rem/Ci and, thus, is negligible by comparison.

h

 Calculated DSRs take into account the contributions from the short-lived decay products of

232U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

j

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

k

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–49

Table A.2.10 Individual Dose-to-Source Ratios for Exposure to Waste

Collectors at Municipal Incinerators a

Individual Effective Dose Equivalent

(rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 6.1×1016 7.8×1014

14C 1.4×1014 2.5×1012

36Cl 1.4×1013 3.7×1012

46Sc 3.6×108 1.9×1013 7.8×1012

55Fe 1.8×1014 7.4×1013

60Co 4.3×108 1.4×1012 3.3×1011

63Ni 2.0×1014 7.0×1013

85Kr 4.4×1011

90Sr + 90Y 8.6×1012 1.9×1010

99Tc 1.5×1014 6.6×1015 1.8×1012

106Ru + 106Rh 3.9×109 3.1×1012 3.3×1011

109Cd + 109mAg 1.8×1010 7.5×1013 1.6×1011

129I 2.7×1010 1.1×1012 3.4×1010

133Ba 8.0×109 5.1×1014 4.1×1012

137Cs + 137mBa 1.1×108 2.1×1013 6.1×1011

147Pm 9.3×1014 2.6×1013 1.3×1012

152Eu 2.1×108 1.4×1012 7.9×1012

204Tl 3.2×1011 1.6×1014 4.1×1012

210mBi + 206Tl 5.8×109 5.3×1012 1.2×1010

210Po 1.6×1013 6.2×1011 2.3×109

Th (natural)d 2.2×108

(4.8×109 rem/g)e

6.2×109

(1.4×109 rem/g)e

3.0×109

(6.6×1010 rem/g)e

U (natural)f 3.5×1010

(2.3×1010 rem/g)g

8.2×1010

(5.6×1010 rem/g)g

3.4×1010

(2.3×1010 rem/g)g

U (depleted)h 5.2×1010

(1.9×1010 rem/g)i

7.8×1010

(2.9×1010 rem/g)i

3.3×1010

(1.2×1010 rem/g)i

241Am 6.3×1010 2.9×109 4.4×109

See following page for footnotes.
A.2–50

Footnotes to Table A.2.10

a

 Models and parameter values are presented in Appendix A.2.4.1. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all incinerators. DSRs for

all exposure pathways should be reduced by a factor of 0.2 if a unit activity of 1 Ci (37 kBq) of

radionuclides per year is assumed to be disposed in landfills plus incinerators (see

Appendix A.2.3.1.5 and A.2.4.1).

b

 1 rem/Ci = 0.27 Sv/Bq. c

 Depending on the physical form of the particular items of concern, DSRs for the exposure

pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.1.5).

d

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

f

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

h

 Calculated DSRs take into account the contributions from the short-lived decay products of 

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–51

Table A.2.11 Collective Dose-to-Source Ratios for Exposure to Waste

Collectors at Municipal Incinerators a

Collective Effective Dose Equivalent

(person-rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 9.2×1014 1.2×1011

14C 2.1×1012 3.8×1010

36Cl 2.2×1011 5.5×1010

46Sc 5.4×106 2.9×1011 1.2×109

55Fe 2.6×1012 1.1×1010

60Co 6.4×106 2.2×1010 4.9×109

63Ni 3.1×1012 1.1×1010

85Kr 6.6×109

90Sr + 90Y 1.3×109 2.8×108

99Tc 2.3×1012 9.8×1013 2.7×1010

106Ru + 106Rh 5.9×107 4.7×1010 5.0×109

109Cd + 109mAg 2.8×108 1.1×1010 2.4×109

129I 4.0×108 1.7×1010 5.0×108

133Ba 1.2×106 7.7×1012 6.2×1010

137Cs + 137mBa 1.6×106 3.1×1011 9.1×109

147Pm 1.4×1011 3.9×1011 1.9×1010

152Eu 3.1×106 2.2×1010 1.2×109

204Tl 4.8×109 2.4×1012 6.1×1010

210mBi + 206Tl 8.6×107 8.0×1010 1.8×108

210Po 2.4×1011 9.2×109 3.5×107

Th (natural)d 3.3×106

(7.4×107 p-rem/g)e

9.3×107

(2.0×107 p-rem/g)e

4.5×107

(9.9×108 p-rem/g)e

U (natural)f 5.2×108

(3.4×108 p-rem/g)g

1.2×107

(8.4×108 p-rem/g)g

5.0×108

(3.4×108 p-rem/g)g

U (depleted)h 7.8×108

(2.8×108 p-rem/g)i

1.2×107

(4.4×108 p-rem/g)i

4.9×108

(1.8×108 p-rem/g)i

241Am 9.4×108 4.4×107 6.6×107

See following page for footnotes.
A.2–52

Footnotes to Table A.2.11

a

 Models and parameter values are presented in Appendix A.2.4.1. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all incinerators. DSRs for

all exposure pathways should be reduced by a factor of 0.2 if a unit activity of 1 Ci (37 kBq) of

radionuclides per year is assumed to be disposed in landfills plus incinerators (see

Appendix A.2.3.1.5 and A.2.4.1). 

b

 1 person-rem/Ci = 0.27 person-Sv/Bq. c

 Depending on the physical form of the particular items of concern, DSRs for the exposure

pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.1.5).

d

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

f

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

h

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–53

Table A.2.12 Individual Dose-to-Source Ratios for Exposure to Workers

at Municipal Incinerators a

Individual Effective Dose Equivalent

(rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 1.7×1018 2.1×1017

14C 3.7×1017 7.0×1016

36Cl 3.9×1016 1.0×1015

46Sc 3.9×1011 5.3×1016 2.1×1015

55Fe 4.8×1017 2.0×1016

60Co 4.9×1011 3.9×1015 9.0×1015

63Ni 5.6×1017 1.9×1016

85Kr 2.2×1014

90Sr + 90Y 2.3×1014 5.1×1014

99Tc 3.8×1016 1.8×1017 4.9×1016

106Ru + 106Rh 3.9×1012 8.6×1015 9.1×1015

109Cd + 109mAg 8.1×1014 2.1×1015 4.4×1015

129I 3.9×1014 3.1×1015 9.2×1014

133Ba 6.0×1012 1.4×1016 1.1×1015

137Cs + 137mBa 1.1×1011 5.7×1016 1.7×1014

147Pm 1.5×1016 7.0×1016 3.5×1016

152Eu 2.1×1011 4.0×1015 2.2×1015

204Tl 1.2×1014 4.3×1017 1.1×1015

210mBi + 206Tl 4.2×1012 1.5×1014 3.2×1014

210Po 1.6×1016 1.7×1013 6.3×1013

Th (natural)d 2.5×1011

(5.4×1012 rem/g)e

1.7×1011

(3.7×1012 rem/g)e

8.3×1013

(1.8×1013 rem/g)e

U (natural)f 2.5×1013

(1.7×1013 rem/g)g

2.3×1012

(1.5×1012 rem/g)g

9.2×1014

(6.3×1014 rem/g)g

U (depleted)h 4.0×1013

(1.5×1013 rem/g)i

2.1×1012

(7.9×1013 rem/g)i

9.0×1014

(3.3×1014 rem/g)i

241Am-241 1.3×1013 8.0×1012 1.2×1012

See following page for footnotes.
A.2–54

Footnotes to Table A.2.12

a

 Models and parameter values are presented in Appendix A.2.4.2. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all incinerators. DSRs for

all exposure pathways should be reduced by a factor of 0.2 if a unit activity of 1 Ci (37 kBq) of

radionuclides per year is assumed to be disposed in landfills plus incinerators (see

Appendix A.2.3.1.5 and A.2.4.2.5).

b

 1 rem/Ci = 0.27 Sv/Bq. c

 Depending on the physical form of the particular items of concern, DSRs for exposure

pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.1.5).

d Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

f

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

h

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–55

Table A.2.13 Collective Dose-to-Source Ratios for Exposure to Workers

at Municipal Incinerators a

Collective Effective Dose Equivalent

(person-rem/Ci)b

Radionuclide External Exposure Inhalation Exposurec Ingestion Exposurec

3

H 5.2×1016 6.4×1015

14C 1.1×1014 2.1×1013

36Cl 1.2×1013 3.0×1013

46Sc 1.2×108 1.6×1013 6.4×1013

55Fe 1.4×1014 6.1×1014

60Co 1.5×108 1.2×1012 2.7×1012

63Ni 1.7×1014 5.8×1014

85Kr 6.5×1012

90Sr + 90Y 7.0×1012 1.5×1011

99Tc 1.2×1013 5.5×1015 1.5×1013

106Ru + 106Rh 1.2×109 2.6×1012 2.7×1012

109Cd + 109mAg 2.4×1011 6.2×1013 1.3×1012

129I 1.2×1011 9.3×1013 2.8×1011

133Ba 1.8×109 4.2×1014 3.4×1013

137Cs + 137mBa 3.3×109 1.7×1013 5.0×1012

147Pm 4.6×1014 2.1×1013 1.1×1013

152Eu 6.4×109 1.2×1012 6.5×1013

204Tl 3.7×1012 1.3×1014 3.4×1013

210mBi + 206Tl 1.3×109 4.5×1012 9.6×1012

210Po 4.8×1014 5.1×1011 1.9×1010

Th (natural)d 7.4×109

(1.6×109 p-rem/g)e

5.1×109

(1.1×109 p-rem/g)e

2.5×1010

(5.4×1011 p-rem/g)e

U (natural)f 7.5×1011

(5.2×1011 p-rem/g)g

6.7×1010

(4.6×1010 p-rem/g)g

2.8×1011

(1.9×1011 p-rem/g)g

U (depleted)h 1.2×1010

(4.4×1011 p-rem/g)i

6.4×1010

(2.4×1010 p-rem/g)i

2.7×1011

(1.0×1011 p-rem/g)i

241Am 4.0×1011 2.4×109 3.6×1010

See following page for footnotes.
A.2–56

Footnotes to Table A.2.13 

a

 Models and parameter values are presented in Appendix A.2.4.2. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all incinerators. DSRs for

all exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 Ci (37 kBq) of

radionuclides per year is assumed to be disposed in landfills plus incinerators (see

Appendix A.2.3.1.5 and A.2.4.2.5).

b

 1 person-rem/Ci = 0.27 person-Sv/Bq. c

 Depending on the physical form of the particular items of concern, DSRs for exposure

pathway could be reduced whenever radioactive materials should be significantly less

dispersible than loose materials in trash (see Appendix A.2.3.1.5).

d

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

f

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

h

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–57

Table A.2.14 Individual and Collective Dose-to-Source Ratios for Exposure to Off-Site

Residents at Municipal Incinerators Due to Airborne Releases During

Incinerator Operations a

Radionuclide

Individual Effective Dose

Equivalent

(rem/Ci)b

Collective Effective Dose

Equivalent

(person-rem/Ci)b

3

H 1.0×1016 5.6×1010

14C 8.7×1019 1.1×1011

36Cl 2.5×1015 1.4×108

46Sc 7.3×1017 2.9×1010

55Fe 1.1×1018 5.7×1012

60Co 1.6×1015 5.9×109

63Ni 9.3×1019 6.8×1012

85Kr 5.6×1019 4.6×1012

90Sr + 90Y 3.9×1017 3.3×1010

99Tc 5.0×1017 4.7×1010

106Ru + 106Rh 7.5×1017 4.2×1010

109Cd + 109mAg 2.5×1017 2.0×1010

129I 3.1×1015 1.3×108

133Ba 5.0×1016 1.8×109

137Cs + 137mBa 1.1×1016 8.0×1010

147Pm 2.9×1018 1.9×1011

152Eu 1.6×1015 5.9×109

204Tl 9.4×1018 4.5×1011

210mBi + 206Tl 6.8×1017 c 4.4×1010 c

210Po 1.6×1015 1.3×108

Th (natural)d 1.7×1014

(3.8×1015 rem/g)e

1.1×107

(2.4×108 p-rem/g)e

U (natural)f 6.4×1015

(4.4×1015 rem/g)g

3.9×108

(2.7×108 p-rem/g)g

U (depleted)h 6.1×1015

(2.3×1015 rem/g)i

3.8×108

(1.4×108 p-rem/g)i

241Am 3.1×1014 2.0×107

See following page for footnotes.
A.2–58

Footnotes to Table A.2.14

a

 Models and parameter values are presented in Appendix A.2.4.3. Except as noted for thorium

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all incinerators. DSRs

should be reduced by a factor of 0.2 if a unit activity of 1 Ci (37 kBq) of radionuclides per year

is assumed to be disposed in landfills plus incinerators (see Appendix A.2.3.1.5 and A.2.4.3).

b

 1 person-rem/Ci = 0.27 person-Sv/Bq. c

 Value estimated based on the calculated DSR for natural thorium, which results primarily from

inhalation exposure, and the assumption that the dose from 210mBi also results primarily from

inhalation exposure.

d

 Calculated DSRs take into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

f

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in

Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

h

 Calculated DSRs take into account the contributions from the short-lived decay products of

