ORAU Team

Dose Reconstruction Project for NIOSH

Savannah River Site

Document Number:

ORAUT-TKBS-0003

Effective Date: 04/05/2005

Revision No.: 03

Controlled Copy No.: ________

Page 1 of 248

Subject Expert: Edward D. Scalsky

Document Owner

Approval: Signature on File Date: 03/31/2005 Edward D. Scalsky, TBD Team Leader

Approval: Signature on File Date: 03/31/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 04/01/2005

Richard E. Toohey, Project Director

Approval: Signature on File Date: 04/05/2005

James W. Neton, Associate Director for Science

Supersedes:

Revision No.: 02

TABLE OF CONTENTS

Section

Record of Issue/Revisions .................................................................................................................

Page

11

Acronyms and Abbreviations ............................................................................................................. 13

Executive Summary........................................................................................................................... 16

1.0 Introduction ................................................................................................................................ 18

1.1 Purpose............................................................................................................................ 18

1.2 Scope............................................................................................................................... 18

1.3 Site Activities and Processes............................................................................................ 19

2.0 Occupational Medical Dose........................................................................................................ 40

2.1 Examination Frequencies................................................................................................. 40

2.2 Equipment and Techniques.............................................................................................. 41

2.2.1 X-Ray Apparatus ........................................................................................................ 41

2.2.2 Collimation.................................................................................................................. 42

2.3 Organ Dose Calculations PA and lateral(LAT).................................................................. 42

2.3.1 Organ Dose Calculation Methods – Chest Photofluorography .................................... 46

2.4 Uncertainty....................................................................................................................... 47

2.5 Dose Reconstructor Instructions....................................................................................... 48

2.5.1 Assignment of Organ Doses from X-ray Procedures: Maximizing 

Approach for Dose Reconstructions ........................................................................... 49

2.5.2 Assignment of Organ Doses from X-Ray Procedures: Best Estimate 

Approach for Dose Reconstructions ........................................................................... 49

2.6 Reconstruction of Organ Dose From Radiography Equipment Using Minimal 

Collimation ....................................................................................................................... 49
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3.0 Occupational Environmental Dose.............................................................................................. 51

3.1 Internal Dose from On-Site Atmospheric Radionuclide Concentrations ............................ 51

3.1.1 On-Site Releases to Air .............................................................................................. 51

3.1.2 Annual Intakes from Resuspension ............................................................................ 58

3.1.3 Annual Doses from the Ingestion of Foodstuffs .......................................................... 59

3.2 Occupational External Dose ............................................................................................. 59

3.2.1 Ambient Radiation ...................................................................................................... 59

3.2.2 Releases of Noble Gases ........................................................................................... 60

3.3 Uncertainty....................................................................................................................... 60

3.4 Dose Reconstructor Instructions....................................................................................... 61

3.4.1 Assignment of Environmental (On-Site Ambient) Doses: Maximizing 

Approach for Dose Reconstructions ........................................................................... 61

3.4.2 Assignment of Environmental (On-Site Ambient) Doses: Best Estimate 

Approach for Dose Reconstructions ........................................................................... 62

4.0 Occupational Internal Dose ........................................................................................................ 64

4.1 In Vitro Minimum Detectable Activities (MDAS), Counting Methods ................................. 64

4.1.1 In Vitro Urine Analysis ................................................................................................ 65

4.1.2 In Vitro Methods for Individual Radionuclides ............................................................. 65

4.1.3 Correcting for Urinalysis Volume ................................................................................ 71

4.1.4 Fecal Sample Analysis ............................................................................................... 71

4.2 In Vivo MDAS, Counting Methods, and Reporting Practices............................................. 71

4.2.1 Whole Body Counting................................................................................................. 72

4.2.2 Chest Counting........................................................................................................... 73

4.3 Personal Air Sampling Data as Applicable ....................................................................... 75

4.4 Interferences, Uncertainty, and Use of Reporting Levels .................................................. 75

4.4.1 Contamination of Samples.......................................................................................... 75

4.4.2 Use of Reporting Levels and Less-Than Values ......................................................... 76

4.4.3 Unmonitored Worker .................................................................................................. 85

4.5 Maximizing Dose.............................................................................................................. 86

4.5.1 Assignment of Internal Doses (from Radionuclides Other than Tritium): 

Maximizing Approach for Dose Reconstructions......................................................... 86

4.5.2 Assignment of Internal Doses from Tritium: Maximizing Approach for 

Dose Reconstructions ................................................................................................ 87

5.0 Occupational External Dosimetry ............................................................................................... 88

5.1 Introduction ...................................................................................................................... 88

5.2 Basis of Comparison ........................................................................................................ 88

5.3 Dose Reconstruction Parameters..................................................................................... 89

5.3.1 SRS Historical Administrative Practices...................................................................... 89

5.3.2 SRS Dosimetry Technology........................................................................................ 91

5.3.2.1 Beta/Photon Dosimeters .................................................................................... 92

5.3.2.2 Neutron Dosimeters........................................................................................... 95

5.3.3 Calibration .................................................................................................................. 96

5.3.3.1 SRS Beta/Photon Dosimeters............................................................................ 96

5.3.3.2 SRS Neutron Dosimeters................................................................................... 96

5.3.4 Workplace Radiation Fields ........................................................................................ 98

5.3.4.1 SRS Workplace Beta/Photon Dosimeter Response ........................................... 98

5.3.4.2 SRS Workplace Neutron Dosimeter Response.................................................. 98

5.3.4.2.1 Plutonium Facilities ................................................................................... 98

5.3.4.2.1.1 Neutron Energy Spectra ................................................................... 99
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5.3.4.2.1.2 Neutron to Photon Ratio ................................................................. 102

5.3.4.2.2 Neutron Source Production Facilities (773-A).......................................... 104

5.3.4.2.2.1 Neutron Energy Spectra ................................................................. 104

5.3.4.2.2.2 Neutron to Photon Ratio ................................................................. 105

5.3.4.2.3 Reactor Facilities (100 Area)................................................................... 105

5.3.4.2.3.1 Neutron Energy .............................................................................. 105

5.3.4.2.3.2 Neutron-to-Photon Ratio................................................................. 106

5.3.4.2.4 Other Facilities ........................................................................................ 107

5.3.4.2.4.1 Calibration Facilities (736-A Area) .................................................. 107

5.3.4.2.4.2 Fuel Fabrication Area (321-M)........................................................ 108

5.3.5 SRS Workplace Recorded Dose Uncertainty............................................................ 108

5.4 Adjustments to Recorded Dose...................................................................................... 110

5.4.1 Photon Dose Adjustments ........................................................................................ 110

5.4.2 Neutron Dose Adjustments....................................................................................... 110

5.4.2.1 Neutron Weighting Factor ................................................................................ 110

5.4.2.2 Neutron Correction Factor................................................................................ 111

5.5 Missed Dose .................................................................................................................. 112

5.5.1 Photon Missed Dose ................................................................................................ 112

5.5.1.1 Facility/Location ............................................................................................... 113

5.5.1.2 Dosimeter Type ............................................................................................... 113

5.5.1.3 Year................................................................................................................. 114

5.5.1.4 Energy Range.................................................................................................. 114

5.5.2 Neutron Missed Dose ............................................................................................... 114

5.5.2.1 Prior to 1971 .................................................................................................... 115

5.5.2.2 Post-1971 ........................................................................................................ 115

5.6 Organ Dose.................................................................................................................... 115

5.6.1 Organ Dose Conversion Factors .............................................................................. 116

5.7 Dose Reconstructor Instructions..................................................................................... 117

5.7.1 Assignment of External Doses from Monitoring Data: Maximizing 

Approach for Dose Reconstructions ......................................................................... 117

5.7.2 Assignment of External Doses from Monitoring Data: Best Estimate 

Approach for Dose Reconstructions ......................................................................... 117

6.0 Tradesworkers.......................................................................................................................... 117

References ...................................................................................................................................... 118

Glossary .......................................................................................................................................... 130

A. Description of Facilities And Processes....................................................................................

Attachments

138

A.1 Reactors......................................................................................................................... 140

A.2 F-Area A-Line Facility..................................................................................................... 140

A.3 221-F B-Line Facility ...................................................................................................... 140

A.4 221-F Canyon Facility .................................................................................................... 140

A.5 New Special Recovery Facility ....................................................................................... 141

A.6 Area Outside Facilities ................................................................................................... 141

A.7 238PuO2 Fuel Form Facility (PuFF) and the 238PuO2 Experimental Facility 

(PEF).............................................................................................................................. 146

A.8 235-F Vaults................................................................................................................... 146
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A.9 A-9 772-F and 772-1F Production Control Laboratories ................................................. 146

A.10 E-Area Solid Waste Disposal Facility (SWDF)................................................................ 147

A.11 Plutonium Storage Facility (PSF).................................................................................... 147

A.12 F/H Effluent Treatment Facility (ETF) ............................................................................. 147

A.13 F & H Cooling Water and Retention Basins.................................................................... 148

A.14 F & H Area Tank Farms.................................................................................................. 148

A.15 Waste Certification Facility (WCF).................................................................................. 148

A.16 221-H B-Line Facility ...................................................................................................... 150

A.17 H-Canyon Facility........................................................................................................... 150

A.18 211-H Area Outside Facilities........................................................................................ 153

A.19 Receiving Basin for Off-Site Fuel (RBOF) and Resin Regeneration Facility 

(RRF) ............................................................................................................................. 153

A.20 Tritium Facilities ............................................................................................................. 153

A.21 M-Area ........................................................................................................................... 153

A.22 320-M, 322-M, and 341-M.............................................................................................. 155

A.23 Fuel Fabrication Facility, 321-M...................................................................................... 155

A.24 S-Area- Defense Waste Processing Facility (DWPF) ..................................................... 155

A.25 Z Area ............................................................................................................................ 156

A.26 A Area Building 773-A.................................................................................................... 156

A.27 Buildings 735-A and 735-11A......................................................................................... 158

A.28 Building 776-A Liquid Waste Handling Facility................................................................ 158

A.29 D-Area Heavy Water Production and Reprocessing....................................................... 158

A.30 CMX and TNX (TNX Semi-works) .................................................................................. 158

B. Occupational Medical Dose......................................................................................................159

B.0 SRS Medical X-Ray.................................................................................................................. 159

B.1 Organ Doses from Medical X-Rays ................................................................................ 159

C. Occupational Environmental Dose............................................................................................163

C.0 Occupational Environmental Dose............................................................................................ 164

C.1 Internal Dose from On-Site Concentrations of Radionuclides ......................................... 164

C.1.1 On-Site Releases to Air ............................................................................................ 164

C.1.2 Resuspension of Radioactive Materials from Soil ..................................................... 164

C.2 External Dose................................................................................................................. 164

C.2.1 Ambient Radiation .................................................................................................... 164

C.2.2 Releases of Noble Gases ......................................................................................... 164

C.3 Maximum Plausible Annual Doses From Ingestion of Foodstuffs at the 

Savannah River Site....................................................................................................... 188

C.3.1 Introduction............................................................................................................... 188

D. Occupational Internal Dose for Monitored Workers................................................................... 188

D.0 Occupational Internal Dose ...................................................................................................... 195

D.1 In Vitro Minimum Detectable Activities (MDAs) and Reporting Levels ............................ 195

D.1.1 In Vitro Urine Analysis .............................................................................................. 195

D.1.2 In Vitro Urine Analyses for Individual Radionuclides ................................................. 196

D.1.3 In Vitro Fecal Analysis .............................................................................................. 197

D.2 In Vivo MDAs and Reporting Practices at SRS............................................................... 197

D.2.1 Whole Body Counting............................................................................................... 197
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D.2.2 Chest Counting......................................................................................................... 198

D.3 Tables of Chronic Intakes of Various Fission Product and Activation Product 

Radionuclides Based on MDA Whole Body Count on Last Day of the Period................. 198

E. Occupational External Dose for Monitored Workers ................................................................. 240

E.0 Occupational External Dose ..................................................................................................... 241

E.1 Years of Exposure.......................................................................................................... 241

E.2 Rate ............................................................................................................................... 241

E.3 Radiation Type ............................................................................................................... 241

E.3.1 Beta/Photon.............................................................................................................. 241

E.3.2 Neutron .................................................................................................................... 241

E.4 Distribution Parameters.................................................................................................. 242

E.4.1 Parameter #1............................................................................................................ 242

E.4.1.1 Resolution of Photon, Neutron and Tritium Dose ............................................. 243

E.4.1.2 Adjustments to Recorded Photon Dose ........................................................... 243

E.4.1.3 Adjustments to Recorded Neutron Dose.......................................................... 244

E.4.1.4 Unmonitored Photon Dose............................................................................... 244

E.4.1.5 Missed Photon Dose........................................................................................ 244

E.4.1.5.1 Facility/Location ...................................................................................... 244

E.4.1.5.2 Dosimeter Type....................................................................................... 245

E.4.1.5.3 Year ........................................................................................................ 246

E.4.1.5.4 Energy Range ......................................................................................... 246

E.4.1.6 Missed Neutron Dose Prior to 1971 ................................................................. 246

E.4.1.7 Missed Neutron Dose After 1970 ..................................................................... 247

E.4.1.8 Organ Dose Equivalent.................................................................................... 247

E.4.2 Parameter #2............................................................................................................ 248

E.4.3 Organ Dose Conversion Factors .............................................................................. 248
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LIST OF TABLES

Table

2.1 Frequency of occupational chest x-rays at the SRS....................................................

Page

40

2.2-1 Description of the X-ray equipment used at SRS........................................................ 41

2.2-2 Technique factors used for each type of X-ray equipment .......................................... 42

2.3-1 Analogues for IREP organs not included in ICRP 34 .................................................. 47

2.6 Substitute dose conversion factors ............................................................................. 50

3.1.1-1 Distance, in meters, from source areas to population centers at SRS......................... 53

3.1.1-2 Overview of tritium monitors at SRS........................................................................... 55

3.1.1-3 Overview of radioiodine sampling and monitoring methods ........................................ 56

3.1.2-1 Radionuclide soil concentrations, pCi g-1, (1 sigma) dry weight (0-8 cm 

depth)......................................................................................................................... 59

3.4-1 On-site ambient organ doses...................................................................................... 63

4.1.1-1 Pu-239/240 – in vitro urinalysis................................................................................... 66

4.1.1-2 Pu-238 – in vitro urinalysis.......................................................................................... 66

4.1.1-3 Activity composition for reference 6% and 12% 240Pu mixtures................................... 67

4.1.1-4 Uranium urinalysis ...................................................................................................... 68

4.2.1-1 Whole body counting .................................................................................................. 72

4.2.1-2 Current MDAs for whole body counting ...................................................................... 72

4.2.1-3 Mean body burdens of 137Cs from fallout in the United States..................................... 73

4.2.2-1 Current MDAs for chest counting (2.5 cm chest wall thickness).................................. 74

4.2.2-2 Current MDAs for chest counting 50th percentile and 95th percentile male

workers....................................................................................................................... 74

4.4.2-1 239Pu chronic inhalation intake assessment based on MDA urinalysis on last 

day of the period......................................................................................................... 76

4.4.2-2 237Np chronic inhalation intake assessment based on MDA urinalysis on last

day of the period......................................................................................................... 77

4.4.2-3 234U chronic inhalation intake assessment based on MDA urinalysis on last 

day of the period......................................................................................................... 77