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2);

assumed mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–59

Table A.2.15 Individual Dose-to-Source Ratios for Exposure to Slag Workers

at Metal Smelter or User of Automobile Containing Recycled Metal a

Radionuclide

Individual Effective

Dose Equivalent

(rem/Ci)b Exposed Individualc Exposure Pathwayd

3

H 1.8×1014 Slag worker Ingestion

14C 5.3×1013 Slag worker Ingestion

36Cl 8.1×1013 e Slag worker Ingestion

46Scf

55Fe 1.6×1013 e Slag worker Ingestion

60Co 2.3×108 Automobile user External

63Ni 1.9×1013 Slag worker Ingestion

85Kr f

90Sr + 90Y 4.1×1011 Slag worker Ingestion

99Tc 3.9×1013 e Slag worker Ingestion

106Ru + 106Rh 2.3×109 Automobile user External

109Cd + 109mAgf

129I 7.6×1011 Slag worker Ingestion

133Baf

137Cs + 137mBa 5.4×109 Automobile user External

147Pm 1.3×1012 Slag worker Inhalation

152Eu 1.1×108 Automobile user External

204Tlf

210mBi + 206Tlf

210Pof

Th (natural)g 2.2×108

(4.8×109 rem/g)h

Slag worker Inhalation

U (natural)i 3.7×109

(2.5×109 rem/g)j

Slag worker Inhalation

U (depleted)k 3.6×109

(1.3×109 rem/g)l

Slag worker Inhalation

241Am 1.7×108 Slag worker Inhalation

See following page for footnotes.
A.2–60

Footnotes to Table A.2.15

a

 Results are obtained from the previous analysis by Hill et al. (1995) as described in

Appendix A.2.5. Except as noted for thorium and uranium, dose-to-source ratios (DSRs) give

annual effective dose equivalents (EDEs) from disposal of unit activity of 1 Ci (37 kBq) of

radionuclides per year by metal recycling.

b

 1 rem/Ci = 0.27 Sv/Bq. c

 Individual, either slag worker or automobile user, receiving the highest dose.

d

 Limiting exposure pathway for determining dose to the slag worker or automobile user.

e

 Hill et al. (1995) erroneously reported that external exposure to the automobile user was the

limiting exposure pathway. Limiting exposure pathway was determined by analogy with the

results for 90Sr, and the annual EDE was obtained by multiplying the result for 90Sr by the

ingestion dose coefficient for the radionuclide relative to the value for 90Sr.

f

 Radionuclide was not included in analysis by Hill et al. (1995).

g

 Calculated DSR takes into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and the

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1.

h

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g).

i

 Calculated DSR takes into account the contributions from the short-lived decay products of

238U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed

mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in Table 3.1.1.

j

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g).

k

 Calculated DSR takes into account the contributions from the short-lived decay products of

238U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed

mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

l

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).
A.2–61

Table A.2.16 Individual Dose-to-Source Ratios for Exposure to Off-Site Residents

at Metal Smelter Due to Airborne Releases During Operations a

Radionuclide

Individual Effective

Dose Equivalent

(rem/Ci)b Radionuclide

Individual Effective

Dose Equivalent

(rem/Ci)b

3

H 1.9×1016 129I 6.7×1014

14C 5.7×1018 133Bac

36Clc 137Cs + 137mBa 6.6×1014

46Scc 147Pm 3.3×1016

55Fe 1.7×1017 152Eu 7.7×1014

60Co 7.4×1014 204Tlc

63Ni 2.2×1017 210mBi + 206Tlc

85Krc 210Poc 

90Sr + 90Y 2.9×1015 Th (natural)d 2.9×1012

(6.3×1013 rem/g)e

99Tc 1.9×1016 U (natural)f 3.1×1012

(2.1×1012 rem/g)g

106Ru + 106Rh 4.1×1015 U (depleted)h 4.7×1012

(1.7×1012 rem/g)i

109Cd + 109mAgc 241Am 4.6×1012

a

 Results are obtained from the previous analysis by Hill et al. (1995) as described in

Appendix A.2.5. Except as noted for thorium and uranium, dose-to-source ratios (DSRs) give

annual effective dose equivalents from disposal of unit activity of 1 Ci (37 kBq) of

radionuclides per year by metal recycling.

b

 1 rem/Ci = 0.27 Sv/Bq. c

 Radionuclide was not included in analysis by Hill et al. (1995).

d

 Calculated DSR takes into account the contributions from the decay products of 232Th, which

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. e

 Value in units of dose per unit mass is based on the assumed activity abundances of 232Th

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 Ci = 4.59 g). f

 Calculated DSR takes into account the contributions from the short-lived decay products of

238U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed

mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in Table 3.1.1.

g

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 1.46 g). h

 Calculated DSR takes into account the contributions from the short-lived decay products of

238U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed

mass and activity abundances of 238U, 235U, and 234U in depleted uranium are given in

Table 3.1.1.

i

 Value in units of dose per unit mass is based on the assumed activity abundances of 238U,

235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2

(1 Ci = 2.7 g).

A.3–1

A.3 GENERIC DISTRIBUTION METHODOLOGY

A.3.1 Introduction

Items that contain radioactive materials and are exempt from licensing requirements are very

diverse in size, shape, and intended use environment. Therefore, distribution of exempt items

can be accomplished by a variety of means. One or more items can be distributed from

manufacturers to users by direct, nonstop commercial truck; by commercial package or mail

delivery, which may involve truck and air transport and intermediate freight-handling terminals;

and by wholesale and retail firms, which may involve all of the above plus warehouses,

distribution centers, and retail stores.

Methods and practices commonly used to distribute commercial and consumer goods have

been characterized in a previous study (Etnier and O’Donnell, 1979). That study and personal

observations of distribution equipment and facilities have been used to define a set of scenarios

(i.e., typical distances and materials between people and packages and durations of exposure

to the packages) that characterize routine distribution practices. (Potential accidents during

distribution are addressed in Appendix A.1.) The defined scenarios, which are a small but

representative subset of the many possible distribution scenarios, include:

 Commercial truck transport, which includes (1) nonstop (express) delivery via small,

large, and tractor-trailer trucks; (2) local delivery via small and large trucks; and

(3) regional or long-distance transport via small, large, and tractor-trailer trucks (see

Appendix A.3.4).

 Warehousing, which includes handling in large warehouses (e.g., truck terminals) and

medium-sized warehouses (e.g., distribution centers) (see Appendix A.3.5).

 Retailing, which involves handling, storage, and display in small and large retail

establishments (see Appendix A.3.6).

 Air transport, which includes handling at receiving and shipping freight terminals and

exposures to flight crew and passengers on a regularly scheduled flight (see

Appendix A.3.7).

The defined scenarios can be combined to build a model that is representative of most common

distribution schemes (see Appendix A.3.3). Not all of the steps given in the scenarios may be

needed for delivery of a specific item. In such cases, the modeler may choose to use only the

appropriate parts of the scenarios. Also, item-specific analyses may be required for items that

have the potential to cause high radiation doses or that are distributed by methods not

characterized by the above.

The defined scenarios and the methods described in Appendix A.3.2 are used to calculate

individual and collective dose factors (DFs). A DF is the effective dose equivalent (EDE)

(sievert (Sv) (rem)) associated with the distribution of an exempt item containing 1 microcurie

(Ci) (37 kilobecquerel (kBq)) of byproduct material or 1 mg of source material.
A.3–2

C  S

E  SV

E2 t

(e EtV 1) VC0

Et (e EtV 1) , (1)

A.3.2 Derivation of the Dose Factors

Except for EDEs due to tritium (3H), all EDEs associated with distribution are due to external

exposures to radiation emitted from packages containing exempt items. These EDEs are

calculated using a personal computer (PC) version of the mainframe CONDOS II methodology

(Computer Codes, O’Donnell et al., 1981) and the exposure conditions given in Appendix A.3.4

through A.3.7. The PC version performs the same calculations as the mainframe version, with

the addition of the capability to calculate EDEs from internal and external exposures. Internal

EDE calculations use the dose conversion factors discussed in Sections 2.1 and 3.1. External

EDE calculations use the rotational geometry factors given in International Commission on

Radiological Protection (ICRP) Publication 51 to convert dose in air (rad), which is the primary

result of the CONDOS II calculations, to EDE (Sv (rem)).

One source configuration is used in all external dose calculations, a 30-cm-long by

7.6-cm-radius concrete-like cylinder modeled as a self-absorbing cylinder with the dose point on

the cylinder’s axis, with or without external shielding (Computer Codes, O’Donnell et al., 1981). 

The source contains either 1 Ci (37 kBq) of a byproduct material or 1 mg of a source material

that is distributed uniformly throughout the cylinder. Because concrete has a density of 2.3

g/cm3

 and the volume of the cylinder is 5600 cm3, the source mass is about 13 kg. Therefore,

the source strength is 7.8×105 Ci/g (2.9 Bq/g) of byproduct material or 7.8×105

 mg/g of

source material. The atomic number of concrete, which is used in the calculation of dose due

to bremsstrahlung, is 9.39.

EDEs due to 3

H exposure are calculated using Equation (4) of Appendix A.1.5 and the average

airborne concentrations of 3

H, which are given by

where C = average airborne concentration of 3

H (pCi/m3) during time t,

t = time (h) over which C is averaged,

C0 = initial airborne concentration of 3

H (pCi/m3),

S = source leak rate (pCi/h),

E = volume ventilation rate (m3

/h), and

V = volume (m3

).

Except for calculations involving exposures that begin after the leaking source has been

removed (e.g., for a person along a truck route after the truck has passed), C0 is set equal to

zero (0). For persons along truck routes, separate calculations were performed for the period

of vehicle passage and for the 8-hour period immediately after vehicle passage, for which C0

was set equal to the final value of C during the period of passage. The source was assumed to

contain 1 Ci (37 kBq) of 3

H and to have a fractional leak rate of 1 Pm/h; therefore, the value of

S used in the calculations is 1×106 Ci/h (3.7×102

 Bq/h).

Tables A.3.1 and A.3.2 are lists of highly exposed and average individual DFs, by nuclide and

delivery mode, for commercial truck transport of byproduct material. Table A.3.3 is a similar list
19 An additional step is required for 3

H. If the assumed leak rate differs from 1 Pm/h, the DFs

for 3

H should be multiplied by the assumed leak rate, which must be expressed in units of Pm/h.

A.3–3

of collective DFs for truck transport of byproduct material (based on average driver conditions). 

Tables A.3.4, A.3.5, and A.3.6 are, respectively, lists of highly exposed and average individual

DFs and collective DFs for truck transport of source material. Tables A.3.7 and A.3.8 are lists

of individual and collective DFs, respectively, for air transport, warehousing operations, and

retail sale of items containing byproduct or source materials.

A.3.3 Application of the Dose Factors

As illustrated below, application of the DFs requires the following six steps:

Step 1. Identification and listing of the steps (scenarios) involved in the chosen

distribution mode.

Step 2. Identification and quantification of the radionuclides to be distributed.

Step 3. Selection of highly exposed individual DFs, from Table A.3.1 or A.3.4 and

Table A.3.7, and collective DFs from Table A.3.3 or A.3.6 and Table A.3.8.

Step 4. Multiplication of the DFs by the quantity of radionuclide19 in the shipment to get

individual and collective EDEs for each step.

Step 5. Selection of highly exposed individual EDE for all steps.

Step 6. Addition of the collective EDEs for all steps to get the total collective EDE for the

shipment.

To illustrate use of the DFs, consider delivery of 10 items, each containing 1 Ci (37 kBq) of

60Co, via parcel delivery. There are several variations of parcel-delivery service. Some involve

delivery only by truck and some involve delivery by truck and airplane. Considered here is the

second delivery scheme, which requires 10 steps to move a shipment from a manufacturer to a

retailer.

Step 1. Transport from manufacturer to parcel-delivery center 1.

Step 2. Process at parcel-delivery center 1.

Step 3. Transport from parcel-delivery center 1 to air freight terminal 1.

Step 4. Process at air freight terminal 1 and load airplane.

Step 5. Transport from airport 1 to airport 2.

Step 6. Unload airplane and process at air freight terminal 2.

Step 7. Transport from air freight terminal 2 to parcel-delivery center 2.
A.3–4

Step 8. Process at parcel-delivery center 2.

Step 9. Transport from parcel-delivery center 2 to retailer.

Step 10. Sell at retail store.

As illustrated in Table A.3.9, Tables A.3.1 through A.3.8 are used to estimate a highly exposed

individual EDE and a collective EDE for each of the above steps. Each step is represented by a

scenario from the tables. Corresponding highly exposed individual DFs for 60Co are taken from

Table A.3.1 or A.3.7 and multiplied by 10 (10 Ci (370 kBq) are being shipped) to give a highly

exposed individual EDE for each step. Step 10, Retailing - small store, gives the highest

individual EDE (2×106

 Sv (2×104rem)) for this shipment. Similarly, collective DFs are taken

from Table A.3.3 or A.3.8 and multiplied by 10 to give the collective EDE associated with each

step. Scenario-specific EDEs then are summed to give the collective EDE associated with the

shipment, which is 4×104

 person-Sv (4×102 person-rem) in this example.

A.3.4 Commercial Truck

Essentially all distribution modes involve commercial truck transport. Three modes of truck

transport—express, local, and regional delivery—can be used to create any desired

truck-transport model.