4.4.2-4 90Sr chronic inhalation intake assessment based on MDA urinalysis on last 

day of the period......................................................................................................... 80

4.4.2-5 144Ce chronic inhalation intake assessment based on MDA urinalysis on last 

day of the period......................................................................................................... 80

4.4.2-6 Parameters for simplified upper bound dose from fission/activation products ............. 82

4.4.2-7 Parameters for reconstructing upper bound dose from fission/activation 

products ..................................................................................................................... 82

4.4.2-8 137Cs chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ............................................................................................. 83

4.4.2-9 144Ce chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ............................................................................................. 83

4.4.2-10 106Ru chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ............................................................................................. 83

4.4.2-11 60Co chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ............................................................................................. 84

4.5.1-1 Doses to non-metabolic and digestive tract organs..................................................... 86

4.5.2-1 Tritium dose................................................................................................................ 87

5.3.1-1 SRS dosimeter type, period of use, exchange frequency, MDL and potential 

annual missed dose.................................................................................................... 91

5.3.2-1 SRS historical dosimetry events ................................................................................. 93

5.3.2.1-1 IARC testing results for US beta/photon dosimeters ................................................... 94
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5.3.2.1-2 Testing results for Hanford two-element and multi-element film dosimeters for 

energy and angular response ..................................................................................... 95

5.3.3.2-1 Common sources of laboratory bias for neutron dosimeter calibration 

parameters. ................................................................................................................ 97

5.3.3.1-1 Common sources of laboratory bias for beta/photon dosimeter calibration 

parameters ................................................................................................................. 97

5.3.4.1-1 Selection of beta and photon radiation energies and percentages for SRS 

facilities ...................................................................................................................... 99

5.3.4.1-2 Common workplace beta/photon dosimeter Hp(10) performance.............................. 100

5.3.4.2-1 Common workplace neutron dosimeter Hp(10) performance.................................... 100

5.3.4.2.1.1-1 Measured dose fractions at different locations in H Area .......................................... 102

5.3.4.2.1.1-2 Measured dose fractions at different locations in F Area........................................... 102

5.3.4.2.1.2-1 Recording periods for selected Hanford plutonium workers ...................................... 103

5.3.4.2.1.2-2 Hanford ratio of recorded dose components............................................................. 104

5.3.4.2.1.1-3 Post-1971 SRS ratio of FB and HB Line recorded dose components ....................... 104

5.3.4.2.2.1-1 Dose fractions for various operations in 773-A area ................................................. 105

5.3.4.2.2.2-1 Post-1971 SRS ratio of 773-A recorded dose components....................................... 105

5.3.4.2.3.1-1 Dose fractions for heavy water reactors at SRS ....................................................... 106

5.3.4.2.3.2-1 Post-1971 SRS ratio of plutonium and reactor recorded dose components .............. 107

5.3.4.2.4.1-1 Dose fractions at the calibration facility (736-A)........................................................ 107

5.3.5-1 Estimates of uncertainty ........................................................................................... 109

5.3.5-2 Common workplace neutron dosimeter Hp(10) performance.................................... 109

5.4.1-1 Adjustments to reported SRS deep photon dose ...................................................... 110

5.4.2-1 Historical neutron quality or weighting factors........................................................... 111

5.4.2.2-1 Summary of dose fractions and associated ICRP 60 correction factors .................... 112

5.5.1-1 Missed photon dose according to SRS facility .......................................................... 113

5.5.1-2 Missed photon dose according to dosimeter type ..................................................... 113

5.5.1-3 Missed photon dose adjustments to recorded deep dose ......................................... 114

5.5.2-1 Summary of SRS neutron dosimeter limits of detection ............................................ 114

5.5.2-2 Summary of neutron to photon ratios by process...................................................... 116

5.6.1-1 Default exposure geometries to calculate organ dose .............................................. 116

A-1.1 Reactors years of operation...................................................................................... 140

A-1.2 Radionuclides of concern for all reactors .................................................................. 140

A-2 F Area A-Line Facility ............................................................................................... 141

A-3 221-F B-Line Facility................................................................................................. 141

A-4.1 221-F Canyon Facility............................................................................................... 142

A-4.2 221-F Canyon Facility............................................................................................... 142

A-4.3 221-F Canyon Facility............................................................................................... 143

A-5 New Special Recovery (NSR) Facility....................................................................... 143

A-6.1 F-Area Outside Facilities .......................................................................................... 144

A-6.2 F Area Outside Facilities........................................................................................... 144

A-6.3 F Area Outside Facilities........................................................................................... 144

A-6.4 F Area Outside Facilities........................................................................................... 145

A-6.5 F Area Outside Facilities........................................................................................... 145

A-6.6 F Area Outside Facilities........................................................................................... 146

A-7 238PuO2 Fuel Form Facility (PuFF) and 238PuO2 Experimental Facility (PEF) ............ 146

A-8 235-F Vaults............................................................................................................. 147

A-9 772-F and 772-1F production control laboratories .................................................... 147

A-10 E-Area Solid Waste Disposal Facility (SWDF).......................................................... 147

A-11 Plutonium Storage Facility ........................................................................................ 148

A-12 F/H Effluent Treatment Facility ................................................................................. 148
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A-13 F & H Area—cooling water and retention basins ...................................................... 149

A-14 F & H Area Tank Farms............................................................................................ 149

A-15 Waste Certification Facility ....................................................................................... 149

A-16.1 221-H B-Line Facility ................................................................................................ 150

A-16.2 221-H B-Line Facility ................................................................................................ 150

A-16.3 221-H B-Line Facility, Np02 and Pu02 Facility ........................................................... 150

A-17.1 H-Canyon Facility ..................................................................................................... 151

A-17.2 H-Canyon Facility ..................................................................................................... 151

A-17.3 H-Canyon Facility ..................................................................................................... 152

A-17.4 H-Canyon Facility ..................................................................................................... 152

A-18.1 H-Area Outside Facilities .......................................................................................... 153

A-18.2 H-Area Outside Facilities .......................................................................................... 153

A-19.1 Receiving Basin for Off-site Fuel (RBOF) ................................................................. 154

A-19.2 Resin Regeneration Facility...................................................................................... 154

A-20 Tritium Facilities........................................................................................................ 154

A-21 Uranium Target Fabrication Facility, 313 M .............................................................. 154

A-22 Uranium Processing Facilities 320-M, 322-M, and 341-M......................................... 155

A-23 Fuel Fabrication Facility, 321-M................................................................................ 155

A-24.1 S Area Defense Waste Processing Facility............................................................... 156

A-24.2 S Area Defense Waste Processing Facility............................................................... 156

A-25.1 Z Area ...................................................................................................................... 157

A-25.2 Z Area ...................................................................................................................... 157

B-1 Average absorbed dose from PA & LAT chest procedures (mGy) for selected 

organs for 1 Gy entrance kerma (air kerma without backscatter) for a beam 

quality of 2.5 mm Al HVL and 3.5 mm Aluminum (HVL)............................................ 160

B-2 Organ doses for Savannah River Site for PA & LAT chest radiography .................... 161

B-3 Average absorbed dose from chest photofluorography (mGy) for selected 

organs for 1 Gy entrance kerma (air kerma without backscatter) for a beam 

quality of 2.5 mm aluminum (HVL)............................................................................ 162

B-4 Organ dose estimates for SRS Photofluorography July 1, 1954 to 1/31/67............... 162

C-1 A Area annual intakes and geometric standard deviations (Bq per year).................. 165

C-2 C Reactor area annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 166

C-3 D Area annual intakes and geometric standard deviations (Bq per year).................. 167

C-4 E Area annual intakes and geometric standard deviations (Bq per year).................. 168

C-5 F Area annual intakes and geometric standard deviations (Bq per year) .................. 169

C-6 H Area annual intakes and geometric standard deviations (Bq per year).................. 170

C-7 S Area annual intakes and geometric standard deviations (Bq per year).................. 171

C-8 Z Area annual intakes and geometric standard deviations (Bq per year) .................. 172

C-9 Forest Service Headquarters calculated air concentration (Bq per m3) .................... 173

C-10 Central Shops annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 174

C-11 L Reactor Area annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 175

C-12 P Reactor Area annual intakes and geometric standard deviations (Bq per

year)......................................................................................................................... 176

C-13 K Reactor Area annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 177

C-14 R Reactor Area annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 178

C-15 Rail Yard annual intakes and geometric standard deviations (Bq per year) .............. 179
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C-16 Average site-wide annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 180

C-17 Maximum site-wide annual intakes and geometric standard deviations (Bq 

per year)................................................................................................................... 181

C-18 Annual intakes for F and H Areas (Bq per year) ....................................................... 182

C-19 Ambient radiation levels for selected SRS locations (mrem/year)............................. 182

C-20 Argon-41 50th-percentile dose to skin and bone surfaces resulting from 

reactor releases (mrem per year) and geometric standard deviation ........................ 184

C-21 Argon-41 50th-percentile dose to whole body and other organs resulting from 

reactor releases (mrem per year) and geometric standard deviation ........................ 186

C-22 Radionuclide soil concentrations, pCi/g, (1 sigma) dry weight (0-8 cm depth)........... 188

C-23 Transfer factors for leafy vegetables (Bq/kg leafy vegetables per Bq/kg soil)............ 189

C-24 Transfer factors for fruit (Bq/kg fruit per Bq/kg soil)................................................... 189

C-25 Transfer factors for root vegetables (Bq/kg root vegetable per Bq/kg soil)................ 190

C-26 Parameters used for calculation of 50-year equivalent dose..................................... 190

C-27 Maximum HTO concentration measured in onsite vegetation at the SRS................. 191

C-28 50-year equivalent organ doses, mrem, based on 60 kilogram per year 

consumption............................................................................................................. 191

D-1 Limits of detection for urinalysis................................................................................ 196

D-2 MDAs for whole body counting ................................................................................. 198

D-3 MDAs for chest counting........................................................................................... 199

D-4 95Zr chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 199

D-5 95Nb chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 200

D-6 54Mn chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 200

D-7 131I chronic inhalation intake assessment based on MDA whole body count on 

last day of the period ................................................................................................ 201

D-8 154Eu chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 201

D-9 65Zn chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 202

D-10 140Ba chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 202

D-11 Ba-140, Type F, f1=0.1, 50-year equivalent dose for years of chronic intake 

(Sv) MDA = 9.3 nCi.................................................................................................. 204

D-12 Ce-144, Type W, 50-year equivalent dose for years of chronic intake (Sv) 

(MDA = 29 nCi)......................................................................................................... 206

D-13 Co-60, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 1 nCi............................................................................................................. 208

D-14 Co-60, Type S, 50-year equivalent dose for years of chronic intake (Sv) MDA 

= 1 nCi...................................................................................................................... 210

D-15 Cs-137, Type F 50-year equivalent dose for years of chronic intake (Sv) 

(MDA = 1 nCi)........................................................................................................... 212

D-16 Eu-154, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 2.4 nCi.......................................................................................................... 214

D-17 I-131, Type F, 50-year equivalent dose for years of chronic intake (Sv) MDA 

= 1.4 nCi................................................................................................................... 216

D-18 La-140, Type F, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 9.3 nCi.......................................................................................................... 218
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D-19 La-140, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 9.3 nCi.......................................................................................................... 220

D-20 Mn-54, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 1 nCi............................................................................................................. 222

D-21 Nb-95, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 2.2 nCi.......................................................................................................... 224

D-22 Nb-95, Type W, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 2.2 nCi.......................................................................................................... 226

D-23 Nb-95, Type S, 50-year equivalent dose for years of chronic intake (Sv) (MDA 

= 2.2 nCi).................................................................................................................. 228

D-24 Ru-106, Type F, f1=0.05, 50-year equivalent dose for years of chronic intake 

(Sv) MDA = 1 nCi .................................................................................................... 230

D-25 Ru-106, Type S, f1=0.05, 50-year equivalent dose for years of chronic intake 

(Sv) MDA = 1 nCi .................................................................................................... 232

D-26 Zn-65, Type S, f1=0.49999, 50-year equivalent dose for years of chronic 

intake (Sv) MDA = 5.1 nCi ....................................................................................... 234

D-27 Zr-95, Type M, 50-year equivalent dose for years of chronic intake (Sv) MDA 

= 2.2 nCi................................................................................................................... 236

D-28 Zr-95, Type S, 50-year equivalent dose for years of chronic intake (Sv) MDA 

= 2.2 nCi................................................................................................................... 238

E-1 IREP dose parameter input screen........................................................................... 241

E-2 Selection of beta and photon radiation energies and percentages for SRS 

facilities and processes............................................................................................. 242

E-3 Selection of neutron radiation energies and percentages for SRS facilities 

and processes .......................................................................................................... 243

E-4 Adjustments to reported SRS deep photon dose ...................................................... 244

E-5 Summary of neutron dose fractions and associated ICRP 60 correction 

factors ...................................................................................................................... 245

E-6 Missed photon dose according to SRS facility .......................................................... 245

E-7 Missed photon dose according to dosimeter type ..................................................... 246

E-8 Missed photon dose according to year ..................................................................... 246

E-9 Neutron-to-photon ratios by SRS process................................................................. 247

E-10 SRS TLND Limits of Detection.................................................................................. 248

E-11 Default exposure geometries to calculate organ dose .............................................. 248

LIST OF FIGURES

Figure

3.1.1-1 SRS Map....................................................................................................................

Page

52

5.3.1-1 SRS monitored workers and collective dose............................................................... 92

5.3.2.1-1 Estimated SRS dosimeter photon response characteristics........................................ 94

5.3.4.2-1 Neutron energy spectra recorded at SRS plutonium facilities with neutron 

energy groups overlay .............................................................................................. 101

5.3.4.2-2 Trends in SRS and Hanford collective neutron dose normalized to plutonium 

production................................................................................................................. 103

5.3.4.2-3 Neutron energy spectra recorded behind the K-Reactor shield door with 

neutron energy groups overlay ................................................................................. 106

5.3.4.2-4 Comparison of various neutron spectra from different plutonium compounds 

and alloys ................................................................................................................. 108
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 11 of 248

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 05/10/2003 00-A Original issue draft technical basis document for the 

Savannah River Site. Initiated by Judson L. Kenoyer.

Draft 05/19/2003 00-B Revised to correct formatting errors and make 

additional text changes. Initiated by Judson L. 

Kenoyer.

Draft 07/03/2003 00-C Revised to incorporate changes based on information 

gathered through contacts with SRS technical subject 

matter experts. Initiated by Edward D. Scalsky.

Draft 07/14/2003 00-D Revised to respond to comments provided by NIOSH 

on July 11, 2003. Initiated by Edward D. Scalsky.

7-15-03 07/15/2003 00 First Approved Issue. Initiated by Edward D. Scalsky.

Draft 08/07/2003 01-A Draft revision to incorporate changes to Sections 2.2, 

3.2.2, Table A-23, Tables B-1, C-17, and C-19; 

addition of last paragraph in “in vitro fission product 

analysis” in section 4.1.2 and section 5.3.5; Additions 

of Dose Reconstructor Instructions: Sections 2.5, 2.5.1, 

2.5.2; 3.4, 3.4.1,3.4.2; 4.5, 4.5.1, 4.5.2, 4.5.3, 4.5.4; 

5.7, 5.7.1, 5.7.2 and Tables D-11 through D-29. 

Additions were also made to Sections 4.4.3 and 4.4.4. 