1. An express delivery is a short-distance, nonstop delivery between, for example, a

manufacturer and a delivery service terminal or a delivery service terminal and a

customer. A typical express delivery covers 48 km in 1 hour of driving—19 km in

0.6 hour through high-population areas (3900 persons/km2

) and 29 km in 0.4 hour

through medium-population areas (730 persons/km2

).

2. A local delivery covers the same distance as an express delivery but involves multiple

stops at several delivery points. A typical express delivery covers 32 km and requires

4 hours, 1 hour driving and 3 hours making deliveries. Driving covers 13 km in 0.5 hour

through high-population areas and 19 km in 0.4 hour through medium-population areas. 

The delivery truck is parked in high-population areas for the 3 hours required to make

deliveries.

3. Regional delivery is used to transport goods over long distances (e.g., between truck

terminals). Over-the-road transport is a combination of regional deliveries that do not

involve unloading cargo at intermediate terminals. A typical regional delivery, which may

have one or multiple stops, covers 400 km and requires 5 hours of driving—20 km in

0.7 hour through high-population areas, 20 km in 0.3 hour through medium-population

areas, and 360 km in 4 hours through low-population areas (6 persons/km2

).

A variety of truck sizes and shapes is available for use in commercial trucking. Since it is not

possible to model every size and shape, three representative truck sizes (small delivery, large

delivery, and tractor-trailer) are used in the assessment:

 A small delivery truck is taken to be a van-like vehicle with no structural barrier between

driver and cargo. Typical dimensions are 305 cm long, 180 cm wide, 130 cm high, an
20 Many trucks fitting this definition of large have, in fact, separate driver and cargo

compartments. For the sake of being reasonably conservative, the case of joined

compartments is used.

A.3–5

enclosed volume of 7.1×106 cm3, and an air-ventilation rate of 5 volume changes per

hour. The driver sits 30 cm from the front and 180 cm from the center of the cargo. 

The truck body provides no shielding for the driver.

 A large delivery truck is taken to be a single-unit vehicle with no structural barrier

between driver and cargo.20 Typical dimensions are 500 cm long, 220 cm wide, 230 cm

high, an enclosed volume of 2.5×107

 cm3, and an air-ventilation rate of 3 volume

changes per hour. The driver sits 61 cm from the front and 340 cm from the center of

the cargo. The truck body provides no shielding for the driver.

 A tractor-trailer rig typically is 1370 cm long, 235 cm wide, and 270 cm high, and has a

cargo volume of 8.7×107

 cm3 with an air-ventilation rate of 1 volume change per hour. 

The driver sits 140 cm from the front and 825 cm from the center of the cargo. The

truck cab and trailer provide a 0.4-cm-thick aluminum shield for the driver. There is no

air ventilation between the cab and trailer.

Tables A.3.10 through A.3.12 provide the steps, types and numbers of persons exposed, and

exposure conditions involved in shipment of goods via express, local, and regional truck

delivery.

A.3.5 Warehousing

Except for direct delivery from a manufacturer to a customer or personal pickup by the

customer, all methods of distribution involve one or more warehouse-like facilities. These

facilities may include commercial truck terminals, large warehouses (chain-store or direct

merchandisers), and smaller distribution centers or stockrooms in retail establishments. In

such facilities, one or more two-person crews load and unload trucks, a forklift operator moves

pallets of material, and one or more sorters direct incoming packages to the appropriate

outgoing truck bay. Two warehouse-like facilities are considered in this appendix:

 A large warehouse is taken to be about 4600 cm long, 3050 cm wide, and 610 cm high,

and to have an enclosed volume of 8.5×109

 cm3 with an air-ventilation rate of 1 volume

change per hour. Each of five truck bays has a volume of 3.0×108

 cm3 with an airventilation rate of 4 volume changes per hour.

 A medium-sized warehouse is taken to be about 3050 cm long, 3050 cm wide, and

370 cm high, and to have an enclosed volume of 3.4×109

 cm3 with an air-ventilation rate

of 1 volume change per hour. Each of three truck bays has a volume of 3.0×108

 cm3

with an air-ventilation rate of 4 volume changes per hour.

The steps, types and numbers of workers, and exposure conditions involved in handling goods

in large and medium-sized warehouse-like facilities are given in Tables A.3.13 and A.3.14,

respectively.
A.3–6

A.3.6 Retailing

Exempt items that are designed for use by members of the public are frequently distributed

through retail stores. In such stores, the items are put on display and are near and handled by

sales clerks and customers. Two retail store sizes, large and small, are modeled in this

appendix:

 A large retail store is taken to be about 3050 cm long, 1520 cm wide, and 610 cm high,

and to have an enclosed volume of 2.8×109

 cm3 with an air-ventilation rate of 4 volume

changes per hour.

 A small retail store is taken to be about 1520 cm long, 910 cm wide, and 460 cm high,

and to have an enclosed volume of 6.4×108

 cm3 with an air-ventilation rate of 4 volume

change per hour.

The steps, types and numbers of persons, and exposure conditions involved in handling goods

in large and small retail stores are given in Tables A.3.15 and A.3.16, respectively.

A.3.7 Air Transport

Distribution of many smaller packages involves air transport. Typically, air cargo is delivered to

an air-freight terminal, where it is unloaded, sorted, and put into containers. (These steps may

be performed by airport or shipper’s personnel.) Containers are transported to and loaded on

an airplane, which transports the cargo to another airport, where the above steps are reversed. 

During the flight, exposed persons include the flight crew, attendants, and passengers:

 An air-freight terminal is taken to be about 3050 cm long, 3050 cm wide, and 610 cm

high, and to have an enclosed volume of 5.7×109

 cm3 with an air-ventilation rate of

1 volume change per hour. Each of three truck bays has a volume of 3.0×108

 cm3 and

an air-ventilation rate of 4 volume changes per hour.

 The cabin area of an airplane is taken to be about 1800 cm long, 460 cm wide, and

305 cm high, and to have an enclosed volume of 2.6×108

 cm3 with an air-ventilation rate

of 3 volume changes per hour. Air freight normally is stowed in two baggage

compartments, a forward and an anterior hold, located below the cabin area.

The steps, types and numbers of persons, and exposure conditions involved in air transport of

goods are given in Tables A.3.17 and A.3.18.
A.3–7

Table A.3.1 Highly Exposed (Package Near Driver) Individual DFs

for Commercial Truck Transport of Byproduct Material a

Express Delivery Local Delivery Regiona

Radionuclide Small Truck Large Truck Semi-Truck Small Truck Large Truck Small Truck Large 

110mAg 2.1×106 5.6×107 1.7×107 4.3×106 1.2×106 9.5×106 1.8×

241Am 4.8×109 1.3×109 3.7×1010 9.6×109 2.7×109 2.2×108 4.1×133Ba 2.6×107 6.7×108 2.0×108 5.2×107 1.4×107 1.2×106 2.2×14C 8.5×1012 2.4×1012 7.0×1013 1.8×1011 5.0×1012 4.0×1011 7.5×109Cd 1.4×109 3.9×1010 1.2×1010 2.8×109 8.5×1010 6.3×109 1.3×36Cl 1.4×109 3.8×1010 1.2×1010 2.9×109 8.1×1010 6.5×109 1.2×60Co 2.0×106 5.5×107 1.7×107 4.1×106 1.2×106 9.2×106 1.8×137Cs 4.5×107 1.2×107 3.6×108 9.1×107 2.5×107 2.0×106 3.8×3H 2.1×1012 1.2×1012 4.3×1013 6.4×1012 3.9×1012 1.2×1011 5.4×129I 3.9×1010 9.1×1011 3.0×1011 7.9×1010 1.9×1010 1.8×109 2.8×85Kr 3.3×109 8.5×1010 2.6×1010 6.6×109 1.8×109 1.5×108 2.8×63Ni 2.7×1013 7.0×1014 2.2×1014 5.5×1013 1.5×1013 1.2×1012 2.3×147Pm 1.9×1011 5.0×1012 1.5×1012 3.8×1011 1.1×1011 8.7×1011 1.7×210Po 5.8×1012 1.5×1012 4.6×1013 1.2×1011 3.2×1012 2.6×1011 4.9×46Sc 1.6×106 4.1×107 1.3×107 3.2×106 8.8×107 7.1×106 1.3×90Sr 1.3×109 3.4×1010 1.0×1010 2.6×109 7.3×1010 5.8×109 1.1×99Tc 8.8×1010 2.3×1010 7.1×1011 1.8×109 5.0×1010 4.0×109 7.6×

a Units are in rem/Ci shipped. 1 rem/Ci = 0.27 Sv/Bq.
A.3–8

Table A.3.2 Average (Package in Center of Cargo Area) Individual DFs 

for Commercial Truck Transport of Byproduct Material a

Express Delivery Local Delivery Regiona

Radionuclide Small Truck Large Truck Semi-Truck Small Truck Large Truck Small Truck Large 

110mAg 2.8×107 2.1×107 1.4×107 6.3×107 5.0×107 3.7×107 7.4×

241Am 6.0×1010 4.6×1010 3.1×1010 1.3×109 1.1×109 7.4×1010 1.5×133Ba 3.4×108 2.5×108 1.7×108 7.5×108 6.0×108 4.5×108 8.9×14C 1.2×1012 8.5×1013 6.0×1013 2.6×1012 2.1×1012 1.5×1012 3.2×109Cd 1.8×1010 1.4×1010 9.7×1011 4.1×1010 3.5×1010 2.4×1010 5.3×36Cl 1.9×1010 1.4×1010 9.6×1011 4.2×1010 3.4×1010 2.5×1010 5.1×60Co 2.7×107 2.0×107 1.4×107 6.0×107 4.9×107 3.6×107 7.3×137Cs 6.1×108 4.4×108 3.0×108 1.3×107 1.0×107 7.9×108 1.6×3H 2.1×1012 1.2×1012 4.3×1013 6.4×1012 3.9×1012 1.2×1011 5.4×129I 5.2×1011 3.8×1011 2.7×1011 1.1×1010 8.0×1011 6.0×1011 1.0×85Kr 4.4×1010 3.2×1010 2.2×1010 9.6×1010 7.6×1010 5.7×1010 1.1×63Ni 3.5×1014 2.7×1014 1.9×1014 7.5×1014 6.0×1014 4.3×1014 8.5×147Pm 2.5×1012 1.9×1012 1.3×1012 5.5×1012 4.5×1012 3.2×1012 6.7×210Po 7.8×1013 5.7×1013 3.8×1013 1.7×1012 1.3×1012 1.0×1012 2.0×46Sc 2.1×107 1.6×107 1.0×107 4.6×107 3.7×107 2.7×107 5.5×90Sr 1.7×1010 1.3×1010 8.6×1011 3.8×1010 3.0×1010 2.2×1010 4.5×99Tc 1.2×1010 8.7×1011 5.9×1011 2.6×1010 2.1×1010 1.5×1010 3.1×

a Units are in rem/Ci shipped. 1 rem/Ci = 0.27 Sv/Bq.
A.3–9

Table A.3.3 Collective DFs for Commercial Truck Transport 

of Byproduct Material a

Express Delivery Local Delivery Regiona

Radionuclide Small Truck Large Truck Semi-Truck Small Truck Large Truck Small Truck Large 

110mAg 3.0×107 2.2×107 1.4×107 7.3×107 5.4×107 3.9×107 8.3×

241Am 6.3×1010 4.7×1010 3.2×1010 1.5×109 1.2×109 7.7×1010 1.6×133Ba 3.7×108 2.6×108 1.7×108 8.8×108 6.5×108 4.7×108 9.8×14C 1.3×1012 9.0×1013 6.0×1013 3.0×1012 2.3×1012 1.6×1012 3.5×109Cd 2.0×1010 1.5×1010 9.9×1011 4.9×1010 3.9×1010 2.6×1010 6.0×36Cl 2.1×1010 1.5×1010 9.7×1011 5.0×1010 3.7×1010 2.6×1010 5.6×60Co 2.9×107 2.1×107 1.4×107 7.1×107 5.4×107 3.8×107 8.2×137Cs 6.5×108 4.6×108 3.0×108 1.6×107 1.1×107 8.3×108 1.7×3H 2.1×1012 1.2×1012 4.3×1013 6.4×1012 3.9×1012 1.2×1011 5.4×129I 5.4×1011 3.8×1011 2.7×1011 1.2×1010 8.4×1011 6.2×1011 1.1×85Kr 4.7×1010 3.3×1010 2.2×1010 1.1×109 8.3×1010 6.0×1010 1.3×63Ni 3.7×1014 2.8×1014 1.9×1014 8.5×1014 6.5×1014 4.5×1014 9.5×147Pm 2.7×1012 1.9×1012 1.3×1012 6.5×1012 4.9×1012 3.4×1012 7.3×210Po 8.3×1013 5.9×1013 3.9×1013 2.0×1012 1.5×1012 1.1×1012 2.2×46Sc 2.3×107 1.6×107 1.1×107 5.4×107 4.0×107 2.9×107 6.1×90Sr 1.8×1010 1.3×1010 8.7×1011 4.5×1010 3.3×1010 2.4×1010 5.1×99Tc 1.3×1010 9.1×1011 6.0×1011 3.0×1010 2.3×1010 1.6×1010 3.5×

a Units are in person-rem/Ci shipped. 1 person-rem/Ci = 0.27 person-Sv/Bq.
A.3–10

Table A.3.4 Highly Exposed (Package Near Driver) Individual DFs for Commercial Truck Transport of 

Express Delivery Local Delivery Regiona

Radionuclide Small

Truck 

Large Truck Semi-Truck Small Truck Large Truck Small Truck Large 

naturalTh (1 yr) 1.1×1010 3.0×1011 9.5×1012 2.2×1010 6.6×1011 4.9×1010 1.0×

naturalTh (20 yr) 2.0×1010 5.3×1011 1.7×1011 4.0×1010 1.2×1010 8.8×1010 1.8×depletedU 7.1×1012 1.9×1012 5.8×1013 1.4×1011 4.1×1012 3.2×1011 6.2×naturalU 8.1×1012 2.2×1012 6.6×1013 1.6×1011 4.7×1012 3.7×1011 7.1×

a Units are in rem/mg shipped. 1 rem/mg = 0.01 Sv/mg.
A.3–11

Table A.3.5 Average (Package in Center of Cargo Area) Individual DFs 

for Commercial Truck Transport of Source Material a

Express Delivery Local Delivery Regiona

Radionuclide Small Truck Large Truck Semi-Truck Small Truck Large Truck Small Truck Large 

naturalTh (1 yr) 1.5×1011 1.1×1011 7.6×1012 3.2×1011 2.7×1011 1.9×1011 4.1×

naturalTh (20 yr) 2.6×1011 2.0×1011 1.3×1011 5.8×1011 4.8×1011 3.4×1011 7.2×depletedU 9.4×1013 7.0×1013 4.8×1013 2.1×1012 1.7×1012 1.2×1012 2.5×naturalU 1.0×1012 8.0×1013 5.5×1013 2.4×1012 1.9×1012 1.4×1012 2.9×

a Units are in rem/mg shipped. 1 rem/mg = 0.01 Sv/mg.
A.3–12

Table A.3.6 Collective DFs for Commercial Truck Transport of Source Material a

Express Delivery Local Delivery Regiona

Radionuclide Small Truck Large Truck Semi-Truck Small Truck Large Truck Small Truck Large 

naturalTh (1 yr) 1.6×1011 1.2×1011 7.7×1012 3.9×1011 3.0×1011 2.0×1011 4.7×

naturalTh (20 yr) 2.8×1011 2.1×1011 1.4×1011 6.9×1011 5.3×1011 3.6×1011 8.1×depletedU 1.0×1012 7.3×1013 4.8×1013 2.5×1012 1.8×1012 1.3×1012 2.8×naturalU 1.2×1012 8.4×1013 5.5×1013 2.8×1012 2.1×1012 1.5×1012 3.2×

a Units are in person-rem/mg shipped. 1 person-rem/mg = 0.01 person-Sv/mg.
A.3–13

Table A.3.7 Highly Exposed Individual DFs for Air Transport, Warehousing, and Retaili

Air Transport Warehousing Re

Radionuclide Freight Terminal Airplane Large Warehouse Medium Warehouse Large Store 

110mAg 1.4×107 7.0×107 2.2×107 1.1×106 1.5×105 

241Am 3.1×1010 1.2×109 4.4×1010 2.0×109 3.2×108 133Ba 1.7×108 8.4×108 2.7×108 1.3×107 1.8×106 14C 6.0×1013 2.8×1012 9.0×1013 4.4×1012 6.0×1011 109Cd 9.6×1011 4.3×1010 1.4×1010 6.9×1010 9.5×109 36Cl 9.5×1011 4.7×1010 1.5×1010 7.3×1010 1.0×108 60Co 1.4×107 6.7×107 2.2×107 1.1×106 1.5×105 137Cs 3.0×108 1.5×107 4.7×108 2.3×107 3.3×106 3H 1.4×1012 2.4×1013 8.8×1013 1.0×1012 1.5×1011 129I 2.4×1011 8.2×1011 3.8×1011 1.5×1010 3.0×109 85Kr 2.1×1010 1.1×109 3.4×1010 1.7×109 2.4×108 63Ni 1.8×1014 7.0×1014 2.6×1014 1.2×1013 1.9×1012 147Pm 1.3×1012 5.9×1012 1.9×1012 2.3×1012 1.3×1010 210Po 3.8×1013 1.9×1012 6.1×1013 3.0×1012 4.2×1011 46Sc 1.0×107 5.2×107 1.7×107 8.2×107 1.1×105 90Sr 8.5×1011 4.2×1010 1.3×1010 6.6×1010 9.2×109 99Tc 5.8×1011 2.8×1010 9.1×1011 4.5×1010 6.3×109 naturalTh (1 yr)b 7.5×1012 3.6×1011 1.2×1011 6.0×1011 7.7×1010 naturalTh (20 yr)b 1.3×1011 6.4×1011 2.1×1011 1.1×1010 1.4×109 depletedUb 4.7×1013 2.3×1012 7.3×1013 3.7×1012 5.0×1011 naturalUb 5.4×1013 2.6×1012 8.4×1013 4.2×1012 5.7×1011 

a Units are in rem/Ci shipped. 1 rem/Ci = 0.27 Sv/Bq. b Units are rem/mg. 1 rem/mg = 0.01 Bq/mg.
A.3–14

Table A.3.8 Collective DFs for Air Transport, Warehousing, and Retailing a

Air Transport Warehousing Re

Radionuclide Freight Terminal Airplane Large Warehouse Medium Warehouse Large Store 

110mAg 7.1×107 1.4×105 1.6×106 6.4×106 2.3×103 

241Am 1.4×109 1.9×108 3.1×109 1.2×108 4.1×106 133Ba 8.4×108 1.7×106 2.0×107 7.6×107 2.6×104 14C 3.0×1012 5.5×1011 5.5×1012 3.3×1011 1.1×108 109Cd 4.8×1010 8.8×109 1.1×109 4.1×109 2.0×106 36Cl 4.8×1010 9.5×109 1.1×109 4.3×109 1.6×106 60Co 6.9×107 1.4×105 1.6×106 6.3×106 2.4×103 137Cs 1.5×107 3.0×106 3.4×107 1.3×106 4.4×104 3H 3.1×1012 3.1×1011 3.8×1012 1.5×1011 4.6×108 129I 1.1×1010 1.3×109 2.3×1010 8.5×1010 2.3×107 85Kr 1.1×109 2.2×108 2.5×109 9.7×109 3.3×106 63Ni 8.5×1014 1.3×1012 1.8×1013 7.0×1013 2.8×1010 147Pm 6.3×1012 1.2×1010 1.4×1011 5.5×1011 2.2×108 210Po 1.9×1012 3.9×1011 4.4×1012 1.7×1011 5.8×109 46Sc 5.2×107 1.1×105 1.2×106 4.8×106 1.7×103 90Sr 4.3×1010 8.5×109 9.8×1010 3.8×109 1.4×106 99Tc 2.9×1010 5.8×109 6.7×1010 2.6×109 9.8×107 naturalTh (1 yr)b 3.9×1011 7.8×1010 9.1×1011 3.5×1010 1.6×107 naturalTh (20 yr)b 6.8×1011 1.4×109 1.6×1010 6.2×1010 2.6×107 depletedUb 2.4×1012 4.7×1011 5.5×1012 2.1×1011 8.0×109 naturalUb 2.7×1012 5.4×1011 6.3×1012 2.4×1011 9.4×109 

a Units are in person-rem/Ci shipped. 1 person-rem/Ci = 0.27 person-Sv/Bq. b Units are rem/mg. 1 rem/mg = 0.01 Bq/mg.
A.3–15

Table A.3.9 Example Estimation of Individual and Collective Effective Dose Equivalents (E

Highly Exposed Individual Collective

Step Scenario Used Source Table

DF

(rem/Ci)a

EDE

(rem)a Source Table

DF

(person-rem/C

1 Express delivery

 Small truck

A.3.1 2.0×106 2×105 A.3.3 2.9×107 

2 Warehousing

 Large warehouse

A.3.7 2.2×107 2×106 A.3.8 1.6×106 

3 Express delivery

 Large truck

A.3.1 5.5×107 6×106 A.3.3 2.1×107 

4 Air transport

 Freight terminal

A.3.7 1.4×107 1×106 A.3.8 6.9×107 

5 Air transport

 Airplane

A.3.7 6.7×107 7×106 A.3.8 1.4×105 

6 Air transport

 Freight terminal

A.3.7 1.4×107 1×106 A.3.8 6.9×107 

7 Express delivery

 Large truck

A.3.1 5.5×107 6×106 A.3.3 2.1×107 

8 Warehousing

 Large warehouse

A.3.7 2.2×107 2×106 A.3.8 1.6×106 

9 Local delivery

 Small truck

A.3.1 4.1×106 4×105 A.3.3 7.1×107 

10 Retailing

 Small store

A.3.7 1.9×105 2×104 A.3.8 3.9×103 

a 1 rem/Ci = 0.27 Sv/Bq; 1 person-rem/Ci = 0.27 person-Sv/Bq. 1 rem = 0.01 Sv; 1person-rem = 0.01 person-
A.3–16

Table A.3.10 Steps, Types and Numbers of Persons Exposed, and Exposure Conditions

Involved in Shipment of Goods Via Express Truck Delivery

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

SMALL TRUCKS

1 driver

 Driving

 - high

 - average

1.0

1.0

0.016

0.016

30

180

Wood

Wood

15

15

 Handling 0.033 0.016 30 Wood 1.0

 In truck 0.3 0.016 90 Wood 15

 Near truck 0.66 0.00011 210 Wood

Aluminum

 30

0.32

120,000 persons in high-population zones

 Along route 0.050 7.7×1010 18,300 Wood

Aluminum

30

0.32

8.0 3.1×1011

34,200 persons in medium-population zones

 Along route 0.022 3.6×1010 18,300 Wood

Aluminum

30

0.32

8.0 1.5×1011

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–17

Table A.3.10 Steps, Types and Numbers of Persons Exposed, and Exposure Conditions 

Involved in Shipment of Goods Via Express Truck Delivery (continued)

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

LARGE TRUCKS

1 driver

 Driving

 - high

 - average

1.0

1.0

0.0091

0.0091

60

340

Wood

Wood

30

30

 Handling 0.033 0.0091 30 Wood 1.0

 In truck 0.50 0.0091 90 Wood 30

 Near truck 1.5 0.00018 310 Wood

Aluminum

60

0.32

120,000 persons in high-population zones

 Along route 0.050 6.5×1010 18,300 Wood

Aluminum

60

0.32

8.0 2.6×1011

34,200 persons in medium-population zones

 Along route 0.022 3.0×1010 18,300 Wood

Aluminum

60

0.95

8.0 1.2×1011

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–18

Table A.3.10 Steps, Types and Numbers of Persons Exposed, and Exposure Conditions 

Involved in Shipment of Goods Via Express Truck Delivery (continued)

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

TRACTOR-TRAILER TRUCKS

1 driver

Driving

 - high 1.0 0.0 140

Wood

Aluminum

60

0.95

 - average 1.0 0.0 830 Wood

Aluminum

60

0.95

Handling 0.033 0.00016 30 Wood 1.0

 In trailer 1.0 0.0037 120 Wood 60

 Near trailer 3.0 0.00025 460 Wood

Aluminum

120

0.32

120,000 persons in high-population zones

 Along route 0.050 3.5×1010 18,300 Wood

Aluminum

120

0.32

8.0 1.4×1011

34,200 persons in medium-population zones

 Along route 0.022 7.7×1010 18,300 Wood

Aluminum

120

0.32

8.0 6.7×1012

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–19

Table A.3.11 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Local Truck Delivery

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

SMALL TRUCKS

1 driver

 Driving

 - high

 - average

2.0

2.0

0.018

0.018

30

180

Wood

Wood

15

15

 Handling 0.33 0.0016 30 Wood 1.0

 In truck 1.0 0.016 90 Wood 15

 Near truck 3.0 0.000078 210 Wood

Aluminum

30

0.32

80,000 persons in high-population zones

 Along route 0.47 3.8×109 18,300 Wood

Aluminum

30

0.32

7.5 1.2×1010

22,800 persons in medium-population zones

 Along route 0.022 3.6×1010 18,300 Wood

Aluminum

30

0.32

8.0 1.5×1011

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–20

Table A.3.11 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Local Truck Delivery (continued)

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

LARGE TRUCKS

1 driver

 Driving

 - high

 - average

2.0

2.0

0.011

0.011

60

340

Wood

Wood

30

30

 Handling 0.033 0.00012 30 Wood 1.00

 In truck 2.0 0.011 90 Wood 30

 Near truck 3.0 0.000079 310 Wood

Aluminum

60

0.32

80,000 persons in high-population zones

 Along route 0.47 3.9×109 18,300 Wood

Aluminum

60

0.32

7.5 1.2×1010

22,800 persons in medium-population zones

 Along route 0.022 3.7×1010 18,300 Wood

Aluminum

60

0.32

8.0 1.5×1011

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–21

Table A.3.12 Steps, Types and Numbers of Persons Exposed, and Exposure

Conditions Involved in Shipment of Goods Via Regional Truck Delivery

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

SMALL TRUCKS

1 driver

 Driving 

 - high

 - average

5.0

5.0

0.019

0.019

 30

 180

Wood

Wood

15

15

125,000 persons in high-population zones

 Along route 0.050 9.2×1010 18,300 Wood

Aluminum

30

0.32

8.0 3.7×1011

23,750 persons in medium-population zones

 Along route 0.022 4.3×1010 18,300 Wood

Aluminum

30

0.32

8.0 1.7×1011

4,500 persons in low-population zones

 Along route 0.015 3.0×1010 18,300 Wood

Aluminum

30

0.32

8.0 1.2×1011

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–22

Table A.3.12 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Regional Truck Delivery (continued)