Initiated by Edward D. Scalsky.

Draft 08/19/2003 01-B Revised to incorporate changes per reviewer 

comments. Added sections 4.4.3 and 4.4.4 to previous 

record of revisions. Corrected several terms and 

definitions in the glossary, sections 1.3, and 2.2. 

Revised section 2.2 for clarity. Reassigned the thyroid 

to Group 2 in section 2.5.2. In section 4.1.2 revised 

MDA/Reporting level in Tables 4.1.1-1 and 4.1.1-2, 

clarified 10 year old 12% Pu mix as claimant favorable, 

added text to the Note re: fission product Sr-90 

sample/whole body counts, corrected absorption types. 

Added 2 paragraphs in section 2.4.1 discussing record 

forms for whole body counts and meaning of terms 

thereon. Corrected 4.4.3-x corrected to 4.4.3-9 in first 

sentence of Note:, page 79. Corrected tables 4.4.3-1 

and 4.4.3-3 Type S and M values respectively and 

corrected units. Added bullets under Note: in section 

4.4.3 “Chronic Intakes for other radionuclide.” Clarified 

tritium equilibrium and corrected calculation in section 

4.4.3, Corrected solubility type in tables 4.4.3-7 and 

4.4.3-8.Added text to sections 4.5 and 4.5.1 and 

deleted sections related to “Best Estimate Approaches” 

and renumbered sections. Deleted footnote in section 

5.3.2.1. Added words to clarify use of annual intakes for 

dose reconstruction. Corrected Table D-5 Absorption 

Type. Initiated by Edward D. Scalsky.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 12 of 248

08/21/2003 08/21/2003 01 Approved issue of Revision 01. Initiated by Edward D. 

Scalsky.

Draft 09/30/2003 02-A Draft revision to simplify internal dose reconstruction. 

Allows use of Table C-17, Maximum site-wide annual 

intakes and geometric standard deviations, for all 

environmental dose reconstruction and clarifies periods 

of applicability. Modified instructions for determination 

of undetected dose from in vitro bioassay results

(except tritium) below detection threshold. Clarified 

instructions for missed dose determination from whole 

body count results below the detection threshold. 

Corrected Ce-144 table 4.3-5. Pages changed are 24, 

57, 58, 62, 72, 80, 81, 84, 85, 86, 89, 156 and 183. 

Initiated by J. Griffin/C. Bloom.

Draft 10/23/2003 02-B Revisions to incorporate reviewer comments. 

Additional minor editorial changes were made 

throughout the document.

The Uncertainty Section (4.4.2) was deleted, because it 

is not specific to SRS and because this avoids 

inconsistencies that might result from dose 

reconstruction programmatic changes. 

Initiated by C. Bloom.

Draft 05/13/2004 02-C Revisions to incorporate medical x-ray changes and 

addition of section on photofluorograpy, addition of 

references, and information on TNX and CMX.

Initiated by Edward D. Scalsky.

Draft 09/28/2004 02-D Revision to incorporate changes to the medical x-ray 

section in response to additional comments regarding 

collimation. Deleted paragraph related to consumption 

of venison. Initiated by Edward D. Scalsky.

Draft 10/05/2004 02-E Changes all dates referring to photofluorography to 

1951-1957. Removes reference to 1967. Initiated by 

Edward D. Scalsky.

10/29/2004 10/29/2004 02 Approved issue of Revision 02. Initiated by Edward D. 

Scalsky.

Draft 01/14/2005 03-A Adds Section 3.1.3 and Attachment C.3 to include 

ingestion of foodstuffs. Initiated by Edward D. Scalsky.

Draft 02/01/2005 03-B Incorporates internal review comments. Initiated by 

Edward D. Scalsky.

Draft 03/09/2005 03-C Incorporates NIOSH review comments. Initiated by 

Edward D. Scalsky.

Draft 03/22/2005 03-D Incorporates additional NIOSH review comments. 

Initiated by Edward D. Scalsky.

04/05/2005 04/05/2005 03 Approved issue of Revision 03. Initiated by Edward D. 

Scalsky.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 13 of 248

ACRONYMS AND ABBREVIATIONS

α alpha particle

AEC (U.S.) Atomic Energy Commission

AECL Administrative Exposure Control Level

AED Atomic Energy Division (of SRP)

AEDE Annual Effective Dose Equivalent 

Am/Cm Americium/Curium Melter

ARU Acid Recovery Unit

AWE Atomic Weapons Employer

β beta particle

BR breathing rate

BSRI Bechtel Savannah River Incorporated

BSRI Bechtel Savannah River Institute 

BTM Berthold Tritium Monitor

CATI Computer Assisted Telephone Interview 

cc cubic centimeter

CEDE Committed Effective Dose Equivalent (replaced the AEDE in 1992)

c/m counts per minute

CMX Code designation for a facility

CND Criticality Neutron Dosimeter

DAC Derived Air Concentration

DAC-h Derived Air Concentration hours

DCF Dose Conversion Factor

DE Dose Equivalent

DL Detection Limit

DNA Delayed Neutron Analysis

dpm disintegration per minute

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL Department of Labor

DPSOP Du Pont Standard Operating Procedure

DWPF Defense Waste Processing Facility

DU Depleted Uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act

ERDA Energy Research & Development Administration

ESE Entrance Skin Exposure

ETF Effluent Treatment Facility

γ gamma ray

HEU Highly Enriched Uranium

HHE High Heat Waste

HHS Department of Health and Human Services

HLW High Level Waste

Hp (0.07) Dose equivalent at 0.07 mm depth

Hp(10) Personal Dose Equivalent at 10 mm depth
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HPAREH Health Protection Annual Radiation Exposure History Database 

HPRED Health Protection Radiation Exposure Database 

HRTM Human Respiratory Tract Model

HTO Tritiated Water

HVL Half Value Layer

I Intake

IXF Ion Exchange Facility

IARC International Agency for Research on Cancer

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis 

IREP Interactive RadioEpidemiological Program

KPA Kinetic Phosphorimetry Analysis

LANL Los Alamos National Laboratory

LHW Low Heat Waste

LLD Lower Limit of Detection

LSC Liquid Scintillation Counter

MDA Minimum Detectable Activity

MDL Minimum Detection Limit

MED Manhattan Engineering District

MEPAS Multimedia Environmental Pollutant Assessment System

MOL Maximum occupational limit

MPBB Maximum Permissible Body Burden

MPLB Maximum Permissible Lung Burden

MPOB Maximum Permissible Organ Burden

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH OCAS Claims Tracking System

NTA Nuclear Track Emulsion

OCAS (NIOSH) Office of Compensation Analysis and Support 

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

PA Posterior-Anterior (medical x-ray)

PAS Personal Air Sampler

PDIS Personnel Dosimetry Intercomparison Studies

PEF 238PuO2 Experimental Facility

PHEF Precipitate Hydrolysis Experimental Facility

PIC Pocket Ionization Chamber (i.e., “Pencil” dosimeter)

PSF Plutonium Storage Facility

PuFF 238PuO2 Fuel Form Facility

PUREX Plutonium Uranium Extraction (Facility/Process)

PWI Plutonium Waste Incinerator

R Roentgen

RAS Regional Air Sampler
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RBE Relative Biological Effectiveness

RBOF Receiving Basin for Offsite Fuel

RCG radioactivity concentration guide

RC & HP Radiological Control and Health Physics Department

rem Roentgen equivalent man

rep Roentgen equivalent physical

RMS Root Mean Square

RQB Radiological Qualification Badge

RRF Resin Regeneration Facility

RTF Replacement Tritium Facility

RU Recycled Uranium

SEC Special Exposure Cohort

SID Source to Image Distance

SMI Stack Monitor Integrator

SRAT/SME Slurry Receipt Adjustment Tank/Slurry Mix Evaporator

SRI Savannah River Institute

SRL Savannah River Laboratory

SRO Savannah River Operations Office

SRP Savannah River Plant

SRS Savannah River Site

SSD Source to Skin Distance

STM Stack Tritium Monitor

SWDF Solid Waste Disposal Facility

TBD Technical Basis Document

TEPC Tissue Equivalent Proportional Counter

TLD Thermoluminescent Dosimeter

TLND Thermoluminescent Neutron Dosimeter

TNX Code designation for a facility

TPO Technical Procedures Office

TRU Transuranic

uCi microcurie

U IV Uranium IV

WCF Waste Certification Facility

WSRC Westinghouse Savannah River Company
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EXECUTIVE SUMMARY

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning the preparation of dose reconstructions at particular sites or categories of sites. 

They will be revised in the event additional relevant information is obtained about the affected site(s). 

These documents may be used to assist NIOSH in the completion of the individual work required for 

each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building or group of buildings 

that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “ Department of Energy facility” as defined in the Energy Employee 

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 7384l (5) and (12)).” 

This technical basis document (TBD) represents a specific support mechanism to the ORAU Team 

concerning documentation of historical practices at the Savannah River Site (SRS). This TBD can be 

used to evaluate both internal and external dosimetry data for unmonitored and monitored workers 

and serve as a supplement to, or substitute for, individual monitoring data. This document provides a 

Savannah River Site (SRS) profile that contains technical basis information used by the ORAU Team 

to evaluate the total occupational radiation dose for Energy Employees Occupational Illness 

Compensation Program Act (EEOICPA) claimants. 

This document also provides supporting technical data to evaluate, with claimant-favorable 

assumptions, the total SRS occupational radiation dose that may reasonably be associated with the 

worker’s radiation exposure. This dose results from exposure to external and internal radiation in SRS 

facilities; to SRS occupationally-required diagnostic x-ray examinations, and to on-site environmental 

releases. Also included is the dose that may have occurred while the worker was unmonitored or the 

dose may have been missed. Over the years new and more reliable scientific methods and protection 

measures have been developed. The methods needed to account for these changes are also 

identified in this document.

The doses are evaluated using the NIOSH Interactive RadioEpidemiological Program (IREP) and the 

Integrated Modules for Bioassay Analysis (IMBA) computer codes. Information on measurement 

uncertainties is an integral component of the NIOSH approach. This document describes how the 

uncertainty for SRS exposure and dose records is evaluated.

The document is divided into five major sections; Facilities and Processes, Occupational Medical 

Dose, Occupational Environmental Dose, Occupational Internal Dose, and Occupational External 

Dosimetry. Each section has an associated attachment, which provides the critical data for the 

specialists reconstructing the doses. 

The facilities and processes used at SRS over the years and their associated source terms are briefly 

described in section one. The tables showing the radionuclides of concern for dose reconstruction and 

the years of operation of the different facilities are in Attachment A. The discussion includes the 

individual processes and the radionuclides for both the internal and external exposure potentials. 

Section 2 discusses the occupational medical dose. The medical doses result from the chest x-rays 

given during pre-employment and annual physical exams. However, certain employees may have had 

x-rays more frequently based on their work and their age. Although x-rays may have been given 

because of on-the-job injuries, these doses are not considered part of the occupational dose. The xray techniques and equipment used over the years have also changed and are presented and 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 17 of 248

discussed in this section. Attachment B presents the pertinent tables for use by the dose 

reconstructors in determining the worker’s occupational medical doses.

Section 3 provides information for calculating internal and external radiation doses to unmonitored 

workers at the SRS. Data are presented on the radionuclides released from the site facilities and 

processes. The methodology to calculate the doses received by unmonitored workers outside the 

facilities is described. Internal doses result from breathing the radioactive material in the air, ground 

level releases, or resuspended from radioactive material deposited on the ground. External doses 

result from the ambient radiation levels and from submersion in a plume of noble gases. The 

equipment used to measure both the internal and external doses changed considerably over the 

years and these changes have been taken into account in calculating the individual worker dose.

A dispersion model was developed using a 5-year average joint frequency distribution. The values 

were converted to X/Q tables for elevated and ground level releases, using the Multimedia 

Environmental Pollutant Assessment System (MEPAS) 4.0 computer code, and plotted using an 

Excel spreadsheet. Various assumptions were made to ensure atmospheric dispersion values were 

not being overestimated. The maximum air concentrations were then calculated for the principal dose 

producing radionuclides at various on-site populated areas for use in calculating the unmonitored 

worker doses. 

Attachment C provides a series of tables of the 50th-percentile and geometric standard deviation 

(GSD) intake values for selected individual radionuclides resulting from releases from various 

facilities. These values will be used by the dose reconstructors to calculate the internal and external 

doses that unmonitored workers may receive from the environmental pathway.

Section 4 discusses the occupational internal dose. The radionuclides identified by Savannah River 

as being the most significant in delivering doses or that were effective tracers for significant doseproducing radionuclides are presented. The methods for analysis of selected individual radionuclides 

are also discussed. Tables providing the minimum detectable activities (MDAs) are included for the In 

vitro analysis of urine and feces for selected radionuclides and in vivo whole body and chest counting.

Attachment D provides a more complete list of radionuclides and MDAs for use by dose 

reconstructors for calculating internal doses. 

Section 5 presents the occupational external dosimetry program for measuring skin and whole body 

doses to the workers. The methods for evaluating external doses to workers have also evolved over 

the years as new techniques and equipment have been developed. Concepts in radiation protection 

have also changed. The dose reconstruction parameters, SRS practices and policies, and dosimeter 

types and technology for measuring the dose from the different types of radiation are discussed in this 

section. Attention is given to the evaluation of doses measured from exposure to beta, gamma, and 

neutron radiation. Test results for various films exposed to different geometries and energies of 

radiation are given in tables.

Sources of bias, workplace radiation field characteristics, responses of the different beta/gamma and 

neutron dosimeters in the workplace fields, and the adjustments to the recorded dose measured by 

these dosimeters during specific years are discussed in detail.

There are sources of potential dose that could be missed because of the limitations of dosimetry 

systems and the methods of reporting low doses. This missed dose is discussed as a function of 

facility location, dosimeter type, year, and energy range. Attachment E describes the use of the 

external dosimetry technical basis parameters to facilitate the efforts of the dose reconstructors.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 18 of 248

1.0

The National Institute for Occupational Safety and Health (NIOSH) was assigned the responsibility for 

developing the technical capabilities and guidance to be used to implement the Energy Employees 

Occupational Illness Compensation Program Act (EEOICPA). Oak Ridge Associated Universities 

(ORAU) led a team, identified as the ORAU Team, to support NIOSH in the conduct of this major 

program. This technical basis document (TBD) represents a specific area of support to the ORAU 

Team concerning documentation of historical practices at the Savannah River Site (SRS). It pertains 

to evaluation of both internal and external dosimetry data for unmonitored and monitored workers and 

is to be used as a supplement to or substitute for individual monitoring data. 

INTRODUCTION

1.1 PURPOSE

The purpose of this document is to provide a Savannah River Site (SRS) profile that contains 

technical basis information used by the Oak Ridge Associated Universities (ORAU) Team to evaluate 

the total occupational dose for EEOICPA claimants. 

1.2 SCOPE

SRS operations played an important role in the US nuclear weapons program (DOE 1997). These 

SRS processes included nuclear fuel fabrication, reactor operation, radiochemical processing, 

uranium recycling, plutonium production, neutron source production, and waste management. 

Supporting documentation has been assembled in this TBD to assist in the evaluation of worker dose 

from these processes using the methodology in the respective NIOSH OCAS-IG-001 External Dose 

Reconstruction Implementation Guideline (NIOSH 2002a) and OCAS-IG-002 Internal Dose 

Reconstruction Implementation Guideline (NIOSH 2002b).