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

LARGE TRUCKS

1 driver

 Driving

 - high

 - average

5.0

5.0

0.012

0.012

 60

 340

Wood

Wood

30

30

125,000 persons in high-population zones

 Along route 0.050 8.9×1010 18,300 Wood

Aluminum

60

0.32

8.0 3.6×1011

23,750 persons in medium-population zones

 Along route 0.022 4.2×1010 18,300 Wood

Aluminum

60

0.32

8.0 1.7×1011

4,500 persons in low-population zones

 Along route 0.015 2.9×1010 18,300 Wood

Aluminum

60

0.32

8.0 1.2×1011

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–23

Table A.3.12 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Regional Truck Delivery (continued)

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

TRACTOR-TRAILER TRUCKS

 1 driver 

Driving

 - high 5.0 0.0 140

830

Wood

Aluminum

60

0.95

 - average 5.0 0.0 Wood

Aluminum

60

0.95

125,000 persons in high-population zones

 Along route 0.050 7.7×1010 18,300 Wood

Aluminum

120

0.32

8.0 3.1×1011

23,7500 persons in medium-population zones

Along route 0.022 3.6×1010 18,300 Wood

Aluminum

120

0.32

8.0 1.5×1011

4,500 person in low-population zones

 Along route 0.015 2.5×1010 18,300 Wood

Aluminum

120

0.32

8.0 1.0×1011

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–24

Table A.3.13 Exposure Conditions for a Large Warehouse

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

4 PRIMARY LOADERS AND UNLOADERS

Handle cargo 0.017 0.0088 30 Wood 1.0

In trailer 1.0 0.0088 120 Wood 60

In trailer bay 2.0 0.00073 460 Wood

Aluminum

120

1.0

16 OTHER LOADERS AND UNLOADERS (2 PER BAY)

1 bay away 2.0 0.00018 910 Wood

Aluminum

230

1.0

2 bays away 2.0 0.0000071 1,400 Wood

Aluminum

340

1.0

3 bays away 2.0 0.000011 1,800 Wood

Aluminum

450

1.0

4 bays away 2.0 0.0000029 2,300 Wood

Aluminum

550

1.0

5 FORK LIFT OPERATORS

Move cargo 0.050 0.00073 120 Wood

Iron

1.0

0.16

Other work 5.0 0.000094 1,000 Wood

Aluminum

Iron

250

10

1.0

10 STOREROOM CLERKS

Handle cargo 0.033 0.000094 30 Wood 1.0

Near cargo 1.0 0.000094 120 Wood 30

Other work 4.0 0.000094 610 Wood

Aluminum

Iron

15

2.0

0.50

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–25

Table A.3.14 Exposure Conditions for a Medium-Sized Warehouse

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

2 PRIMARY LOADERS AND UNLOADERS

Handle cargo 0.033 0.0088 30 Wood 1.0

In trailer 2.0 0.0088 120 Wood 60

In trailer bay 4.0 0.00073 460 Wood

Aluminum

120

1.0

4 OTHER LOADERS AND UNLOADERS (2 PER BAY)

1 bay away 6.0 0.00019 910 Wood

Aluminum

230

2 bays away 6.0 0.000048 1,400 Wood

Aluminum

340

1.0

3 FORKLIFT OPERATORS

Move cargo 0.050 0.00076 120 Wood

Iron

1.0

0.16

Other work 40 0.00029 1,000 Wood

Aluminum

Iron

250

10

1.0

6 STOREROOM CLERKS

Handle cargo 0.033 0.00029 30 Wood 1.0

Near cargo 8.0 0.00029 120 Wood 30

Other work 32 0.00029 610 Wood

Aluminum

Iron

15

2.0

0.50

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–26

Table A.3.15 Exposure Conditions for a Large Retail Store

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

(cm)

2 PRIMARY CLERKS

Handle

 product

3.3 0.000088 30 Wood 1.0

Near stored

 product

250 0.000088 460 Wood 3.0

Near display 250 0.000088 310 Wood

Aluminum

75

1.0

Other activities 1,500 0.000088 910 Wood

Aluminum

230

1.0

10 OTHER CLERKS

Near stored

 product

100 0.000088 460 Wood 3.0

Near display 130 0.000088 460 Wood

Aluminum

75

1.0

Other activities 1,800 0.000088 910 Wood

Aluminum

230

1.0

100,000 CUSTOMERS

Examining

 product

0.083 0.000088 30 Wood 1.0

Near display 5.0 0.000088 310 Wood

Aluminum

75

1.0

General

 shopping

55 0.000088 910 Wood

Aluminum

230

1.0

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–27

Table A.3.16 Exposure Conditions for a Small Retail Store

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

(cm)

1 PRIMARY CLERK

Handle product 3.3 0.00039 30 Wood 1.0

Near product 250 0.00039

310

Wood

Aluminum

3.0

1.0

Other activities 1,800 0.00039 460 Wood

Aluminum

120

1.0

2 OTHER CLERKS

Near display 100 0.00039 310 Wood

Aluminum

3.0

1.0

Other activities 1,900 0.00039 460 Wood

Aluminum

120

1.0

10 ,000 CUSTOMERS

Examining

 product

0.083 0.00039 30 Wood 1.0

Near display 9.9 0.00039 150 Wood

Aluminum

50

1.0

General

 shopping

50 0.00039 460 Wood

Aluminum

120

1.0

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–28

Table A.3.17 Exposure Conditions for an Air-Freight Terminal

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

(cm)

2 PRIMARY LOADERS AND UNLOADERS

Handle cargo 0.017 0.030 30 Wood 1.0

In truck 0.5 0.030 90 Wood 30

In trailer bay 2.0 0.00073 310 Wood

Aluminum

120

1.0

4 OTHER LOADERS AND UNLOADERS (2 PER BAY)

1 bay away 2.0 0.00073 910 Wood

Aluminum

230

1.0

2 bays away 2.0 0.00073 1,400 Wood

Aluminum

 340

3 FORKLIFT OPERATORS

Move cargo 0.050 0.000078 120 Wood

Iron

1.0

0.16

Other work 2.0 0.00010 1,000 Wood

Aluminum

Iron

250

10

1.0

6 SORTERS

Handle

 packages

0.017 0.00010 30 Wood 1.0

Near

 packages

2.0 0.00010 310 Wood 30

12 FREIGHT CLERKS

Fill container 0.017 0.00085 310 Wood 1.0

Near

 container

1.0 0.000065 120 Lucite

Wood

1.5

60

Other work 1.0 0.000033 460 Lucite

Wood

Aluminum

1.5

120

1.0
A.3–29

Table A.3.17 Exposure Conditions for an Air-Freight Terminal (continued)

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

(cm)

12 PLANE LOADERS

Move

 container

0.083 0.0000066 90 Lucite

Wood

Aluminum

1.5

1.0

0.16

Load

 container

0.25 0.00036 310 Lucite

Wood

Aluminum

Iron

1.5

250

10

1.0

Other work 0.75 0.00000019 410 Lucite

Wood

Aluminum

Iron

1.5

250

10

1.0

a

 1 pCi/m3 = 0.037 Bq/m3.
A.3–30

Table A.3.18 Exposure Conditions in an Airplane

Exposure

Event

Exposure

Time

(h)

Concentration

of 3

H in Air

(pCi/m3

)

a

Exposure

Distance

(cm)

Absorbers

Material Thickness

 (cm)

3 FLIGHT CREW MEMBERS

In cockpit 2.5 0.0011 1,100 Lucite

Aluminum

Wood

1.5

2.0

 20

3 FLIGHT ATTENDANTS

Either hold 0.50 0.0011

90

Lucite

Aluminum

Wood

1.5

1.0

10

Other

 locations

2.0 0.0011 1,000

1,200

Lucite

Aluminum

Wood

Aluminum

Wood

1.5

3.0

100

3.0

100

1 MAXIMALLY EXPOSED PASSENGER

Over hold 2.5 0.0011

90

Lucite

Aluminum

Wood

1.5

1.0

10

120 AVERAGE PASSENGERS

Both holds 2.5 0.0011

200

Lucite

Aluminum

Wood

1.5

2.0

20

a

 1 pCi/m3 = 0.037 Bq/m3.
A.4–1

A.4 OTHER MODELING METHODOLOGIES

A.4.1 Estimation of External Photon Dose Due to Bremsstrahlung Produced

by Low-Energy Electrons in Beta Decay

A.4.1.1 Introduction

In some of the dose assessments presented in this report, estimates of external dose from

exposure to photons emitted by radionuclides were obtained using CONDOS II (Computer

Codes, O’Donnell et al., 1981). For radionuclides that undergo beta decay, external doses

calculated using the CONDOS II code include a contribution due to bremsstrahlung, which is the

continuous spectrum of photons that results when an emitted electron (beta particle) is

decelerated by scattering in matter. This appendix presents an evaluation of the validity of 

CONDOS II in estimating external dose due to bremsstrahlung, and it describes the method used

to adjust the dose estimates given by CONDOS II to obtain more realistic results.

In the continuous spectrum of bremsstrahlung resulting from beta decay of radionuclides, the

number of photons per unit energy decreases rapidly with increasing energy between zero (0)

and a maximum energy equal to the endpoint energy of the continuous spectrum of beta

particles, and the energies of most of the photons are only a small fraction of the beta endpoint

energy (Evans, 1955). Thus, for example, when the endpoint energy of the beta spectrum is a

few hundred keV, the energies of most of the photons in the spectrum of bremsstrahlung are a

few tens of keV or less. At these low energies, estimates of external dose are the most

uncertain, especially when exposed individuals are assumed to be located in close proximity to

small sources. The difficulty in estimating external dose in these cases results, in part, from the

substantial variations in the energy and angular distributions of the radiation field over the body

surface and the rapid decrease in dose with decreasing photon energy and with increasing depth

in the body.

During routine use of some products or materials containing radionuclides that emit only beta

particles, external exposure is the only credible pathway and the calculated dose is due entirely

to bremsstrahlung. However, use of CONDOS II to evaluate doses raises the concern that it may

significantly overestimate the external dose due to bremsstrahlung, especially for radionuclides

that emit only low-energy beta particles.

The evaluation of CONDOS II in this appendix focuses on calculations of external dose from

exposure to low-energy beta-emitting radionuclides. Particular radionuclides of concern include

14C, 63Ni, and 147Pm, which have beta endpoint energies of 156 keV, 66 keV, and 225 keV,

respectively (Kocher, 1981). For 3

H, the dose due to bremsstrahlung is assumed to be zero (0)

as explained in Section A.4.1.4. As noted above, it is for such low-energy beta-emitting

radionuclides that the spectrum of bremsstrahlung is dominated by very low-energy photons and

estimates of external dose are the most uncertain.

The validity of CONDOS II in estimating external dose due to bremsstrahlung was evaluated by

comparing estimates of dose from exposure to beta-emitting radionuclides obtained using the

code with estimates given in Federal Guidance Report No. 12 (EPA–402–R–93–081). Based on

this evaluation, a simple adjustment factor was developed and applied to calculations using

CONDOS II to estimate external doses due to bremsstrahlung in assessments of products or
A.4–2

materials containing low-energy beta-emitting radionuclides. Other approaches to obtaining

more realistic estimates of the external dose due to bremsstrahlung were also considered. 

However, they were not adopted for the reasons discussed below.

A.4.1.2 Calculation of Bremsstrahlung Dose in CONDOS II Code

CONDOS II uses the method developed by Evans (1955) to determine the external dose due to

bremsstrahlung. Specifically, the external dose is calculated by using an approximate

representation of the continuous spectrum of photons produced by the scattering

of beta particles emitted by radionuclides. In this approximation, the number of photons of

energy Eγ per MeV per beta emission, denoted by d(Eγ)/dEγ, is given by

 . (1)

dI(Eγ)

dEγ

 1.02kZabs

W0

W1

(W W1)N(W) dW

W0

1

N(W) dW

 

In this equation, k is a constant equal to 7×104

 MeV1, Z abs is the atomic number of the

absorbing material in which the bremsstrahlung is produced, W is the total energy of a particular

beta particle in units of its rest-mass energy of 0.51 MeV given by

W  , (2) E

0.51  1

where E is the kinetic energy of the beta particle in MeV, W 0 is the value of W corresponding to

the endpoint energy of the beta spectrum in the decay of the radionuclide, W 1 is the value of W

corresponding to the photon energy Eγ, and N(W) dW is the continuous spectrum of electrons

produced in beta decay represented by

N(W) dW  const (W . (3) 2 1)1/2(W0 W)

2

W dW

In this approximation, the total number of bremsstrahlung photons per beta decay depends on

the atomic number of the absorbing material, Z abs, but the shape of the spectrum of photons

does not depend on the absorbing material.

There may be considerable uncertainty in the approximations used to calculate the spectrum of

bremsstrahlung photons described above. The constant of k in Equation 1 appears to be

uncertain by about a factor of 2 (Evans, 1955). An uncertainty in the value of k results in the

same uncertainty in the total number of bremsstrahlung photons, but there would be no effect on

their energy distribution. The representation of the continuous spectrum of beta particles given

by Equation (3) is an approximation that is exact only for a radionuclide with Z = 0 (Evans, 1955). 

The uncertainty in this approximation affects the energy distribution of bremsstrahlung photons.
A.4–3

In implementing Equations (1) through (3) in CONDOS II, the range of photon energies between

zero (0) and the endpoint energy of the beta spectrum is divided into a number of energy

intervals (Table A.1 of O’Donnell et al. (Computer Codes, 1981)), and the photon intensities and

resulting external doses are calculated at each of the discrete energies used to represent these

intervals. In the lowest energy interval, which includes the greatest number of bremsstrahlung

photons and is of greatest interest in evaluating the validity of calculations using CONDOS II, all

photons with energies between zero (0) and 12 keV are assumed have an energy of 10 keV.

A.4.1.3 Comparison of CONDOS II Code With Federal Guidance Report

Federal Guidance Report No. 12 (EPA–402–R–93–081) contains current Federal guidance on

external dose coefficients (i.e., external dose-equivalent rates per unit concentration) for

exposure to radionuclides in air, water, and soil. These dose coefficients were obtained using

sophisticated numerical methods that represent the current state-of-the-art in calculations of

external dose due to exposure to radionuclides dispersed in the environment.

For beta-emitting radionuclides, the external dose coefficients given in Federal Guidance Report

No. 12 include a contribution due to bremsstrahlung. In contrast to the approximate methods

used in CONDOS II, the dose due to bremsstrahlung included in Federal Guidance Report

No. 12 was calculated using realistic representations of the energy spectrum of bremsstrahlung

for a particular energy of an emitted beta particle and the energy spectrum of electrons in beta

decay of a particular radionuclide. These representations were obtained using sophisticated

numerical methods.

Another important difference between CONDOS II and the methods used in Federal Guidance

Report No. 12 is in the approach to considering the lowest energy bremsstrahlung photons in any

beta decay and the bremsstrahlung resulting from beta spectra with low endpoint energies. In

particular, the following two assumptions were used in Federal Guidance Report No. 12 in

calculating external dose due to bremsstrahlung:

 The external dose due to bremsstrahlung was assumed to be zero (0) for all photon

energies less than 10 keV.

 The external dose due to bremsstrahlung was assumed to be zero (0) for any beta

endpoint energy less than 100 keV.

The second assumption is based on the first because, for beta endpoint energies less than

100 keV, most of the bremsstrahlung photons have energies less than 10 keV.

Thus, the dose calculations in Federal Guidance Report No. 12 ignore the lowest energy

bremsstrahlung photons by using energy cutoffs in the spectrum of photons and in the endpoint

energy of beta spectra, but no such cutoffs are used in CONDOS II. This difference is

particularly important when nearly all photons have very low energies, as is the case for low beta

endpoint energies. The primary justification for use of energy cutoffs in Federal Guidance

Report No. 12 is that the external dose from the lowest energy bremsstrahlung photons should

be insignificant in any credible scenarios for exposure to radionuclides in the environment, when

all radionuclides and exposure pathways are taken into account.
A.4–4

In this evaluation, external doses calculated using CONDOS II are compared with the dose

coefficients tabulated in Federal Guidance Report No. 12 (EPA–402–R–93–081). The

calculations were performed for different beta-emitting radionuclides having a wide range of beta

endpoint energies. The source was assumed to be a uniformly contaminated ground surface,

which can be represented as a very large disk in CONDOS II (Computer Codes, O’Donnell et al.,

1981), and the dose was calculated at 1 meter above the ground. Of the source distributions

included in Federal Guidance Report No. 12, a contaminated ground surface is the most

appropriate for this comparison because the sources of concern in assessments of products or

materials containing beta-emitting radionuclides are usually assumed to be point sources and a

plane source can be regarded as an infinite array of unshielded point sources at varying

distances from a receptor.

The comparison of external dose coefficients for a contaminated ground surface calculated using

CONDOS II with the corresponding dose coefficients given in Federal Guidance Report No. 12 is

shown in Table A.4.1. Excluding the radionuclides with beta endpoint energies less than

100 keV, the results of this comparison may be summarized as follows:

 For all radionuclides, the dose calculated using CONDOS II is higher than the

corresponding value given in Federal Guidance Report No. 12.

 The degree of overestimation in the doses calculated using CONDOS II is the highest

when the beta endpoint energy is close to the cutoff of 100 keV used in Federal Guidance

Report No. 12. The substantial differences in these cases presumably reflect, at least in

part, the use of a photon energy cutoff of 10 keV in Federal Guidance Report No. 12, in

contrast to the inclusion of all low-energy photons in CONDOS II, because the spectrum

of bremsstrahlung is dominated by the lowest energy photons. However, other

differences between the approximate methods used in CONDOS II and the more rigorous

methods used in Federal Guidance Report No. 12 also could be important.

 The difference between the two dose estimates decreases as the beta endpoint energy

increases. This result presumably reflects the decreasing importance of the lowest

energy photons as the maximum photon energy in the spectrum of bremsstrahlung

increases.

 The difference in the dose estimates for 90Sr/90Y is only 25%. This comparison suggests

that the approximate methods used in CONDOS II give reasonable results for

radionuclides having the highest beta endpoint energies (i.e., when there are significant

intensities of higher energy photons in the spectrum of bremsstrahlung).

For radionuclides with beta endpoint energies less than 100 keV, the comparisons at higher

endpoint energies suggest that CONDOS II substantially overestimates the dose due to

bremsstrahlung. The degree of overestimation cannot be determined, but it probably is greater

than that found for radionuclides with beta endpoint energies slightly above 100 keV.

A.4.1.4 Approach to Estimating Dose Due to Bremsstrahlung

As noted in Section A.4.1.1, the particular beta-emitting radionuclides of concern to this report in

regard to estimating external dose due to bremsstrahlung include 14C, 63Ni, and 147Pm. These

radionuclides have relatively low beta endpoint energies and, as indicated in Table A.4.1,
A.4–5

calculations using CONDOS II probably overestimate the external dose due to bremsstrahlung

by a substantial amount, i.e., by more than a factor of 10. For 3H, the dose due to

bremsstrahlung is assumed to be zero (0) as explained below.

A simple approach that should provide more realistic estimates of external dose from exposure to

14C and 147Pm is to reduce the doses calculated using CONDOS II based directly on the ratios of

doses given in Table A.4.1. In particular, these results suggest that calculated doses should be

reduced by a factor of 20 for 14C and a factor of 15 for 147Pm. Such a reduction should be

reasonable even though the comparison in Table A.4.1 applies to exposure to a contaminated

ground plane, whereas exposure to a point source is the usual assumption in the assessments of

products or materials containing these radionuclides.

A similar reduction in calculated doses from exposure to 63Ni based on the results in Table A.4.1

is more uncertain. One option would be to ignore external doses for 63Ni calculated using

CONDOS II (i.e., a dose of zero (0) could be assumed in all assessments). This option would be

consistent with current Federal guidance (EPA–402–R–93–081). A second option would be to

reduce calculated doses for 63Ni by a factor of 20, based on the reduction factor for 14C obtained

from Table A.4.1 and the observation that the discrepancy between doses calculated using the

CONDOS II code and results given in Federal Guidance Report No. 12 increases with

decreasing beta endpoint energy. With this assumption, the dose estimates for 63Ni should be

more realistic but still conservative. The second option is used in this report, primarily because

the external dose due to bremsstrahlung resulting from decay of 63Ni is important only in

assessments where no other exposure pathways are assumed to occur.

The beta endpoint energy for 3

H, which also is included in Table A.4.1, is very low. Since the

energies of nearly all bremsstrahlung photons are less than 1 keV, the most reasonable option is

to assume that the dose due to bremsstrahlung is zero (0), which again would be consistent with

current Federal guidance. This assumption also can be justified on the grounds that all

assessments of products or materials containing 3

H assume that some release of activity from a

source occurs, and the resulting dose due to inhalation or absorption through the skin generally

would be much more important than the external dose due to bremsstrahlung.

A.4.1.5 Alternatives to Estimating Dose Due to Bremsstrahlung

For the low-energy beta-emitting radionuclides listed in Table A.4.1, the alternative of replacing

dose estimates obtained using CONDOS II with estimates obtained using more rigorous and

sophisticated methods, such as those described in Federal Guidance Report No. 12

(EPA–402–R–93–081), could be considered. Such calculations would require the use of

complex computer codes.

In considering whether the use of more rigorous and sophisticated calculations in estimating

external dose from exposure to low-energy beta-emitting radionuclides is justified, the limitations

of any such calculations in regard to obtaining realistic estimates of dose due to bremsstrahlung

should be recognized. These limitations result from two factors: (1) the primary importance of

very low-energy photons in the spectrum of bremsstrahlung for these radionuclides, and (2) the

assumption of idealized exposure conditions in any dose assessment (e.g., exposure at a fixed

distance from a point source shielded only by air or a simple configuration of another material). 

Even the most sophisticated calculations would not provide an accurate accounting of the

significant scattering and absorption of the lowest energy photons in materials used in the source
A.4–6

mounting, other materials located close to the source (e.g., in an instrument housing), and

clothing worn by an exposed individual. Furthermore, it is very difficult to accurately account for

the substantial variations in the energy and angular distributions of low-energy photons over the

body surface of an exposed individual located close to a source, with the result that there would

be considerable uncertainty in estimated doses at different depths in the body. Finally, the

energy and angular distributions of low-energy photons at the body surface, and thus the dose at

different depths in the body, would be quite sensitive to the assumed distance of an exposed

individual from a source.

Based on these considerations, calculations of external dose due to bremsstrahlung performed

using complex computer codes are unlikely to produce realistic results for low-energy betaemitting radionuclides. Therefore, this alternative is unlikely to have a substantial benefit

compared with an approach of applying simple reduction factors to dose estimates obtained

using CONDOS II.

A.4.1.6 Conclusion

A simple approach to addressing the likely overestimates of external dose due to bremsstrahlung

obtained using CONDOS II is adopted for use in this report. In this approach, simple

radionuclide-specific reduction factors are applied to calculated doses for the low-energy beta

emitters of concern. These reduction factors are given in Table A.4.2. The reduction factors for

14C and 147Pm are obtained directly from a comparison of doses calculated using CONDOS II

with the results given in Federal Guidance Report No. 12 for the same source configuration (see

Table A.4.1). For 63Ni, which has a beta endpoint energy less than 100 keV, the dose due to

bremsstrahlung is assumed to be zero (0) in Federal Guidance Report No. 12. In this case, dose

estimates obtained using CONDOS II are adjusted using the assumed reduction factor for 14C. 

The dose estimate obtained using this reduction factor should be more realistic but still

conservative. Finally, the external dose due to exposure to 3

H is assumed to be zero (0) in all

cases, due to the very low energies of all bremsstrahlung photons and the assumption in all

assessments that pathways of internal exposure would occur.

A.4.2 Generic Dose Modeling for Sources in Close Proximity to the Body

Several of the exposure scenarios involve the placement of the item/source in close proximity of

the body/skin surface. For example, several misuse scenarios call for the user to carry the

exempt item in the shirt pocket or coveralls over an extended time period. Other routine

exposure scenarios involve the item/source being in contact with the skin, such as for a watch

dial using tritiated paint.

The calculations of the effective dose equivalent (EDE) and the localized skin dose from a small

source in close proximity to the body is complicated by several factors. Radiation fluence will

decrease rapidly with increased distance from the source; electron attenuation by clothing

becomes critical for skin dose calculations. For the calculation of EDE, the body will not be

exposed to a uniform radiation field. Therefore, conventional methods for correlating fluence to

effective dose equivalent are not appropriate. Use of the 1 cm deep-dose equivalent, as

specified in 10 CFR 20, or the fluence-to-dose conversion factors of the International

Commission on Radiation Units and Measurements (ICRU) (ICRU 47) would yield unrealistic

estimates of the EDE. For calculation of the skin dose, placing the source in a pocket of clothing
A.4–7

will reduce the electron energy level and fluence. Assumptions regarding thickness and density

of clothing can have a significant effect on resultant skin dose calculations.

For the calculation of the localized skin dose, VARSKIN MOD2 (Computer Codes, Durham,

1992) provides a method for performing the calculations. Generic assumptions regarding

distances and shielding materials (cloth covering) are needed in the calculations for consistency. 

However, for the EDE, attempting to calculate doses to the different internal organs and applying

weighting factors becomes quite involved, requiring analytical modeling techniques that do not

lend themselves to simple applications like these. A simple method is needed to estimate the

EDE based on assumptions concerning location and tissue depth. Such a method would allow

the use of readily available radiation shielding approximation methods, such as MicroShield

(Computer Codes, Grove Engineering, 1996), for estimating the EDE.