The methods and concepts of measuring radiation exposure to workers have evolved since the 

beginning of SRS operations in 1952. An objective of this document is to provide supporting technical 

data to evaluate, with claimant-favorable assumptions, the total SRS occupational dose that may 

reasonably be associated with worker radiation exposure as covered under EEOICPA legislation. 

This dose includes occupational external and internal exposure in SRS facilities; SRS occupationallyrequired diagnostic x-ray examinations, and on-site exposure to SRS environmental releases. The 

documentation addresses evaluation of unmonitored and monitored worker exposure, and missed 

dose. Consistent with NIOSH Implementation Guidelines, this document identifies how to adjust the 

historical occupational dose to account for current scientific methods and protection factors. 

This document also presents the technical basis of methods used to prepare the SRS worker dose 

records for input to the NIOSH Interactive RadioEpidemiological Program (IREP) and the Integrated 

Modules for Bioassay Analysis (IMBA) computer codes used to evaluate worker dose. Because 

information on measurement uncertainties is an integral component of the NIOSH approach, this 

document describes how the uncertainty for SRS exposure and dose records is evaluated. 

The main body of text in this document provides the description of the facilities and processes, 

historical information related to worker internal and external exposures, and environmental data for 

use when actual monitoring data are unavailable. The attachments represent the critical data and 

tables required by dose reconstructors for performing the individual claimant dose reconstructions. 

These attachments should suffice as a stand-alone document for dose reconstruction. Additional 

details, if necessary, could be found in the main body of the text.
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1.3 SITE ACTIVITIES AND PROCESSES 

This section briefly describes the facilities and processes while their associated source terms have 

been captured in Attachment A. These data would be used by dose reconstructors only when 

monitoring data are unavailable and/or other methods may not be amenable for claimant dose 

reconstruction. As indicated, the tables listed below and referenced in the text refer to the tables in 

Attachment A.

The following tables or sections are provided in Attachment A:

Table Facility or Process

A-1.1 Reactors years of operation......................................................................................

Page

140

A-1.2 Radionuclides of concern for all reactors .................................................................. 140

A-2 F Area A-Line Facility ............................................................................................... 141

A-3 221-F B-Line Facility................................................................................................. 141

A-4.1 221-F Canyon Facility............................................................................................... 142

A-4.2 221-F Canyon Facility............................................................................................... 142

A-4.3 221-F Canyon Facility............................................................................................... 143

A-5 New Special Recovery (NSR) Facility....................................................................... 143

A-6.1 F-Area Outside Facilities .......................................................................................... 144

A-6.2 F Area Outside Facilities........................................................................................... 144

A-6.3 F Area Outside Facilities........................................................................................... 144

A-6.4 F Area Outside Facilities........................................................................................... 145

A-6.5 F Area Outside Facilities........................................................................................... 145

A-6.6 F Area Outside Facilities........................................................................................... 146

A-7 238PuO2 Fuel Form Facility (PuFF) and 238PuO2 Experimental Facility (PEF) ............ 146

A-8 235-F Vaults............................................................................................................. 147

A-9 772-F and 772-1F production control laboratories .................................................... 147

A-10 E-Area Solid Waste Disposal Facility (SWDF).......................................................... 147

A-11 Plutonium Storage Facility ........................................................................................ 148

A-12 F/H Effluent Treatment Facility ................................................................................. 148

A-13 F & H Area—cooling water and retention basins ...................................................... 149

A-14 F & H Area Tank Farms............................................................................................ 149

A-15 Waste Certification Facility ....................................................................................... 149

A-16.1 221-H B-Line Facility ................................................................................................ 150

A-16.2 221-H B-Line Facility ................................................................................................ 150

A-16.3 221-H B-Line Facility, Np02 and Pu02 Facility ........................................................... 150

A-17.1 H-Canyon Facility ..................................................................................................... 151

A-17.2 H-Canyon Facility ..................................................................................................... 151

A-17.3 H-Canyon Facility ..................................................................................................... 152

A-17.4 H-Canyon Facility ..................................................................................................... 152

A-18.1 H-Area Outside Facilities .......................................................................................... 153

A-18.2 H-Area Outside Facilities .......................................................................................... 153

A-19.1 Receiving Basin for Off-site Fuel (RBOF) ................................................................. 154

A-19.2 Resin Regeneration Facility...................................................................................... 154

A-20 Tritium Facilities........................................................................................................ 154

A-21 Uranium Target Fabrication Facility, 313 M .............................................................. 154

A-22 Uranium Processing Facilities 320-M, 322-M, and 341-M......................................... 155

A-23 Fuel Fabrication Facility, 321-M................................................................................ 155

A-24.1 S Area Defense Waste Processing Facility............................................................... 156

A-24.2 S Area Defense Waste Processing Facility............................................................... 156
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A-25.1 Z Area ...................................................................................................................... 157

A-25.2 Z Area ...................................................................................................................... 157

The data in the tables in Attachment A have all been extracted from ESH-HPT-960197, Facility 

Descriptions (U), Thomas R. La Bone, October 1, 1996.

All stated assumptions are also from La Bone (1996).

The following definitions apply to all tables in Attachment A that contain information on the Activity 

Fraction:

The activity fraction is the fraction of the total activity represented by a particular radionuclide.

The units in the tables in Attachment A are conventional units. To convert from mrem/nCi to Sv/Bq, 

multiply by 2.7E-7. 

To convert from nCi to Bq, multiply by 37.

Where solubility classes were available in the Savannah River Site literature, their designations have 

been changed to the ICRP absorption type designations in the tables. According to the International 

Commission on Radiological Protection ICRP Publication 68 “Dose Coefficients for Intakes of 

Radionuclides by Workers” the ICRP 30 D, W, and Y absorption classes correspond broadly to the 

absorption Types F, M, S respectively. These values are given in the second column of the tables in 

Attachment A.

The differences in the radionuclides for the various processes in Attachment A are related to the 

mode of exposure. To minimize the number of pages that the dose reconstructor would need to 

search, both internal and external radionuclides have been combined into one table for each of the 

facilities/processes addressed for the SRS. Radionuclides identified as “being of concern” for internal 

exposure may be different from those identified as providing the greatest external exposure. The last 

column of some of the tables is “Significant to External Exposure.” An “X” in this column indicates that 

the radionuclide contributes to external dose. While 90Sr is typically an internal exposure hazard, 

under certain circumstances it could also be an external exposure hazard. It is listed in this document 

as both for completeness. 

Dates of operation of the various facilities have been taken from:

Cummins, et al, Radioactive Releases from the Savannah River Site, 1954 – 1989, WSRC-RP-91-

684,

DOE/EM-0319, Linking Legacies, January, 1997, and

Till, et al, Savannah River Site Environmental Dose Reconstruction Project, Phase II, Risk 

Assessment Corporation, 30 April 2001.

The discussions that follow relate directly to the table represented in the individual section heading, 

e.g., A-1 Reactors will relate to Table A-1 in Attachment A, etc. 

A-1 Reactors (LaBone 1996, Till 1999)

Five heavy-water reactors designated R, P, L, K, and C were constructed at SRS in the early 1950s, 

each of which is in the locations designated R-Area, P-Area, L-Area, K-Area, and C-Area respectively. 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 21 of 248

All five reactors were operating in 1955. Reactor shutdown began in 1964 with R-Reactor. C-Reactor 

was shut down in 1987, L-Reactor was shut down in 1968, restarted in 1985, and then shut down 

again in 1988. K-Reactor and P-Reactor operated with few interruptions until 1988. A restart of KReactor began in 1991, but was not completed. None of the reactors are currently operational.

Depleted uranium, 237Np, and lithium targets were irradiated to produce 239Pu, 238Pu, and tritium, 

respectively. The targets are irradiated and transferred to the Tritium Facilities, H Area Separations 

(

238Pu), and F Area Separations (239Pu) for processing. Spent reactor fuel is sent to H Area 

Separations for processing. In the past the reactors were also used to convert thorium to 233U as well 

as to irradiate targets that produced transplutonic elements such as curium and californium.

The composition and relative abundance of the radionuclides in the reactor environment are not 

known. Representative samples are taken annually to determine the isotopic makeup of 

contamination in reactor areas. When the reactors are operating, fission neutrons are an external 

exposure concern in some areas. For the purpose of determining internal or external exposure in the 

reactor areas, the radionuclides of concern in the reactor areas are presented in Table A.1-1.

Internal exposure

Because the reactors were heavy-water reactors, intakes from the reactor buildings would be so 

dominated by tritium that intakes of fission/activation products can be ignored. If, in a specific case, 

the bioassay data indicate otherwise, then see the fission product discussion for the 221-F Canyon 

Facility. 

. Discussions by T.R. La Bone with Health Physics Operations personnel and a 

review of ANSI 343 were used to identify the radionuclides of concern. Tritium is assumed to 

produce most of the personnel exposure because of the large quantities present.

External exposure

A-2 F-Area-A-Line Facility

. During operation, fission neutrons are also a concern in certain areas. 

The A-Line facility accepts dilute, depleted uranyl nitrate solution from the 221-F Facility and converts 

it into uranium trioxide powder. Dilute uranyl nitrate solution is sent to A-Line from the solvent 

extraction process in the 221-F Facility where it is initially decanted to remove tributyl phosphate nparaffin solvent. It is then concentrated by evaporation. A second evaporation yields a hydrate of 

uranyl nitrate that is heated in a denitrator to yield the uranium trioxide product.

The A-Line facility also has a dissolver to prepare a uranyl nitrate solution from the uranium trioxide 

product and nitric acid when it is required as a process feed stream back into the solvent extraction 

process in the 221-F Facility. The dissolver is also used to recycle off-standard A-Line product. The 

uranium trioxide product is placed into drums for storage on-site or into packages for off-site 

shipment.

The A-Line Facility can be broken down into unit operations that perform specific functions in the 

creation of the final product. These processes consist of:

• Evaporation, that accepts uranyl nitrate from the 221-F Warm Canyon and concentrates it; 

• Purification, which is an optional step, but by using silica gel absorbs residual zirconium and 

niobium out of the uranyl nitrate solution;

• Hydrate evaporation, where uranyl nitrate from the silica gel or evaporator is again 

concentrated; 
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• Denitration, where material from the hydrate evaporation is received and converted from 

uranyl nitrate hexahydrate to uranium trioxide; 

• Material handling, where the uranium trioxide powder is removed from the denitrators and 

placed into storage bins from which it is loaded into containers, and 

• Oxide dissolution, where 221-F occasionally requires a uranyl nitrate-nitric acid solution for 

use as a process stream.

• The fume recovery system converts oxides of nitrogen into nitric acid for reuse using 

exhausters, a venturi scrubber, off-gas coolers, an absorption column, and various pumps and 

tanks. (The venturi scrubber removes uranium trioxide particles from the denitrator off-gas 

and is the only piece of the system that is of radiological concern.)

Internal exposure. The chemical forms that are produced in A-Line are uranyl nitrate and uranium 

trioxide. These materials are Types F and M, respectively. See also the discussion in 4.1.2, “In Vitro 

Bioassay for Uranium,” for guidance on interpreting uranium urinalyses for natural or depleted 

uranium. Although the trace contaminants in the uranium provide little to “compliance” dose (i.e., 50-

year committed doses), they may contribute dose to different organs than the uranium and they will be 

more significant if the time from intake to diagnosis is short relative to 50 years. Hence, the trace 

radionuclides should be included in intake calculations in the ratio relative to 238U as shown in Table 

A.2. 

External exposure

The activity fractions given in the table indicate the predominant radionuclides of concern for external 

dose would be 238U, 141Ce, 144Ce, 103Ru, 106Ru, and 95Nb. 

. Depending on the age of the material in the A-Line, it is assumed a mixture of 

long-lived fission products would be the major contributors of external dose. The long-lived fission 

products of interest would include 95Nb, 103Ru, 106Ru, 141Ce, 144Ce, 144Pr, 151Sm, and 155Eu. Other 

contributors would be 235U, 236U, 238U, and 239Pu.

A-3 221-F B-Line Facility

The F B-Line Facility is located in the 200-F Separations Area and is used to convert plutonium nitrate 

into plutonium metal or plutonium oxide. It also has the capability to recover plutonium from on-site 

and off-site scrap. Initially the plutonium produced in this facility is primarily 239Pu in a dilute nitric acid 

solution. From there it undergoes many chemical processes that produce an end product.

The individual processes in the 221-F B-Line Facility consist of:

• Cation exchange, where dilute plutonium solution transferred from the canyon storage tanks to 

the F B-Line receiving tanks is sent through a cation exchange column that removes the 

plutonium along with some impurities; 

• Drying and conversion, where the plutonium trifluoride taken from the filtration process is dried 

and roasted to produce plutonium dioxide and plutonium tetrafluoride;

• Reduction, where the plutonium dioxide-plutonium tetrafluoride mixture is weighed, mixed with 

metallic calcium and placed into a magnesium crucible whose inside is filled with magnesium 

oxide sand, sealed, pressurized with argon, and heated to a molten state. Pu metal flows to 

the bottom of the crucible and forms a button shaped solid, which is then cooled, separated 

from the calcium slag, and die-stamped for identification;
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• Button finishing, where the stamped button is submerged in nitric acid, rinsed in water and 

dried (after which one or two grams of the plutonium button are removed using a drill), placed 

in a can, crimp-sealed, placed in a plastic bag, a second can, crimp-sealed, and sent to 

storage.

The drill turnings are sent to the 772-F Laboratory for analysis and recovery, where plutonium is 

extracted from the solid scrap from on-site and off-site sources, as well as miscellaneous solutions 

from F B-Line by using an anion exchange resin. The scrap is initially dissolved using a mixture of 

aluminum nitrate and nitric acid, after which the plutonium scrap is sorbed on the anionic resin and 

sent back to the canyon for processing.

External exposure considerations

Neutrons are also emitted by spontaneous fission along with associated beta and gamma radiations 

from 238Pu and 240Pu and progeny. Plutonium-241 decays by beta emissions to 241Am which emits a 

60 keV photon and x-rays with an energy of 17 keV. These photons are the primary external exposure 

concerns for 241Pu.

. The alpha emitting isotopes of plutonium are not usually an 

external exposure concern. However, when they exist in a fluoride form, the alpha particles interact 

with the fluorine by an (α, n) reaction to produce a significant source of neutrons. Therefore, neutrons 

are a major external exposure concern in the F B-Line.

Many different mixtures of plutonium isotopes are possible in the 221-F B-Line, but it is assumed that 

the mixture will be either 6% or 12% Pu. The percent refers to the mass percent of 240Pu. The 

radionuclides of concern in Table A-3 are the same for both internal and external exposure.

Internal exposure

A-4 221-F Canyon facility

. The plutonium chemical compounds in the process are varied, and the lung 

inhalation/solubility types vary with these compounds so Table A-3 does not list a specific type. If the 

bioassay data are not robust enough to imply the type, the dose reconstructor should assume type M 

for nitrate form and type S for oxide form. However, old contamination, even if originally in the nitrate 

form, slowly oxidizes and could be considered type S. If no information is available concerning the 

chemical form, assume a type that is claimant-favorable: type M will increase the dose to systemic 

organs; type S will increase the dose to the lung and GI track organs. 

The 221-F Canyon Facility consists of two parallel canyons that constitute the process area. These 

canyons contain the high-activity and low-activity radioactive materials and are referred to as the hot 

and warm canyons, respectively.