An in-depth evaluation of the EDE for photon radiation sources external to the body was

performed by Reece, et. al., for the Electric Power Research Institute (EPRI) (Reece, 1993). 

This evaluation presented calculations of the EDE for point sources located on the external body

surface. Mathematical models of the adult male and female body were coupled with Monte Carlo

modeling of photon source and transport for the calculations. Doses to internal organs were

calculated for specified source-organ geometries; organ weighting factors were applied; and the

EDE was calculated by summing the weighted organ doses.

Using this method, the EDE was calculated with the point source at varying locations on the torso

of the body. The position of the source was incrementally increased circumferentially around the

body and vertically up from the base of the torso to 61 cm. The EDE was calculated for a total of

52 locations for a point source located on the torso of an adult male and adult female. For the

male, the highest EDE was for the source located at the front part of the torso at a 6 cm height

above the base, which resulted primarily from the dose to the gonads. For the female, the

highest calculated EDE was with the source located in front at a height of 61 cm, which resulted

primarily from the dose to the breast.

The methods employed in the EPRI study were very detailed, requiring extensive modeling and

computer application. For this study, a simpler technique was needed. The approach selected

was to establish an effective tissue depth that would allow the use of the point kernel radiation

shielding code MicroShield (Computer Codes, Grove Engineering, 1996) for estimating an EDE

for sources in close proximity to the body. Using MicroShield, the deep dose equivalent at a

10 cm depth in tissue was compared with the results from the EPRI study. The 10 cm depth was

considered representative of the typical depth to the radiosensitive organs, as used for

calculating the EDE. 

As evaluated in the EPRI study, three different photon energies were evaluated – 0.08, 0.3 and

1.0 MeV. For 0.08 MeV photons, calculations using MicroShield (Computer Codes, Grove

Engineering, 1996) for the deep dose equivalent were a factor of five greater than the maximum

calculated EDE in the EPRI study for a point source in contact with the torso of the body. For

0.3 MeV photons, MicroShield calculations were a factor of 3 greater and a factor of 2 greater for

1.0 MeV photons. Based on these comparisons, it was concluded that using MicroShield and

calculating the deep dose equivalent at a 10 cm tissue depth provided a reasonably conservative

approach for estimating the EDE for sources in close proximity to the body.
21 MicroShield calculations performed with a total distance of 11 cm, comprised of 1 cm of air

(source distance from body), 9 cm of tissue (shielding), and 1 cm of air (recognizing that the

deep dose equivalent fluence-to-dose factors are based on a 1 cm depth).

A.4–8

In applying this approximation method for the different exemptions, the source cannot be

expected to be in direct contact with the body on a continuous basis. Therefore, for most

applications, a 1 cm separation has been assumed from the source to the skin of the body. 

Additionally for the evaluation of the electron dose component, attenuation by typical clothing has

been included. The source is assumed to be placed in a clothing pocket having a thickness of

0.07 cm and a density of 0.4 g/cm3

, yielding an effective thickness of 0.028 g/cm2 which is a

value typical for cotton coveralls (Martz et al., 1986).

Thus, for the purposes of generic modeling the following assumptions have been made:

 source located in a pocket of clothing, typically 1 cm from the body surface, 

 clothing has a thickness of 0.07 cm and a density of 0.4 g/cm3

, yielding an effective

thickness of 0.028 g/cm2

, and

 effective dose equivalent is estimated by calculation of the deep dose equivalent using

MicroShield at a tissue depth of 10 cm,21
A.4–9

Table A.4.1 Comparison of External Doses Due to Bremsstrahlung for Exposure

to Contaminated Ground Surface Calculated Using the CONDOS II Code 

With External Dose Coefficients Given in Federal Guidance Report No. 12

External Dose Coefficient

(Sv/s per Bq/m2

)

Radionuclidea CONDOS II Federal Guidance Report No. 12b Ratioc

3

H (19 keV) 3.7×1021 0d —

63Ni (66 keV) 5.6×1020 0d —

14C (156 keV) 3.4×1019 1.6×1020 21

147Pm (225 keV) 5.1×1019 3.4×1020 15

36Cl (710 keV) 2.1×1018 6.7×1019 4.6

90Sr/90Y (2,284 keV)e 7.0×1018 5.6×1018 1.25

a

 Entry in parentheses is the endpoint energy of the continuous spectrum of electrons in beta

decay (Kocher, 1981). Radionuclides are listed in order of increasing beta endpoint energy.

b

 Dose coefficients given in Table III.3 of Federal Guidance Report No. 12

(EPA–402–R–93–081).

c

 Value obtained using CONDOS II divided by the value in Federal Guidance Report No. 12.

d

 Dose is assumed to be zero (0) when beta endpoint energy is less than 100 keV.

e

 Beta endpoint energy is the value for the short-lived decay product 90Y, which is assumed to be

in activity equilibrium with 90Sr; beta endpoint energy for 90Sr is 546 keV.
A.4–10

Table A.4.2 Dose Reduction Factors Applied to Estimates

of External Dose Due to Bremsstrahlung Obtained

Using the CONDOS II Code a

Radionuclide Dose Reduction Factor

3

H b

14C 20

63Ni 20

147Pm 15

a

 For the other beta-emitting radionuclides of concern to this report, correction of external doses

estimated using CONDOS II is unimportant, either because the correction factor is small for highenergy beta emitters or because the radionuclide also emits gamma rays or X-rays, or pathways

of internal exposure are assumed to occur.

b

 Dose due to bremsstrahlung is assumed to be zero (0), because the energies of the

bremsstrahlung photons are very low and pathways of internal exposure also are assumed to

occur.
APPENDIX B
B–1

GLOSSARY

bastnasite—a light lanthanide (Ln) fluoride carbonate that occurs in an unusual type of magmaderived deposit in which the Ln-elements have been enhanced. 

beta backscatter/transmission devices—devices that use particles from a variety of sources to

measure the thickness or density of thin films and thin coatings on other materials.

byproduct material—any radioactive material (except special nuclear material) yielded in or made

radioactive by exposure to radiation during the process of producing or utilizing special nuclear

material.

calibration source—a source of a known purity and activity that is used to determine the variation

in accuracy of a measuring instrument and to ascertain necessary correction factors.

chemical detectors—devices used to monitor for harmful or toxic gases and a variety of vapors.

collective dose—the sum of the individual doses received in a given period of time by a specified

population from exposure to a specified source of radiation.

decommission—to remove safely from service and reduce residual radioactivity to a level that

permits release of the property for unrestricted use and termination of license.

depleted uranium—the source material uranium in which the isotope uranium-235 is less than

0.711 percent by weight of the total uranium present. Depleted uranium does not include special

nuclear material.

disposal—isolation of radioactive wastes from the biosphere inhabited by man and containing his

food chains by emplacement in a land disposal facility.

dose—a general term denoting the quantity of radiation or energy absorbed per mass of tissue. 

For special purposes it must be appropriately qualified. If qualified, it refers to absorbed dose.

dose commitment—the total radiation dose to a part of the body that will result from retention in

the body of radioactive material. For purposes of estimating the dose commitment, it is assumed

that from the time of intake the period of exposure to retained material will not exceed 50 years. 

effective dose—the sum of the weighted equivalent doses in all the tissues and organs of the

body.

electron capture detectors for gas chromatographs—devices used to identify molecules in the

effluent stream from gas chromatographs.

exempt concentrations—some generally licensed items contain only small quantities of

byproduct material, and these items are potential candidates for exemption from licensing

requirements.
B–2

external dose—that portion of the dose equivalent received from radiation sources outside the

body.

fission—the splitting of a nucleus into at least two other nuclei and the release of a relatively large

amount of energy. Two or three neutrons are usually released during this type of transformation.

general license—the possession and use of specified quantities of certain radionuclides without

the need for applications or issuance of licensing documents to the particular persons using the

radioactive materials.

incandescent gas mantles—mantles containing thorium that are available in a variety of designs

and sizes, each intended to fit into one of the many different lighting devices in use, such as

camping lanterns, recreational vehicle lights, and outdoor gaslights.

incinerator workers—individuals who engage in sweeping or other cleanup activities while

located at the edge of a partially enclosed tipping area where garbage trucks unload waste at the

facility.

ion generating tube—devices designed for ionization of air that contains, as a sealed source or

sources, byproduct material consisting of a total of not more than 19 megabecquerel

(500 microcurie) of polonium-210 or 1.9 gigabecquerel (50 millicurie) of tritium per device.

internal dose—that portion of the dose equivalent received from radioactive material taken into

the body.

landfill workers—individuals who are located on top of the waste pile during operations and

perform such tasks as dumping of waste, grading of the waste following dumping, and covering

of the waste at periodic intervals.

liquid scintillation counters—devices that measure light emitted by a scintillator medium in which

radioactive materials are intimately dispersed and estimate the concentration of the radioactive

material from the light intensity.

maximum permissible concentration (MPC)—an acceptable upper limit for the concentration of a

specified radionuclide in a material taken into the body, below which continuous exposure to the

material will result in acceptable health risks to the specified population involved.

monazite—a rare earth phosphate, one of the most abundant rare earth minerals.

ore—a natural mineral compound of the elements of which at least one is a metal.

piezoelectric ceramics—used in many different shapes and sizes in consumer products that

require an electromechanical coupling device. Produced when pressure is applied to certain

classes of crystalline materials, where the crystalline structure produces a voltage proportional to

the pressure. Such consumer products include pacemakers, electronic telephone ringers,

microphones, patio grills, and games and toys.

rad—the unit of absorbed dose equal to 0.01 joule (J)/kg or 0.01 gray in any medium.
B–3

radioactive tracers—substances used to label specific atoms, molecules, living organisms, or

other entities. The tracer may be used to study the kinetics of exchange, distribution,

metabolism, turnover, conversion, and excretion of the labeled compound.

radiation monitoring—the measurement of radiation levels, concentrations, surface area

concentrations, or quantities of radioactive material and the use of the results of these

measurements to evaluate potential exposures and doses.

rem—a special unit of dose equivalent. The dose equivalent in rem is numerically equal to the

absorbed dose in rad multiplied by the quality factor, the distribution factor, and any other

necessary modifying factors.

sealed source—any licensed material that is encased in a capsule designed to prevent leakage

or escape of the licensed material.

self-luminous devices—devices that use byproduct material to create light without outside

activation. Examples are wristwatches, leveling bubbles, automobile lock illuminators, gun

sights, and aircraft and building exit signs.

sievert—the unit of radiation dose equivalent that is used for radiation protection purposes for

engineering design criteria and for legal and administrative purposes; equal to 1.0 joule (J)/kg.

source material—uranium or thorium, or any combination thereof, in any physical or chemical

form or ores that contain by weight one-20th of 1 percent (0.05%) or more of uranium, thorium, or

any combination thereof.

spark gap irradiators containing cobalt-60—irradiators designed to minimize spark delay in some

electrically ignited commercial fuel-oil burners by generating free electrons in the spark gap. 

These free electrons are produced from beta particles emanating from the cobalt-60 plating on

the irradiator.

special nuclear material—plutonium, uranium-233, uranium enriched in the isotope 233 or 235,

and any other material that the Nuclear Regulatory Commission, pursuant to the provisions of

section 51 of the Atomic Energy Act, determines to be special nuclear material, but does not

include source material; or any material that is artificially enriched by any of the foregoing but

does not include source material.

static eliminators—devices that work on the principle that static charges can be neutralized by

ionized particles. These devices use up to 200 millicurie of polonium-210 to ionize the air where

static charges may build up. These devices can be portable or stationary.

thermoluminescent dosimetry readers—devices used to determine the radiation dose to an

exposed piece of thermoluminescent material by measuring the light output as the material is

heated.

thorium vacuum tubes—vacuum tubes containing thoriated tungsten cathodes in many varied

designs.
B–4

tungsten inert-gas (TIG) arc welding—a process in which an electrical arc is struck between an

inert, gas-cooled, nonconsumable electrode (also called a welding rod), and the metal work

pieces. Many electrodes used in TIG welding consist of tungsten wire that contains thorium

dioxide or another metal oxide.

uranium milling—any activity that results in the production of byproduct material.

waste collectors—individuals who collect waste from the generating site, haul the waste to

garbage trucks, and transport the waste to landfills.

x-ray fluorescence analyzers—devices designed for use in nondestructive analysis to determine

the elemental chemical composition of solid and liquid samples.
B–5

SYMBOLS AND UNITS

Bq becquerel

C coulomb

Ci curie

cm centimeter

d day

g gram

Gy gray

h hour

kg kilogram

km kilometer

L liter

m meter

min minute

mg milligram

mL milliliter

mm millimeter

MT metric ton

ppb parts per billion

ppm parts per million

p-rem person-rem

R roentgen

s second

Sv sievert

E exa 1018

P peta 1015

T tera 1012

G giga 109

M mega 106

k kilo 103

c centi 102

m milli 103

micro 106

n nano 109

p pico 1012

f femto 1015
I–1

INDEX

abandoned mine .................................................... 3–63, 3–67

aggregate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–23, 3–30, 3–65, 3–155

aircraft engine part . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxv

aircraft safety devices . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–111

ALARA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1–5

area/process monitor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–145

asphalt ................................................................ 3–23

automobile glass . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–149, 3–181

automobile lock illuminator ................................................. 2–77

automobile shift quadrant .................................................. 2–98

baggage . . . . . . . . . . . . . . . . . . 2–17, 2–19, 2–20, 2–22, 2–40, 2–44, 2–45, R–20, R–21, A.3–6

balances of precision . . . . . . . . . . . . . . . . . . . xxx, xxxi, 2–87, 2–88, 2–90-93, 2–95, 2–106, 4–4

barium titanate ......................................................... 3–209

bastnasite . . . xxvii, 3–33, 3–147-150, 3–152-158, 3–162, 3–170, 3–171, 3–173, 3–174, 3–179,

3–181, 3–182

benchtop instrument ..................................................... 2–145

beryllium ............................................................... 3–75

boiler slag . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–22

borosilicate ............................................................ 2–195

bottom ash . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–22

bow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–196

bremsstrahlung . . . . . . 1–11, 2–50-52, 2–60, 2–63, 2–73, 2–120, 2–124, 2–127, 2–216, 2–224,

3–195, 3–219, 4–17, 4–19, 4–20, 4–31, 4–34, 4–37, 4–42, 4–54, 4–55, 4–59,

4–64, A.3–2, A.4–1-6, A.4–9, A.4–10

calibration source . . . . . . . . . . . . . . . . . 2–167, 2–172, 2–173, 2–178, 2–191, 3–86, 3–90, 4–64

casting . . . . . . . . . . . . . . . . . . . . . . . 3–23, 3–29, 3–44, 3–50, 3–51, 3–241-243, 3–302, 3–303

casting material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–23

ceramic or alumina bead .................................................. 2–161

ceramic tableware . . xxv, xxvii, xxx, xxxi, 3–191, 3–192, 3–194-201, 3–203, 3–204, 3–207, R–17

cerium . . . xxvii, 2–182, 3–75, 3–103, 3–147-151, 3–153, 3–155-163, 3–169-171, 3–173, 3–174,

3–176, 3–180-182, R–11

chemical detector . . . . . . . . . . . . . . . . . . xxiii, 2–218, 2–224, 2–226-228, 2–234, 2–236, 2–238

chlorinator bed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–50, 3–51

class Y material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–108, 3–113

cloisonne jewelry ........................................................ 3–232

coal ash . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–20, 3–22-24, 3–30, 3–31, 3–41, 3–52-56

communication equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–119

compost . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–23

contaminated steel . . . . . . . . . . . . . . . . . . . . . 2–19, 2–20, 2–24, 2–25, 2–27, 2–30, 2–33, 2–48

crushed base ........................................................... 3–23

cyanoacrylate .......................................................... 2–196

decolorization of glass . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–149

dental porcelain ...................................... 3–20, R–1, R–15, R–19, R–22

dental product ........................................................... 3–24

depleted uranium . xxvi, 3–2, 3–3, 3–5, 3–8, 3–19, 3–20, 3–40, 3–42, 3–43, 3–249-251, 3–258,

3–260, 3–269, 3–270, 3–274, 3–277, 3–279, 3–281-283, R–4, R–7, R–12, A.1–17,
I–2

A.1–19, A.1–22, A.2–33, A.2–35, A.2–37, A.2–39, A.2–41, A.2–44, A.2–46,

A.2–48, A.2–50, A.2–52, A.2–54, A.2–56, A.2–58, A.2–60, A.2–61, B–1

elastomeric . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–195

electric lamp . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–126

electromechanical coupling device ...................................... 3–209, B–2

electron tube . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–119-121, 2–123-127, 2–131, 2–137

explosive . . . . . . 2–16, 2–17, 2–19, 2–20, 2–22, 2–23, 2–25, 2–40, 2–44, 2–45, 4–4, 4–5, R–20,

R–21

fertilizer ................................................................ 3–23

Fiestaware® . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–192, 3–193

fluorescent lamp . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–123-127, A.1–9

fly ash . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–22, 3–30, 3–41

gas chromatograph . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxviii, 4–30, 4–36, 4–37, 4–40

gaseous tritium ................................................... 2–195, A.1–9

gemstone . . . . . . . . . . . . . . . . . . . . . . . . . . 2–17, 2–21, 2–22, 2–26, 2–46, 2–48, 3–163, 3–167

general license .... 1–10, 2–139, 2–162, 2–194, 3–233, 4–1, 4–3, 4–15, 4–18, 4–29-31, 4–41,

4–43, 4–53, 4–57, 4–58, 4–70, 4–71

glass enamel ................................................ 3–231, 3–232, R–18

glass enamel frit ............................................. 3–231, 3–232, R–18

glaze . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–165, 3–191, 3–193-196, 3–198-204, 3–207, R–4

gun sight . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–207-209, 2–213

hafnium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–184, 3–20, 3–22, 3–29, 3–44, 3–52, 3–53

handgun ......................................... 2–196, 2–206-209, 2–213, 2–214

handheld instrument ...................................................... 4–16

ilmenite . . . . . . . . . . . . . . . . . . . . . 3–27, 3–28, 3–32-35, 3–42, 3–44, 3–50, 3–51, 3–60, 3–148

incandescent lamp ................................................ 3–163, 3–167

insoluble oxide . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–50, 3–123, 3–135

internal calibration . . . . . . . . . . . . . . . . . . . . . . . . 2–139, 2–141-145, 2–147, 2–189, 4–54, 4–57

ion exchange resin ................................................ 2–140, 2–144

jewelry . . . 2–58, 2–64, 2–77, 2–78, 2–84, 2–85, 2–89, 2–90, 2–93, 2–98, 2–103, 2–114, 2–118,

3–231, 3–232, A.1–12

lead metaniobate ........................................................ 3–209

lead zirconate titanate .............................................. 3–209, 3–210

liquid crystal display . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–195

liquid scintillation counter ................................................... 4–53

lock illuminator ............................................. 2–74-78, 2–84, 2–85

magnesium-thorium alloy . . . . . . . . . . . . . . . . . xxx, xxxi, 3–239-243, 3–245, 3–248, 3–302-304

magnetron . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–95-98, 3–100, 3–239

marine compass . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–105, 2–108

medical glassware . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–149

microwave oven . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–95-102, R–19

microwave receiver protector tube . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–119, 2–120, 2–125-127

microwave tube . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . R–7

mineral sand . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–27, 3–35, 3–62, 3–65, 3–66, 3–68

monazite ...... 3–6, 3–7, 3–28, 3–29, 3–33, 3–41, 3–62-68, 3–147-152, 3–154, 3–155, 3–172,

3–177, 3–178, R–8

narcotic . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–218

navigational instrument ............................................. 2–105, 2–108

oil and gas production and process waste ...................................... 3–24
I–3

oil well . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–162

ophthalmic glass . xxvi, 3–20, 3–21, 3–26, 3–32, 3–35, 3–38, 3–59, 3–60, 3–150, 3–153, 3–154,

3–156, 3–159, 3–162, 3–181, 3–182, R–2

ore-buying station . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–61, 3–69

paving . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–31, 3–36

personnel neutron dosimeter ......................................... 3–188, 3–190

phosphate ore ................................................. 3–23, 3–31, 3–47

phosphate slag .... xxvi, 3–20, 3–23, 3–24, 3–30-32, 3–36, 3–37, 3–47, 3–52-55, 3–57, 3–59,

3–60

phosphogypsum ................................................ 3–23, 3–47, R–5

photographic camera ........................... xxv, 3–287, 3–290, 3–292-295, 3–299

photographic film ..................................... xxx, xxxi, 1–3, 3–233, 4–4, 4–6

piezoelectric ceramic . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxx, xxxi, 3–209-212, 3–214

pointer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–114, 2–115, 2–118

policy statement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1–2

porcelain teeth . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–20, R–1

pyrex . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–195, 2–196, 3–219

Pyrex® . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–194

quantities of byproduct material . . . . xxiii, 2–2, 2–17, 2–33, 2–35-39, 2–121, 2–123, 2–133-136,

2–139, 2–142, 2–147, 2–163-180, 2–182-188, 2–190, B–1

radar transmit-receiver switch .............................................. 2–120

radiation measuring instrument . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–141

radiation-resistant glass . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–149

radio broadcasting ........................................................ 3–98

railroad ballast ........................................................... 3–23

rare earth . . . xxiv, xxvii, xxx, xxxi, 3–2, 3–22, 3–26, 3–34, 3–50, 3–62, 3–147-159, 3–161-171,

3–177, 3–178, 3–181, R–10, B–2

refrigerant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–165

resin-coated sand . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–161

road construction .................................................... 3–22, 3–37

rockwool insulation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–23

rutile . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–42, 3–44, 3–50, 3–51

sapphire .............................................................. 2–196

scandium ....................... 2–161, 2–162, 2–186, 3–147, 3–149, 3–166, R–3, R–8

scintillation detector ...................................................... 2–140

scrap metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–241, 3–244, 3–245, 3–248, 3–306-308

sealed tube . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–195

self-luminous ..... xxx, xxxi, 1–3, 1–6, 1–9, 2–58, 2–59, 2–73, 2–74, 2–87, 2–95, 2–96, 2–105,

2–111, 2–193-197, 2–199-203, 2–205-214, 2–216, 3–166, B–3

self-luminous paint ............................. 2–73, 2–74, 2–87, 2–95, 2–96, 2–111

shipping container . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–269, 3–272, 3–273, 4–15, 4–29

silicon semiconductor ....................... 2–16-18, 2–22-24, 2–28, 2–31, 2–34, 2–48

smelting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–45, 3–50, 3–217, 3–306, A.2–5, A.2–28

smoke detector . . . . . . 2–217, 2–219-223, 2–227, 2–229, 2–233, 3–310, 3–312, R–17, A.1–5-7

soil stabilization .......................................................... 3–23

solvent . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–45, 3–156, 3–174

sources for calibration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–164, 4–53

spark gap irradiator . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxiv, 2–151

spark gap tube . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–127
I–4

special application metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–29, 3–30

sports arena . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–138-140, 3–143, 3–145

stabilization material ...................................................... 3–23

static eliminator . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4–4-7, 4–10, 4–13

storage pad . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–67

structural fill . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–22

sulfur hexafluoride . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–218

sunlamp . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxx, 3–139, 3–140, 3–143, 3–145

tape dispenser ................................. 3–27, 3–29, 3–35, 3–36, 3–59, 3–60

television camera . . . . . . . . . . . . . . . . . . . . . . . . xxv, 3–286, 3–289, 3–290, 3–292-295, 3–298

thermostat dial . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–114, 2–115, 2–118

thorium oxide . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–23, 3–41, 3–75, 3–76, 3–105, 3–240, 3–285

timepiece . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–49-53, 2–58-60, 2–63-66, 2–69

tin . . . . . . . . . . . . . . . . . . . . 2–12, 2–188, 3–29, 3–41, 3–50, 3–52, 3–53, 3–148, 3–210, 3–231

titanium . . . . 3–20, 3–22, 3–29, 3–42, 3–44, 3–45, 3–50-53, 3–62, 3–148, 3–164, 3–167, 4–30,

4–31, 4–38

topaz . . . . . . . . . . . . . . . . . . . . . . . . . xxiii, 2–16, 2–17, 2–21-28, 2–34, 2–35, 2–46, 2–48, R–18

tracer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–161, 2–162, 3–166, B–3

tritiated paint . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–55, 2–59, 2–111, A.4–6

tritium gas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2–105, R–17

tungsten-thorium alloy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–241-244, 3–248, 3–305

ultraviolet absorption glass . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–149

unprocessed ore ........................... xxv, xxx, xxxi, 3–1, 3–61, 3–62, 3–66, 3–69

uranium mine . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–63, 3–65, 3–67, 3–68, 3–70, R–3

uranium oxide . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–41, 3–191, 3–210, 3–217, 3–231

vacuum tube . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–100-102, R–20

watch . . . . . . 1–3, 2–49-53, 2–56-60, 2–62-64, 2–69, 2–77, 2–89, 2–98, 2–114, 2–127, 2–195,

2–197-205, 2–209-212, 3–100, 3–166, R–23, A.1–12, A.1–23, A.4–6

water treatment sludge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–20, 3–23, 3–30, 3–31

welding rod . . . . . . . . . . . . . . . . 3–23, 3–103, 3–111, 3–112, 3–114, 3–115, 3–120, 3–121, B–4

wristwatch . . . . . . . . . . . . . . . . . . . 2–49, 2–50, 2–55-58, 2–195, 2–198-203, 2–206, 2–210-212

xenotime ........................................................ 3–147, 3–149

x-ray fluorescence analyzer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxix, 4–44

yttrium . . . . . . . . . . . . . . . . . . . . . . . . . 2–188, 3–75, 3–103, 3–147, 3–149, 3–150, 3–167, R–8

zircon flour . . . . . . . . . . . . . . . . . 3–23, 3–27, 3–29, 3–32, 3–34, 3–35, 3–37, 3–38, 3–59, 3–60

zircon sand . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3–23, 3–27-29, 3–32-36, 3–38, 3–58, R–3

zirconium . . . . 2–188, 3–20-23, 3–26, 3–27, 3–29, 3–32, 3–36, 3–44, 3–45, 3–52, 3–53, 3–62,

3–103, 3–148, 3–154, 3–159, 3–162, 3–210, R–12, R–21