The F-Canyon facility uses the Plutonium Uranium Extraction (PUREX) process to recover 239Pu, 

237Np, and 238U from reactor-irradiated targets. The plutonium is transferred to the B-Line and the 

uranium is transferred to the A-Line facility where they are processed into a solid form.

The 221-F Canyon processes are:

Head End 

Irradiated targets are received from the reactor areas. The aluminum cladding is dissolved in 

hydroxide-sodium nitrate solution, and the uranium metal is then dissolved in nitric acid.
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Solvent Extraction

The first solvent extraction process separates the plutonium and the uranium from the fission 

products using tributyl phosphate in a hydrocarbon diluent. In the second cycle, plutonium and 

uranium are separated and purified.

Anion Exchange

The aqueous waste produced in the first solvent extraction cycle and the second uranium 

cycle contain neptunium and plutonium in a nitric acid solution, which are recovered using an 

anion exchange resin column. After the solution is sent through the column, the diluted 

actinides are sent to the first solvent extraction cycle or warm canyon for further processing 

and separation. If the concentration of the actinides is such that the solution is sent to the 

warm canyon, a further series of columns are used to purify and separate the plutonium from 

the neptunium. At this stage, 232Th and 234Th are also removed.

Waste Evaporators

Aqueous waste from the first extraction cycle, the second uranium cycle, acid strip 

condensate, ventilation system sumps, and concentrate from the second stage high activity 

waste evaporator are all combined and evaporated in the first stage continuous evaporator. 

This process reduces the volumes of waste going to the high level waste storage tanks.

Locations of Increased Exposure Potential

The hot and warm crane cabs in 221-F have increased potential for exposure to radioactive 

materials. These areas are the major concern because the majority of the process is isolated 

and run remotely due to the high radiation fields.

Internal exposure

Some fission products will be more volatile than others during the dissolution so hopefully whole body 

counts will provide the ratios of the gamma-emitting radionuclides in the body. If fission products are 

detected in a worker from this facility, and no 90Sr or plutonium bioassay data are available, assume

the ratios provided in Table A-4. An intake of 241Pu using the ratio to 239Pu provided in Table A-3 or 

for standard 6% or 12% plutonium mixture (Table 4.1.1-3) should also be included. To be claimantfavorable, an intake of 91Y the same magnitude as 106Ru should be assumed. 

. The majority of the dose from an intake of material from the Head End Stream 

comes from plutonium, cerium, and ruthenium, but because different radionuclides target different 

organs in the body, none of the radionuclides listed in Table A-4 should be ignored. Intakes of the 

fission and activation products would be detectable with good sensitivity by whole body counts, except 

for 90Sr and 91Y (the progeny of 91Sr). Not listed, but undoubtedly present, was 89Sr. However, 89Sr 

would have been counted as part of the gross beta count for 90Sr, and assigning a radiostrontium 

gross beta result to 90Sr is claimant favorable. If fission products were detected by whole body counts 

and no urinalysis for 90Sr was available, an intake of 90Sr equal to the intake of 137Cs should be made. 

Because 137Cs is generally more mobile in the reactor and Head End work environment, this 

assumption should be conservative. This assumption would account for the dose from 91Sr/91Y as 

well, which because of the short physical half-lives, produce negligible dose compared to the 90Sr. 

In the second uranium cycle stream (Stream 2) the majority of the dose comes from 238U and 239Pu; 

however, the same considerations concerning fission products and 241Pu stated in the preceding 

paragraph should be applied (See Table A.4-2). 

The third stream is known as the second cycle plutonium stream (Stream 3) in which plutonium 

dominates (See Table A.4-3).
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Personnel in this facility are exposed to all processes. Since that is the case, it is not possible to 

determine the uranium ratios required to determine the Minimum Detectable Intake (MDI). Since 238U 

is the main isotope in every process stream, the MDI was figured specifically for 238U. 

External exposure

The majority of the external dose from Stream 1 comes from cerium, zirconium, and ruthenium, all 

photon and beta emitters. In the second stream 238U, which feeds 234mPa, makes up the majority of 

the external dose along with zirconium which emits both photons and betas. In the third stream the 

majority of the external dose comes from 241Pu and 95Zr.

. Fission products are present in the 221-F Canyon Facility and, depending on the 

age of the material (i.e., the length of time out of the reactor), the isotopes of concern and the fraction 

of the activity of these materials will change. 

A-5 New Special Recovery (NSR) Facility

The NSR facility recovers plutonium from a variety of sources including scrap plutonium from offsite 

sources and from the F B-Line facility. The processed plutonium is conditioned so that it can be 

blended back into the plutonium nitrate feed stream to the F B-Line process.

The New Special Recovery Facility consists of the following processes: 

• Material Receiving and Assaying, where the material is assayed to determine its plutonium 

content before any processing begins; 

• Feed Preparation, which is dependent on the type of scrap; 

• Dissolution and Filtration, which converts the solid feed material into a solution that is suitable 

for further purification by either anion exchange or by PUREX solvent extraction; 

• Anion Exchange Purification, which removes impurities such as fluoride, aluminum, 

americium, sulfate, and uranium;

• Waste Handling, which comes from sludge, residues, slurries, and from solid materials such 

as tools, filters, gloves, etc., and 

• Analytical and Assay Operations, where samples are analyzed and returned to the process.

The function of the NSR is to recover plutonium from a variety of materials. These materials also 

contain impurities such as americium and uranium. The isotopic content is not constant in the material 

or between materials. For example, the unirradiated reactor cores percent weight range from 6% to 

12% for 240Pu, 0% to 10% for 241Am, and 60% to 80% for uranium. It is assumed that the feed 

material is principally enriched uranium.

Internal exposure. The isotopes that deliver the majority of the internal dose are 239Pu, 240Pu, 241Am, 

235U, and 236U (assuming enriched uranium). 

External exposure. The isotope that delivers the majority of the photon (external) dose is 241Am. 

Some neutron dose will also be present due to spontaneous fission and an (α, n) reaction with 

fluoride, if present. 
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A-6 F-Area Outside Facilities

The F-Area Outside Facilities provide general support, principally to the processing of irradiated fuels 

and targets in Building 221-F. The term “Outside Facilities” is used to describe a wide variety of 

processes, utilities, and services that are ancillary to the primary 200-F Area operations. The main 

processes of radiological concern are as follows: 

Water Handling Facilities

This facility is used to provide process water and acidified water streams for the canyons, 

retain discard water in holding tanks for analysis, and decant spent solvent from waste water.

Acid Recovery Unit (ARU)

This facility concentrates nitric acid for reuse. All material in this location can contain 

radioactive material. 

General Purpose Evaporator:

The General Purpose (GP) evaporators concentrate aqueous waste having activity that is in 

excess of disposal limits yet is low enough to be evaporated in unshielded equipment.

General Purpose Waste Tanks

The tanks in this facility are used to collect GP evaporator feed and to receive and dispense 

evaporator overheads for recycle to the process.

Waste Handling Facilities

The F-Area Waste Handling facilities are used for storage and transfer of high- and low-activity 

wastes that are primarily from the Metallurgical Building 235-F, Savannah River Laboratory 

(SRL) Waste Concentration Building 776-A, and the Production Control Facilities, Building 

772-F and 100-Area sources.

Segregated Solvent Facilities

Segregated Solvent facilities provide solvent purification and tank storage before the solvent is 

returned to F-Canyon for reuse.

Transfer Tanks

These tanks provide intermediate pumping stations for those solutions transferring from 

Building 221-F.

Sump Collection Tanks

Two sump collection tanks are located in separate underground vaults. One tank receives 

radioactive waste condensate, by jet stream, from sumps or tanks related to the canyon air 

exhaust system and transfers the waste to the hot or warm canyon waste evaporator feed 

tanks. Discharges to hot canyon cell sumps are first transferred to the second sump collection 

tank and then to the canyon for analysis and disposition.

Recycle Sump

The Recycle Sump collects drainage and overflow from the canyon auxiliary tanks and holds 

contaminated or recycled liquids.

Auxiliary Systems and Support Facilities

The only auxiliary system and support facility of concern is the plant laundry located in Building 

723-F.
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Internal exposure. The radionuclides of concern for the ARU, water handling, and the basin transfer 

tanks are 95Zr, 95Nb, 103Ru and 106Ru. 

External exposure

The radionuclide distribution in the basin transfer tanks is assumed to be similar to the water handling 

distribution.

. The radionuclides of concern for the ARU, water handling, and the basin transfer 

tanks are 95Zr, 95Nb, 103Ru, and 106Ru. The primary radionuclides of external dosimetric concern for 

the GP evaporators, Segregated Plutonium Solvent, and Laboratory Waste are 241Pu, 103Ru, 106Ru, 

95Zr, and 95Nb. The radionuclides of concern for the Segregated Uranium Solvent are 103Ru, 106Ru, 

95Zr, 95Nb, and 238U which feeds 234Pa.

A-7 238PuO2 Fuel Form Facility (PuFF) and the 238PuO2 Experimental Facility (PEF) 

The primary function of the PuFF facility is to produce encapsulated plutonium oxide fuel forms. This 

is accomplished by converting 238Pu oxide powder into fuel forms for heat sources by powder ceramic 

and metallurgical process. The final products are compacted PuO2 fuel shapes. The PEF provides 

space for the testing of processes to be used in the PuFF facility. The plutonium oxide blends used in 

the Puff and PEF facilities are very similar.

The PuO2 is fabricated and encapsulated in six manipulator cells filled with inert gas, three air-filled 

manipulator cells, five wing cabinets, and one hood.

The fuel pellets are made by hot-pressing a blended PuO2 shard mixture prepared from calcined Pu 

oxalate powder received from H B-Line. Feed processing includes oxygen-16 enrichment, ball milling, 

compaction, granulation, and sintering. After final heat treatment, the fuel pellets are encapsulated in 

iridium-clad vent sets by tungsten inert gas welding. The conversion of plutonium oxide powder into 

dense PuO2 fuel forms is conducted in the PEF.

Internal exposure. The majority of the dose from the PuFF and PEF material comes from plutonium. 

External exposure

A-8 235-F Vaults

. The majority of the external dose from the PuFF and the PEF comes from the 

neutrons generated by spontaneous fission of 238Pu and the photon/beta emissions from 241Am, 232Th, 

and 232Np. 

Vaults in the 235-F Facility were built to store plutonium at different production steps prior to 

completion and were created not only for safety reasons but also to establish safeguards for security 

purposes. The 235-F facility receives plutonium and plutonium dioxide and fabricates it into a variety 

of items used in the nuclear program. There are three vaults located in the 235-F facility:

Plutonium Oxide Vault

Used to store plutonium oxide powder from the PuFF and PEF, and the Actinide Billet 

Fabrication Facility. Neptunium-237 powder and billets are also stored in this facility.

Finished Product Vault

Used for the storage of uranium and transuranics in powder form like the plutonium oxide 

vault. In addition, this vault contains a water cooled, shielded storage area for the storage of 

finished 238Pu products.
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Scrap Vault

Used to store metal and oxide scraps of uranium, thorium, and plutonium that are in the form 

of chips, sweepings, powder, blended samples, or metal turnings.

Internal exposure. Radioisotopes that are stored and found in the 235-F vaults are 238Pu, 239Pu, HEU, 

and 237Np. All are of concern for determining internal dose.

External exposure

A-9 772-F and 772-1F Production Control Laboratories

. Detailed information on the composition of the radionuclides stored and found in 

these vaults is not available; therefore, the radionuclides listed in the associated table in the 

Attachment are considered to be pure. The major source of external exposure is 238Pu spontaneous 

fission neutrons and betas from 238Pu progeny in addition to gammas and x-rays.

The production control laboratories provide analytical support for the 200-F and 200-H separations 

processes. Support is also provided to Waste Management, Reactors, and Raw Materials.

The radionuclides of concern are plutonium, uranium, and fission products. The major source of 

external exposure is 238Pu spontaneous fission neutrons and betas from 238Pu progeny in addition to 

gamma, x-ray photons, and betas from fission products.

A-10 E-Area Solid Waste Disposal Facility (SWDF)

Solid and liquid low-level waste, generated at the SRS and the SRL, and some waste from off-site are 

stored at the SWDF.

Solid wastes are located above and below ground. The SWDF is located in areas labeled 643-E and 

643-7E. The old burial ground is designated 643-E and ceased operations in 1972. Burial operations 

were phased in during 1969 to 1972 in the present SWDF designated as 643-7E. Most of the solid 

waste above ground is TRU waste, mainly 238Pu and 239Pu contaminated waste. Contaminated 

process equipment contained in steel boxes is stored above ground.

Liquid waste is stored in below ground tanks. Thirty two tanks have been used to store liquids. As of 

1988, only three tanks contained liquids. Liquid radioactive waste is in the form of degraded organic 

process solvents, waste oil, scintillation solutions, and contaminated hazardous waste chemicals.

Due to the diverse nature of the radioactive waste, the radionuclide content of each type of waste has 

not been defined. The radionuclides of highest concentration (Ci/vol) are tritium, cobalt, plutonium, 

and curium. 

External Exposure

A-11 Plutonium Storage Facility (PSF)

. The radionuclides of external dosimetric significance are the fission and activation 

products associated with reactor operations, such as 60Co, Cs, Sr, Ni, Ce, Rh, and Zr.

The PSF is located in the 200-F Separations Area and is used to receive, store, monitor, retrieve, and 

ship packaged plutonium. Plutonium in solid and powdered form (such as scrap) contained in 

approved shipping containers may be temporarily stored there prior to shipment.

The Receipts Assay Facility is the primary location for material accountability. Specifically, it is the 

area used for the unpacking of off-site material. Material from off-site is unpacked in a shielded 

glovebox, the inner package is removed, and it is passed on to the loading station where the inner 

package is weighed, labeled and sealed, then sent to the nondestructive assay laboratory.
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Internal exposure. The radionuclide of concern in PSF is plutonium. Many different mixtures of 

plutonium isotopes are possible, but it is assumed that the mixture is either 6% Pu or 12% Pu, the 

percent referring to the mass percent of 240Pu.

External exposure

Neutrons are also emitted by spontaneous fission along with associated beta and gamma radiations 

from 238Pu and 240Pu.

. The alpha-emitting isotopes of plutonium are not usually an external exposure 

concern. However, when they exist in a fluoride form, the alpha particles interact with the fluorine by 

an (α, n) reaction to produce a significant source of neutrons. 

Plutonium-241 decays by beta emissions to 241Am which emits a 60 keV photon and x-rays with an 

energy of 17 keV. These photons are the primary external exposure concerns for 241Pu.

All radionuclides identified in Table A-11 are applicable for determining internal and external dose.

A-12 F/H Effluent Treatment Facility (ETF)

The ETF consists of large, impermeable storage basins (with lift stations located in both F and H 

Areas) and a treatment facility (Building 241-84H located in H Area). ETF is designed to remove 

hazardous chemical and radioactive contaminants from the 200-Area liquid effluent waste streams 

and concentrate them for disposal. The effluent processing steps include feed adjustment (which 

consists of oxidation, grit collection and removal) and pH adjustment; filtration, which consists of three 

parallel trains with three stages of filtration; organic removal, consisting of mercury removal ion 

exchange columns (3) in parallel followed by 2 activated carbon beds in series; reverse osmosis, 

which removes the bulk of the solid material in the 200-Area liquid effluent wastes using a three-stage 

reverse osmosis process that concentrates dissolved solids; ion exchange, which removes ionic 

species of radionuclides and mercury; and evaporation (which is fed concentrate from the filtration 

and reverse osmosis systems, regeneration solution from the ion exchange process, and the cleaning 

solution from the filtration and the reverse osmosis processes). Evaporator steam condensate is 

collected in a condensate receiver and pumped to the H-Area segregated water system for analysis. 

The overheads are transferred to the organic removal feed tank or recycled to the waste water 

collection tanks for reprocessing. The evaporator bottoms are sent to the H-Area waste tank farm for 

transfer to the Saltstone Facility. 

Internal exposure. The majority of the annual internal dose from the F/H effluent material comes from 

plutonium, cerium, and ruthenium. 

External exposure

A-13 F/H Cooling Water and Retention Basins

. The majority of the external dose from the F/H effluent material comes from 

cerium, cesium, zirconium, strontium, and ruthenium.

The F and H Areas consist of chemical separation plants and appropriate support facilities for the 

production of purified radionuclides, principally uranium, plutonium, and tritium. The cooling water and 

retention basins are support facilities for the separations and waste management areas. Low-level 

radioactive liquid wastes from the separations areas are routed to the basins when activity is 

detected.

Cooling Water Basins

When radioactivity is encountered in a cooling water system, immediate action is taken to 

divert the water and isolate the leak. The cooling water system is flushed until the release 

guidelines are met.
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Retention Basins

Both F and H Areas use a retention basin when it is necessary for temporary storage of storm 

drains that may be contaminated. After sampling and analysis, the water may be further 

processed, transferred to ETF, or released.

Internal exposure. The majority of the annual internal dose from the F and H Area basins comes from 

plutonium and uranium. 

External exposure

A-14 F/H Area Tank Farms

. The majority of the external dose from the F Area and H Area basins comes from 

137Cs, 90Sr, 106Ru, 144Ce, and 238U feeding 234mPa. 

High-level liquid radioactive wastes from SRS are received and managed in large underground tanks 

in the waste tank farms in 241-F and 241-H. The two farms contain 51 large subsurface tanks and 

related facilities required for safe handling, processing, and temporary retention of liquid wastes. The 

liquid waste comes primarily from fuel reprocessing operations in F and H Areas with quantities from 

the Receiving Basin for Off-site Fuel and Resin Regeneration Facility (RBOF-RRF) operations, SRL, 

and the 100 Areas. Liquid wastes are classified as either high heat waste (HHW) or low heat waste 

(LHW), depending on whether or not the concentration of radioactivity is such that forced cooling is 

required to maintain waste temperature within operating guidelines.

F Area and H Area Reprocessing Waste

Liquid waste from Buildings 221-F and 221-H is made up of many waste streams generated 

during the recovery and purification of the reprocessing operation, i.e., the HHW from 221-F & 

H and the LHW from F Area and H Area.

RBOF-RRF

Waste from RBOF (244-H) and RRF (245-H) is produced during regeneration of ion-exchange 

beds, backwashing filters, cleaning and handling of fuel and target elements, and similar 

incidental operations. Radioactivity in these liquids is approximately 400 μCi/gal, but the 

volume ranges from 1.0 to 2.3 million gal/yr.

SRL and 100-Area Waste

Batches of liquid waste from SRL and the reactor areas that are not suitable for discharge are 

delivered to 211-F where the waste is evaporated, adjusted for alkalinity, and transferred to 

the waste storage tanks. Reactor waste is handled differently if it contains insoluble materials, 

such as sludges or slurries, at which time it would be unloaded directly to the appropriate tank.

Waste Processing Operations

Liquid waste from separations areas and other locations are routinely processed through a 

series of operations allowing segregation and consolidation of the waste components prior to 

interim storage.

Waste Evaporation

Radioactive waste received in the waste tank facilities is reduced by evaporation, and the 

concentrated solutions are immobilized by solidification of the residual salts. The evaporator 

condensate is continuously monitored; and if the radioactivity exceeds1500 dpm/mL gamma, 

the condensate is diverted and recycled.

The radionuclide contents of the waste depend on the age of the waste and vary with the source 

stream and prior processing. 
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Internal exposure. The majority of the annual internal effective dose equivalent in the F Area 

combined waste tank is delivered by 90Sr, 144Ce, and 244

Cm. The majority of the annual internal 

effective dose equivalent in the H Area combined waste tank is delivered by 90Sr, 144Ce, and 238Pu. 

External exposure

A-15 Waste Certification Facility (WCF)

. The majority of the external dose in the F Area Combined Tank Waste is 

delivered by 90Sr, 144Ce, 137Cs, and 106Ru. The majority of the external dose in the H Area Combined 

Tank Waste is delivered by 90Sr, 144Ce, and 238Pu.

The Waste Certification Facility (WCF), Building 724-8E, located in the 643-E Solid Waste Disposal 

Facility, is used to certify and package drums of transuranic (TRU) waste. The functions of the WCF 

are to assay and certify the contents of TRU waste drums and to package, load, and ship the drums. 

Only physical handling of sealed drums is performed in the WCF.

Typical generic contents of the TRU waste drums include:

Combustibles: paper, gloves, sweepings, cloth rags

Noncombustibles: HEPA filters, crucibles, glassware

Chemicals: evaporated sludge, caustic soda, soda lime, spent resins

Equipment: glove boxes, motors, scales.

Normally the drums contain either 238Pu or 239Pu depending on the source of the TRU waste.

Internal exposure. The radionuclides of concern for internal exposure are the same as for the Solid 

Waste Disposal facility, 643-E, and also include tritium, curium, plutonium, and activation products.

External exposure

A-16 221-H B-Line Facility

. The radionuclides of concern for external dosimetric purposes are the same as 

those for the Solid Waste Disposal facility, 643-E and also include curium, fission products, and 

activation products. 

The new HB Line was designed to replace the aging HB-Line production facility. The HB Line consists 

of three separate facilities: the Scrap Recovery Facility, Neptunium Oxide Facility, and Plutonium 

Oxide Facility.

The Scrap Recovery Facility is designed to routinely generate nitrate solutions of 238Pu or 235U/239Pu 

scrap suitable for purification by anion exchange or solvent extraction in the canyon. Scrap is received 

and dissolver batches prepared, based on assay of the scrap. The solid scrap is dissolved in hot nitric 

acid containing trace fluoride ion, transferred through a filter bag, collected in a tank, sampled for 

accountability and process control, then diluted with nitric acid and transferred to the proper canyon 

vessel as a nitrate solution.

The radionuclides of concern in the scrap material are composed of either 238Pu, in oxide form, or 

235U/239Pu mixture. The 235U/239Pu mixture normally contains greater than 20% enriched uranium.

In the Neptunium Oxide Facility, neptunium nitrate solutions are received from H Canyon and F 

Canyon and transferred to a receipt tank located in the HB Line. The solutions are precipitated to yield 

neptunium and plutonium oxalate. The neptunium and plutonium oxalate cake is then loaded into a 

furnace and calcined to NpO2 and PuO2.

In the H B-Line it is assumed that the NpO2 is 100% 237Np for purposes of dose calculations.
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In the Plutonium Oxide Facility plutonium nitrate solution from H Canyon is converted to plutonium 

oxide powder by the oxalate precipitation and calcinations method. 

A-17 H-Canyon Facility

The H-Canyon building was designed for the separation and recovery of 239Pu and 238U from irradiated 

natural uranium by the PUREX process. Over the years, other processes have been added. These 

operations include the processing of irradiated enriched uranium to recover 235U, the processing of 

irradiated neptunium targets to separate and recover 238Pu and 237Np, and the processing of thorium 

to recover 233U.

In the uranium cycle, standard enriched uranium fuel is dissolved in nitric acid catalyzed by mercuric 

nitrate. The dissolved fuel is clarified in head end, and the adjusted solution is fed into the solvent 

extraction system. Uranium and neptunium are extracted from the fission products and separated 

from each other in the first cycle. The uranium product solution is transferred out of the building for 

further processing. The neptunium product solution is transferred to the HB Line for oxide conversion.

In the neptunium cycle, irradiated neptunium targets are dissolved in nitric acid. Plutonium-238 and 

neptunium are separated from fission products by a series of anion exchange resin columns. The 

plutonium and neptunium product solutions are then concentrated, precipitated to oxalate, and 

calcined to oxides. The plutonium oxide is either shipped offsite or sent to Building 235-F for formation 

into heat sources. The neptunium oxide is sent to Building 235-F for refabrication into billets.

The radionuclides of concern will vary with the age of the material.

Internal exposure. The majority of the annual internal dose for the First Cycle Stream comes from 

plutonium and cerium. The majority of the annual dose for the Second Uranium Cycle Stream comes 

from uranium. The majority of the annual dose for the Second Neptunium Cycle Stream comes from 

plutonium and cerium. The majority of the annual dose for the Work Stream comes from plutonium 

and cerium. 

External exposure

A-18 211-H Area Outside Facilities

. In all streams (i.e., Uranium, Neptunium, and Work Streams), the radionuclides of 

greatest concern for external exposure are the fission products cesium, cerium, zirconium, ruthenium, 

and strontium. 

H-Area Outside Facilities include a number of processes that support the separations function of the 

200-H Area. The main processes and services of radiological concern are A-Line, Water Handling 

Facilities, Acid Recovery Unit (ARU), General Purpose Evaporator, Segregated Solvent Facilities, 

Transfer Tanks, Sump Collection Tanks, Recycle Sumps, Auxiliary Systems, and Support Facilities.

Many of the processes and services do not have any connection to each other in terms of function in 

the Outside Facilities. Due to this, it will in some cases be necessary to treat each process separately 

in terms of radiological protection.

A-Line

The H-Area A-Line receives a dilute aqueous solution of uranyl nitrate hexahydrate (UNH) 

enriched in 235U from the canyon where it is temporarily stored and then shipped off-site in 

tank trucks.
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Water Handling Facilities

The primary equipment for water handling consists of tanks, skimmers, and coolers, all of 

which can contain radioactive materials. The facility is operated to provide process water and 

recycle water for the canyons, retain recycle water prior to analysis to permit disposal, decant 

and discharge skimmed solvent, and store weak acid feed for ARU.

Acid Recovery Unit

Nitric acid is concentrated in the ARU for reuse.

General Purpose Evaporator

Low-activity aqueous waste is concentrated in the General Purpose (GP) Evaporator reducing 

the volume prior to sending the waste to the tank farm.

Segregated Solvent Facilities

The final purification and storage of the solvent occurs in the Segregated Solvent Facilities 

prior to its return to the Building 221 canyons for reuse. Activity in the spent solvent is caused 

primarily by degradation of n-paraffin and Tri Butyl Phosphate (TBP) when solvent is exposed 

to high levels of radiation. The degradation produces complex zirconium and ruthenium which 

subsequently contaminate both solvent and product streams. One of the purposes of the 

segregated solvent facilities is to remove the degradation end products and radioactive 

contaminants from the solvent.

Transfer Tanks

Seven transfer tanks provide intermediate pumping stations for solutions transferred from 

Building 221-H and are located in open basins, each containing a sump and a pump. The 

basins are considered a place of potential exposure to airborne radioactivity.

Sump Collection Tank

This tank receives radioactive waste condensate by steam jet from sumps and tanks related to 

the canyon air exhaust system and send it by jet stream to other tanks that feed the hot or 

warm canyon waste evaporators.

Recycle Sump

This sump collects the drainage and overflow for all 221-H tanks that contain contaminated or 

recycled liquids.

Recycle Vent System

The recycle vent system provides a means to filter contaminated air from tanks and vessels 

prior to discharge to the atmosphere.

Auxiliary Systems and Support Facilities

The process well water system is the only system of concern for these facilities. There are two 

cooling water streams that are of importance, the segregated cooling water stream and the 

recirculating cooling water stream. Both streams are continuously monitored for radioactivity 

and can be diverted into retention or seepage basins to prevent a release of radioactive 

material.

Internal exposure. The radionuclide of concern for the A-Line is enriched uranium: 234U, 235U, 236U, 

and 238U.
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External exposure

The radionuclides of concern for all other facilities are 95Nb, 95Zr, 103Ru, 106Ru, and 137Cs.

. The radionuclide of concern for the A-Line is natural uranium and the major 

contributor to the external dose will be photons and x-rays from 234U, 235U, and 236U.

A-19 Receiving Basin for Off-site Fuel (RBOF) and Resin Regeneration Facility (RRF)

The RBOF facility handles casks and fuels of various shapes, sizes, and content. Casks of spent fuel 

are received and washed in the cask basin to remove any dirt accumulated during transit. After the 

cask is loaded into the fuel basin and the loaded fuel basket within the cask is removed and 

transferred to a fuel transfer bucket, the fuel transfer bucket then moves the fuel underwater to either 

the inspection basin, disassembly basin, repackaging basin, or storage basin. There are facilities for 

inspecting the fuel by mechanical and optical means, disassembling fuel, cutting fuel by cross cutting 

or slitting, and weighing fuel before and after packaging or other alterations.

The RRF is capable of regenerating both anion and cation type resins. The mixed resin is first 

depleted with NaOH and then the cation and anion resins are separated after which they are 

regenerated, rinsed, and returned to the deionizer. Other operations performed at RRF include the 

removal and disposal of the filter cake from the portable 100-Area filter, chemical cleaning of the filter 

stones in the portable 100-Area filter, and chemical cleaning of target slugs prior to processing in 

Building 232-H.

The irradiated fuel received at RBOF contains uranium and fission/activation products. Since the work 

is performed underwater, the inert gases that escape the fuel bundles are of primary radiological 

concern.

In addition to irradiated fuel, RBOF performs other operations such as basin purification, resin 

regeneration, and target cleaning. It was determined that tritium, fission/activation products, and 

plutonium could be present during these activities.

Internal exposure. The mixture of cation and anion resins in the RRF contains fission and activation 

products. The radionuclides of concern are 134Cs, 137Cs, and 60Co.

External exposure

A-20 Tritium Facilities

. The RRF receives a mixture of anion and cation exchange resins from the basin 

waste deionizer. The resins contain fission and activation products. 

The Tritium Facility includes Buildings 232-H, 233-H, 234-H, and 238-H. Irradiated lithium targets are 

processed in the tritium facilities to recover, purify, and package tritium.

The radionuclides of concern for internal exposure are elemental tritium and tritiated water. In 

addition, the activation product, 65Zn is associated with the processing of lithium targets in Building 

232-H. The activation product 65Zn is assumed to be the primary radionuclide of concern for external 

exposure at the tritium facilities. 

A-21 M Area – Uranium Target Fabrication Facility, 313-M

The Uranium Target Fabrication Facility was designed and built to manufacture aluminum clad targets 

for irradiation in the SRS reactors.

Depleted uranium metal, received from the Fernald plant, was fabricated into target rods that were 

used for transmutation of 238U to plutonium.
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There is no information on the distribution of radionuclides in the 313-M material, and the mixture 

given in Table A-21 is assumed. 

Internal exposure. The majority of the annual internal dose from this uranium mixture is delivered by 

238U. 

External exposure

A-22 320-M, 322-M, and 341-M 

. The majority of the external dose from the uranium mixture is delivered by 238U 

which decays to 234mPa which emits a high energy beta.

Process descriptions for these facilities are not available. However, target extrusion was conducted in 

Building 320-M, and Building 322-M was a Chemical and Metallurgical Laboratory. The M-Area 

facilities processed uranium, lithium, and aluminum into fuel and target components for the nuclear 

reactors. Processing included aluminum alloy formation, degreasing, etching, metal extrusion, hot-die 

size bonding, and nickel plating.

Uranium and/or enriched uranium are assumed to have been processed. The assumed distributions 

for depleted and enriched uranium are given in Table A-22. All radionuclides given in the tables are of 

concern for both internal and external exposures.

A-23 Fuel Fabrication facility, 321-M

The Fuel Fabrication Facility was designed and built to manufacture aluminum clad fuel elements for 

irradiation in the SRS reactors. In this process, enriched uranium metal is alloyed with aluminum, in 

concentrations required for reactor irradiation. The alloy is cast into hollow cylindrical ingots from 

which pre-extrusion billet cores are machined. These cores are extruded into logs, which are then 

machined into sections, encased in aluminum, and co-extruded into tubes. Cores for other types of 

tubes are fabricated and assembled into billets in other SRS facilities (Li-Al in Building 320-M and 

NpO2-Al in Building 235-F) before they are received at 321-M. The materials processed in Building 

321-M undergo the following processes: charge preparation; U-Al alloy storage; and U-Al casting, 

core machining, billet assembly, and billet outgassing.

Three radionuclide sources were identified in Building 321-M: uranium, plutonium, and neptunium. 

The majority of the annual internal dose from the identified uranium mixture is delivered by 234U. 

Plutonium was also processed in 1980. 

The majority of the external dose from the uranium mixture is delivered by 234U, 235U, and 236U.

When neptunium targets are prepared, photoemissions from neptunium and its daughters (i.e., 233Pa) 

present a much higher exposure potential than is normally encountered with uranium based targets.

A-24 S-Area – Defense Waste Processing Facility (DWPF) 

Radioactive wastes are stored in existing H-Area waste tank facilities in the form of settled sludge, 

supernatant liquid, and salt cake. The insoluble solids (sludge) is chemically treated and washed to 

remove aluminum in preparation for the feed to DWPF. The aqueous fraction undergoes treatment 

with sodium tetraphenylborate to precipitate the cesium and potassium salts. Also, a slurry of sodium 

titanate is added to absorb the soluble strontium and plutonium. The precipitate is washed and 

concentrated in preparation for feed to the DWPF. The aqueous product is subsequently blended with 

the formatted sludge feed and adjusted with glass formers. The adjusted feed is vitrified to a 

borosilicate waste glass in a slurry-fed joule-heated melter. The waste glass is encapsulated in 

stainless steel canisters and stored onsite in the glass waste storage building.
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The radionuclide composition of the incoming DWPF feed stream may vary depending on which tanks 

of waste are being processed.

Two process streams represent the high and low ends of the range of 238Pu/90Sr concentrations.

Internal exposure

The precipitate-slurry feed stream (Stream 2) has a specific activity of 37 Ci/gal. The majority of the 

annual internal dose comes from the 137Cs and 144Ce. 

. The sludge-slurry feed stream (Stream 1) has a specific activity of 56 Ci/gal. The 

majority of the annual internal dose is delivered by 90Sr and 238Pu. 

External exposure

The precipitate-slurry feed stream has a specific activity of 37 Ci/gal with the majority of the external 

dose being delivered by the same radionuclides as from the sludge-slurry stream. 

. The majority of the external dose is delivered by 144Ce, 137Cs, 106Ru, 90Sr, 238Pu, 

and 241Pu in the sludge-slurry feed (Stream 1) .

A-25 Z Area

Waste tank supernate (low-level waste) is transferred from H Area to Z Area via pipeline. At Z Area, 

the salt solution (supernate) is mixed with cement, fly ash, and blast furnace slag; and the resulting 

grout is pumped to vaults where it sets into a saltstone monolith.

The radionuclide composition of the incoming feed stream may vary depending on which tanks are 

being processed.

Two process streams have been identified on the basis of which radionuclides would deliver the 

majority of the dose and which radionuclides could be used for tracers. The process streams are the 

decontaminated salt solution and the saltstone.

Internal exposure. The majority of the annual internal dose from the decontaminated salt solution is 

delivered by 238Pu and 106Ru. The majority of the annual dose from the saltstone is delivered by 238Pu 

and 106Ru. 

External exposure

A-26 A Area Building 773-A

. The majority of the external dose from the salt solution and the saltstone is 

delivered by 238Pu, 137Cs, 90Sr, and 106Ru.

The radionuclide distribution found in some of the various sections is not available, and the 

information provided is based on the type of work performed.

Building 773-A has many processes including the following:

Actinide Technology

This section provides technical support for the Pu and enriched uranium production facilities; 

Building 221-F, FB Line, Building 221-H, and HB Line. The material used here should be 

similar to that found in Building 221-F, FB Line, Building 221-H, and HB Line. The 

radionuclides of concern would include plutonium, neptunium, americium, and uranium. 
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Analytical Development

This section provides analytical support for research programs and conducts research and 

development in the area of process control and analyzer development. The material used in 

this section may include fission products, uranium, neptunium, curium, and/or californium.

Defense Waste Processing Technology

This section provides support and conducts research and development in the area of waste 

processing. The material used in this section would be similar to the material processed in the 

S and Z area and should contain principally fission/activation products and plutonium.

Hydrogen Technology

This section is involved in development of hydride process technology and testing materials for 

use in corrosive and hydrogen environments. The type of radioactive material used in this 

section is unknown.

Interim Waste Technology

This section is developing the technology for the Defense Waste Processing facility (DWPF) 

and the Effluent Treatment facility (ETF). Work is performed on supernate processing, 

precipitation of strontium and cesium, and filtration and acid hydrolysis of the resulting 

sludges. The characterization of waste and the investigation of the treatment of effluents by 

ion exchange and reverse osmosis are also being developed. The material used in this section 

is assumed to be similar to the material processed in the S and Z Areas and should contain 

principally fission/activation products and plutonium.

Environmental Technology

This section processes environmental samples. The material used in this section may contain 

129I and 14C.

Laboratory Services

The areas of primary interest in this section include the High Level Cells and the F-Wing 

Experimental Area. The High Level Cells provide the shielding and confinement necessary for 

analysis and testing of highly radioactive material. The F-Wing Experimental Area has 

research and pilot scale production facilities for work with alpha and neutron emitters such as 

239Pu, 244Cm, 243Am, and 252Cf.

Materials Technology

This section consists of a Robotics Development Laboratory, Powder Metallurgy Facility, Onsite Uranium Reprocess facility, and the Hot Machine Shop. The purpose of the Powder 

Metallurgy Facility is to develop a new process for fabricating reactor fuel tube cores. The Onsite Uranium Recycle Process facility is a prototype facility to assist in the design and 

operation of a plant scale Fuel Production Facility. The Hot Machine Shop provides a variety of 

functions and services including machining uranium metal. The material used in this section 

may include fission/activation products, tritium, and depleted and/or enriched uranium. 

A-27 Buildings 735-A and 735-11A

Buildings 735-A and 735-11A house the Radiological and Environmental Science facilities. It is 

assumed that mainly low-level environmental samples are present in these facilities.

A-28 Building 776-A Liquid Waste Handling Facility

Located in Building 776-A, this facility collects the aqueous waste from Building 773-A. Facilities in 

this building are provided to strain solids from the waste streams, collect waste in receiving tanks, 
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sample and adjust the pH of the waste tank contents, and transfer the waste to tank trailers for 

disposal. The radionuclides are believed to be a composite of all the material that is used in Building 

773-A operations.

A-29 D-Area Heavy Water Production and Reprocessing

A heavy water production plant in D Area began operation early (1952) in SRS history to concentrate 

heavy water from Savannah River water to moderate and cool the Site’s reactors. The facility stopped 

production in 1981 because there was a sufficient supply of heavy water.

A-30 CMX and TNX (TNX Semi-works)

The very first facilities to be finished and put in operation at the Savannah River plant were two 

technical development facilities; CMX (1950-1984) was a reactor support facility to determine 

whether Savannah River water would present any unexpected problems in reactor cooling and TNX 

provided a wide range of technical support and development for Separations.. “CMX” and “TNX” were 

purely code designations and had no logical derivation.

CMX was used to flow test all new types of fuel and target assemblies before they were approved for 

use in the reactors. CMX operated from 1952 to 1984.

TNX (also called the TNX Semi-works) (1952- 2003) was used as a staging area for receipt and 

testing of large process equipment destined for use in production facilities within the SRS. Since 

1978, the TNX Area expanded from 3 original buildings to 32 buildings. The 32 buildings include 

administrative offices, process buildings for large scale experimental demonstrations, laboratories for 

both research and analytical work, pilot scale facilities, bulk tank storage, industrial wastewater 

processing facilities, and warehouse storage for a wide range of chemical and specialty equipment. 

TNX is no longer in existence and is well into the decommissioning process. 

Historically, some areas within buildings 672-T, 677-T, and 678-T contained radioactive materials. 

Building 672-T, constructed in 1983, consisted of a chemical processing facility, an electrical and 

instrumentation shop, and a control room. The chemical processing facility contained the Integrated 

DWPF Melter System, the Slurry Receipt Adjustment Tank/Slurry Mix Evaporator (SRAT/SME), the 

Ion Exchange Facility (IXF), and the Americium/Curium (Am/Cm) Melter.

Building 677-T, consisted of a maintenance shop, a laboratory where small-scale demonstration 

projects were operated, a bioremediation laboratory, and a bay area where larger-scale pilot projects 

were demonstrated. Equipment in Building 677-T consisted of the Plutonium Waste Incinerator (PWI) 

and Mobile Offgas System, the Uranium IV (U IV) facility, the Geometrically Favorable Dissolver, and 

the Centrifugal Contactor. Equipment and surrounding facilities housing the Geometrically Favorable 

Dissolver and the Centrifugal Contactor were radioactively contaminated. The U IV facility has a 

process history of radioactive work.

Building 678-T, and the 678-T Annex, constructed in 1979, was a chemical processing facility and 

supporting laboratories. Equipment in 678-T included the Offgas Components Test Facility, the Mott 

Filter Test Stand, and a tank farm. Historically, the tank farm in this facility was used to store raw 

material feed makeup for the Precipitate Hydrolysis Experimental Facility (PHEF), for development of 

melter feed, and for storage of aluminum nitrate for calibration of separations frames columns. 

Building 678-T was the former site of an evaporator which exploded in 1953. The tank farm, the 

former site of the evaporator, the laboratory, and a concrete pad outside the west wall of the building 

were all radioactively contaminated.
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Buildings 677-T and 678-T were contaminated with depleted uranium (EA-1231, July, 1998).
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2.0

SRS required pre-employment and annual physical examinations as part of the occupational health 

and safety program. These medical examinations, which were required as a condition of 

employment, typically included diagnostic chest x-rays. Doses from these required diagnostic x-ray

procedures depended not only on the characteristics of the x-ray machine and the procedure used, 

but also on the frequency of the examination. 

OCCUPATIONAL MEDICAL DOSE

2.1 EXAMINATION FREQUENCIES

The frequency of the required x-ray examinations differed significantly over the years and may have 

occurred more frequently than on the schedule indicated in Table 2.1. This information should be 

recorded in the file provided by DOE. All required diagnostic radiographs in the worker file are to be 

included in the reconstruction of occupational dose.

Table 2.1. Frequency of occupational chest x-rays at the SRS.

Period Frequency Comment

1950–1988 Annually All employees.

Annuallya Construction asbestos workers.

1989–1993 Annually Employees over 50 years old.

Biennially Employees 40 to 49 years old.

Every 3 years Employees under the age of 39.

Biennially Asbestos workers.

Annually DOE, USFS and WSI employees

1994–1998 Biennially Employees over 50 years old.

Every 3rd year Employees 40 to 49 years old.

Every 5th year employees under the age of 39

Biennially Asbestos workers.

Annually DOE, USFS and WSI employees

1995 Biennially Pensioners with documented uptake of 5 rem

1999–2002 Every 5th year Employees in a health surveillance program

Annually DOE, USFS and WSI employees

Biennially Asbestos workers.

a. The annual chest x-ray was a minimum received by construction and asbestos 

workers. Individual workers may have received chest x-rays as often as every 6

weeks. This would be recorded in the file provided by DOE on the individual worker.

X-ray frequency was determined by the age of the employee and the type or classification of the 

employee. Table 2.1 lists the frequencies of chest x-rays for different age groups through the years 

and also identifies specific groups of workers. Specific organ doses to be attributed for PA chest xrays calculated on the basis of the dose conversion factors found in ICRP Publication 34 are given in 

Attachment B, Table B-1. For lateral chest radiography the organ doses can be determined by 

increasing the PA chest entrance air kerma by a factor of 2.5 times to determine the lateral entrance 

air kerma. This air kerma value, when multiplied by the appropriate dose conversion factors (DCF) for 

mGy per Gy from Tables A2 through A8 in ICRP 34 (1982), provides the specific organ doses for the 

lateral view . These organ doses are somewhat higher than those from the corresponding PA doses 

primarily because of the greater mAs exposure per radiograph and the somewhat smaller Source to 

Skin (SSD). 

For organs not listed in ICRP Publication 34 but specified in the IREP code, doses were determined 

by analogy with anatomical location (Table 2.3-1). Thus IREP code organs in the thoracic cavity, but 

not mentioned in the ICRP Publication 34 were assigned the same dose as the lungs; doses to the 
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organs in the head and neck were assigned the same dose as the thyroid. The head and neck organ 

dose estimates (i.e. eye/brain), should be somewhat greater than doses actually incurred (hence 

claimant favorable), because of geometry considerations and at least in the case of brain, because of 

attenuation by the bony cranium. To ensure claimant favorability in the view of the variations in organ 

dose described in ICRP Publication 34 (p. 51), the doses for females, which are slightly higher than 

those for males, were used. 

2.2 EQUIPMENT AND TECHNIQUES

2.2.1

Although the Savannah River Site (SRS) medical practices are assumed to have followed the 

adoption of standards of radiology practice during the 1930s and 1940s to minimize dose to the 

patient, there is the potential for significant dose from occupational medical x-ray examinations 

depending upon the type of equipment, the technique factors, the number of photo-fluoroscopic 

examinations typical in the early years (Cardarelli 2002) and the number of radiographic 

examinations. SRS medical records include notations in individual worker files regarding both the date 

and the purpose of the x-ray examinations. 

X-Ray Apparatus

X-ray organ dose estimates for occupational x-rays administered at the SRS are made for Type I 

equipment (used from 1950 through 1970), Type II equipment (used from 1971 to July 1985), Type III 

equipment (used from August 1985 to May 1999), and Type IV equipment (used from June 1999 

through 2002). A description of the x-ray equipment used at the SRS is included in Table 2.2-1. The 

specific technique factors for these machines are in Table 2.2-2. Since no technique factors were 

identified by the SRS for Type I equipment, organ doses based on assumed technique factors, were 

developed on the basis of x-ray techniques contemporary with the time period (i.e. 1950-1970) with 

due consideration given to claimant favorability. Accordingly, it was assumed that the operating kVp 

was 80, somewhat higher (and hence claimant favorable) than the kVp values typically used at the 

time (Glasser et al. 1952; Morgan and Carrigan, 1955; Laughlin et. al. 1957). To offset this lower kVp, 

an increased exposure would be required as compared with Type II exposures, and this was 

conservatively taken to be 30 mAs, based on exposure time of 0.1 second at 300 mA. External 

filtration was assumed to be 1.5 mm Al as compared with 2.5 mm Al used subsequent to 1970, which 

would further increase the dose. The SSD was assumed to be 152cm based on a SID of 183cm less 

a chest thickness of 26cm and an addition of 5cm to account for cassette thickness.

Table 2.2-1. Description of the x-ray equipment used at SRS.

Technique Period Equipment

Type I 1950–1970 Dupont 2DC Safety Screens, Dupont Cronex 7 film, Picker X-ray Tube, No 

grid, Manual/Hand processing, Manual collimator.

Type II 1971–7/1985 Dupont Daylight Hi-Speed Screens, Dupont Cronex 7Film, Picker X-ray

Tube, Stationary 12:1 Grid, PAKO 3-Minute Film Processor, Manual 

Collimator.

Type III 8/1985–5/1999 Dupont Daylight Cronex 10 TL Film, Quanta III 

Screens, TecRad Manual Collimator, Technomed Recipromatic Upright

Bucky 12:1 Grid, Kodak M6AW 90 sec. X-ray Film Processor, Eureka X-ray

Tube.

Type IV 6/1999–2003 AGFA Cronex 10TK Film, AGFA Curix Cassettes-Ortho Regular Screens, 

AGFA Multiloader Processor, TecRad Manual Collimator, Eureka X-ray

Tube Rad 68, Technomed Recipromatic Upright Bucky 12:1 Grid.

For both posterior/anterior and lateral views, a standard source to image distance (SID) of 72 inches 

(183 cm) was maintained for all types. Additional information indicated that all of the x-ray machines 
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Table 2.2-2. Technique factors used for each type of X-ray equipment.

Machine View Current (mA) Voltage (kVp) Exposure time (sec.)

Type I Techniques unknown. No records were kept with regard to the 

technique factors used during this time period. The Radiologic 

Technologists employed during this period are all deceased; 

therefore, no known resources exist.

Type II PAa 300 110-120 1/30

Type III PA 300 120 1/40

Type IV PA 300 120 1/40

a. PA indicates a posterior/anterior view, the average PA chest measures 26 cm. 

The average Lat. chest measures 34 cm.

were single phase and that there was no air gap between the patient and the film. It is assumed in 

this TBD that all equipment had an HVL of 3.5 mm Al eq. (see Table B.2 of NCRP 102). Tables B-2 

lists organ doses for PA 14” × 17” chest films.

Although the SRS has not provided any information on early medical x-rays, there is indication that a 

mobile x-ray unit was used that may have resulted in higher doses. DPSP-57-01-11 indicates that an 

evaluation of the mobile x-ray unit was conducted and that full size chest radiographs were 

recommended to reduce personnel exposures. This implies that this unit may have delivered higher 

exposures, which may have been a photofluorographic machine using 70 mm film.

2.2.2

Collimation is an important determinant of organ dose. As indicated in Table 2.2-1, procedures at 

SRP called for the use of manual external collimation, and there is considerable indirect evidence and 

documentation to support the conclusion that beams at SRP were in fact externally collimated since 

the beginning of operations, the assumption has been made that there was in fact no collimation or 

cones used prior to 1970, i.e., for Type I machines and this is reflected in the dose calculations below. 

Collimation

2.3 ORGAN DOSE CALCULATIONS PA AND LATERAL(LAT)

ICRP Publication 34 (1982) provides tables of average absorbed dose (mGy) in selected organs for 

selected x-ray projections at 1 Gy entrance kerma (i.e., air kerma without backscatter), for selected 

views (including PA), and for selected beam qualities (i.e., various HVLs). These tables provide the 

basic dose conversion factors for converting air kerma to organ dose. Air kerma was obtained from 

Table B.3 of NCRP Report 102 (1989). The average air kerma rates for the different machines are 

calculated using the cGy per mAs provided in NCRP Report No. 102 (1989) for specific voltage, 

current, phase of the machine, and distance to the film. 

Finally, the SRS organ doses are found by multiplying the ICRP 34 (1982) organ dose conversion 

factors (DCF) in Tables A2-A8 by the air kerma dose in mGy (without backscatter) by the DCF for 

1Gy entrance kerma. The resulting x-ray organ dose conversion factors and doses for all machines 

are in Attachment B, Tables B-2 and B-4. The doses are shown in units of rem, assuming a quality 

factor of 1.0 for x-rays. SRS records will indicate the view and in most cases only one view was taken 

per medical examination. 

PA View X-Ray Dose Determination

The following formulas were used to calculate the PA x-ray procedures doses in Table B-2.

Based on the techniques in Table 2.2-2, the mAs for the types of equipment used were calculated for 

each view:
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The calculation method for entrance skin kerma for PA prior to1970 is as follows: from NCRP Report 

N0. 102. 

For Type I equipment assumed to be operating at 80 kVp, the entrance kerma is 0.15 cGy/100 mAs, 

(from NCRP Report 102 1989). 

The Current (mA) × Exposure Time (sec) = Current for View (mAs) (2–1)

Example for Type I PA view: 

300 mA × 1/10 s = 30 mAs 

30 mAs exposure x 0.15 cGy/100 mAs = 0.045 cGy at 183 cm. 

Correcting for distance (183/152)2 and filtration (40% for reduction from 2.5 to 1.5 mm Al) yields an

entrance skin kerma of 0.108 cGy.

[(183/152)2 x (0.045 cGy)/0.6 = 0.108 cGy] (2-2)

For the Type I time period, this approach gives values that are comparable with those reported in 

Laughlin et al. 1958, Braestrup and Wycoff 1958, and elsewhere. Note that this approach is 

reasonable when compared to other DOE sites (e. g. Hanford) where measurements are available for 

x-ray equipment from the early time period.

Example for Type II PA view:

Current (mA) × Exposure Time (sec) = Current for View (mAs)

300 mA × 1/30 s = 10 mAs (2–3)

The air kerma rate for 120 kVp was determined to be 0.3 cGy per 100 mAs (see Table B.3 of NCRP

Report No 102) and the air kerma was calculated after converting the rate to air kerma per mA.

Current for View (mAs) × Corrected Air Kerma Rate (cGy/mAs) = Air Kerma (cGy) (2–4)

10 mAs × 0.003 cGy/mAs = 0.03 cGy

The air kerma was corrected for the thickness of the chest (26 cm) and for distance between the chest 

and the plane of the film (5 cm) to obtain the air kerma at skin entrance.

Air kerma at 183 cm × SID squared ÷ SSD squared = air kerma at skin entrance (2–5)

Example for Type II PA view: 0.03 cGy × (183 cm)2 ÷ (152 cm)2 = 0.044 cGy

Air kerma at skin entrance was multiplied by the dose conversion factors in Table A.2 through A.8 of 

ICRP Publication 34 (1982) for PA chest and an HVL of 3.5 mm Al eq.

Air Kerma (cGy) × Dose conversion factor = Dose for View (cGy)
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Example for Type II PA view, dose to thyroid: 

0.043 cGy × 62 mGy/Gy × 1 Gy/100cGy × 1 rad/10 mGy = 2.7 E-3 rad (rem) (2-6)

Example for Type III and IV PA view:

Current (mA) × Exposure Time (sec) = Current for View (mAs) (2–7)

300 mA × 1/40 s = 7.5 mAs 

The air kerma rate for 120 kVp was determined to be 0.3 cGy per 100 mAs (see Table B.3 of NCRP 

Report No.102) and the air kerma was calculated after converting the rate to air kerma per mA.

Current for View (mAs) × Corrected Air Kerma Rate (cGy/mAs) = Air Kerma (cGy) 7.5 mAs × 0.003 

cGy/mAs = 0.023 cGy (2–8)

The air kerma was corrected for the thickness of the chest (26 cm) and for distance between the chest 

and the plane of the film (5 cm) to obtain the air kerma at skin entrance.

Air kerma at 183 cm × SID squared ÷ SSD squared = air kerma at skin entrance (2–9)

Example for Type III and IV PA view: 0.023 cGy × (183 cm)2 ÷ (152 cm)2 = 0.033 cGy

Air kerma at skin entrance was multiplied by the dose conversion factors in Tables A2 through A8 of 

ICRP Publication 34 (1982) for PA chest and an HVL of 3.5 mm Al equivalent.

Air Kerma (cGy) × Dose conversion factor = Dose for View (cGy) 

Example for Type III and IV PA view, dose to thyroid: (2-10)

0.033 cGy × 62 mGy/Gy × 1 Gy/100cGy × 1 rad/10 mGy = 2.0 E-3 rad (rem)

Lateral X-Ray View Dose Determination

Organ doses for lateral (LAT) views for Type I -IV machines can be determined using the PA entrance 

air kerma dose times a factor of 2.5 to obtain the LAT air kerma entrance dose. Using ICRP

Publication 34 (1982) Tables A2 through A8, and the LAT view dose conversion factors listed for the 

individual organs times the 1Gy/100 cGy (1000mGy/1Gy) gives the specific organ dose in mGy. The 

organ dose in rad (or rem) is the product of 1 mGy/10 rad/mGy. The IREP Analogous organ dose can 

be determined for the LAT view using the same method as used for the PA view. These doses are 

also listed in Table B-2. The LAT doses were not distance corrected as the 2.5 factor is large enough 

to account for the decrease in distance

For Type 1 equipment used prior to 1970 and as recommended in ORAU-OTIB-0006 Rev 3.0, no 

collimation was assumed. Therefore, organs not normally in the primary beam for a PA chest exam 

were included in the primary beam by using ICRP 34 (1982) organ dose conversion factors for 

procedures where those organs would normally be included in the primary beam. For example, 

assuming minimal collimation the ovaries would be in the primary beam, and dose conversion factors 

for the abdomen were used. Abdomen dose conversion factors were used for ovaries, testes, uterus, 

and their analogues, and the cervical spine dose conversion factor was used for the thyroid and its 

analogues. However manual collimation was used for all x-ray procedures and apparatus for Type II 

through IV x-ray machines.
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Photofluorography

Photofluorography, also known as photoroentgenography, was utilized for routine chest radiography 

and, as is well documented in the literature, typically produces higher patient doses than conventional 

radiography (Braestrup 1958, p. 140; Laughlin et al. 1957; Moeller, Terrill and Ingraham 1953). A 

review of available records from the Savannah River Plant (SRP) indicates that on July 7, 1953 and 

on February 19, 1954, SRP Project 8980 – AXC 71255 issued specifications for a mobile x-ray unit 

with capability for 70 mm photofluorography and conventional 14 × 17 inch radiographs. According to 

the bid specifications “This mobile x-ray unit shall consist of a complete x-ray unit providing facilities 

for rapid mass chest x-rays on roll film ...and for conventional 14” × 17” radiography.” The 

specifications required a mobile x-rays unit capable of operation at 100 kVp and 200 mA, with eight 

timer steps ranging from 0.05 (1/20) second to 1.5 seconds in the conventional radiography mode, 

and an f 1.5 lens and automatic operation in the photofluorographic mode. Bids were received from 

Westinghouse, General Electric and Powers X-Ray Services. Powers X-Ray Services, who specified 

Picker x-ray equipment submitted the low bid that was accepted by SRS. Powers X-Ray Services 

placed an order on June 18, 1954, for a Picker Catalog #6042 unit which produced a collimated 

rectangular field measuring field measuring 16 x 16 inches at the patient and was equipped with an f 

1.5 lens with a 4 3/8 inch focal length for photofluorography. For conventional radiography, the unit 

was equipped with a stepped timer with times ranging from 1/20 to 5 seconds, an milliammeter with a 

range of 0-200 mA indicating capability of operation at 200 mA, and a 20 step kVp control with a 

maximum of 140 kVp.

The plan was to x-ray 10,000 to 14,000 employees annually1

Assistance was given the Project Department and the Medical Department in evaluating 

several proposed methods for reducing personnel exposures during routine chest x-rays. 

It was recommended that full size 14” × 17” chest radiographs be used in order to reduce 

personnel exposure.

under contract, using the mobile x-ray 

unit, but it is not known if there were conventional 14 × 17 inch radiographs performed in addition to 

the photofluorographic exposures. A very limited review of a few worker medical records indicates 

that the conventional 14 × 17 inch radiograph was the procedure of choice, and indeed used from 

1953 on as indicated by those records examined. Despite the inference that might be drawn from the 

limited examination of medical records, clearly the capability for photofluorography existed at SRP 

from perhaps as early as mid-1954, depending upon when delivery of the unit was effected and when 

it was placed into service, and at least two documents indicate that at least some photofluorography 

was carried out. The Works Technical Department Monthly Progress Report for November 1957 

(DPSP 57-1-11, p. 411) makes the following statement, reproduced in its entirety, with respect to the 

mobile x-ray unit:

A history of the site, published in 1990, while not explicitly identifying photofluorography, states:

For a time, chest x-rays of employees in plant operations were done with portable 

equipment in a van. This gave the employee higher radiation exposure than did the fixed 

installation in Central Medical and was discontinued. (Bebbington 1990, p. 180).

An internal document dated November 4, 1966, dealt with the calibration of medical x-ray units at 

SRP (Cooley 1966). This document makes mention of the Picker mobile unit, which was then not in 

use, and recommended recalibration of the machine in January 1967 when it would again be used. 

Review of the measurement data in this report is indicative of conventional radiography; indeed, the 

 1. SRS Subcontract Requisition dated July 7, 1953, Order No. AXC 71255, with Powers X-Ray Products, Inc. specifies 

taking, developing and interpretation of 70 MM chest X-Ray pictures for approximately 10,000 to 14,000 SRS 

employees, based on their bid proposal Dated June 30, 1953. 
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only measurements reported were for the stationary Westinghouse unit in building 719-A. However, 

SRS used 100 mm film (2.75” X 3”) from 1951 through 1957 which indicates the use of 

photofluorography. 

In any case, it is reasonable to presume that at least some of the medical diagnostic chest x-rays at 

SRP were accomplished by PFG, but specific SRS data regarding measured exposures or 

photofluorographic technique factors are not available. Therefore, a claimant favorable approach 

would be to assume that PFG was used. However, there may be evidence in the individual claimant 

medical file such as notations of technique factors or the presence of 14 x 17 inch films that indicate 

that conventional 14 x 17 inch radiography and not PFG was in fact used. 

Typical operating parameters reported for 1950’s photofluorography were 24 mAs at 83 kVp at a 

target to film distance of 36 inches (Braestrup 1958, p. 143), and 30 mAs at 90 kVp with a target to 

film distance of 40 inches and 2.4 mm added filtration. Added filtration of 2.5 mm Al was assumed for 

SRS PFG dose determinations and is claimant favorable. While kVp and mA settings could be 

manipulated by the technician, the exposure time was regulated by the amount of light generated in 

the PFG unit, with a cutoff or maximum exposure time. An exposure of 15 mAs (150 mA for 0.1 

second) was sufficient to produce a satisfactory image on 35 mm film; larger film required greater 

exposures (Sante 1954, p. 129). The reported gonadal exposure doses from PFG equated to 0.15 

and 0.36 mrad, respectively, for females and males in the United Kingdom (Stanord and Vance 1957), 

and 1 and 2 mrad, respectively for females and males, in the American study (Laughlin et al. 1957). 

In another study in the U.S., (Feldman et al.) reported gonadal exposures equivalent to doses of 0.73 

for males and 15 mrad for females, the large difference being attributable to assumed collimation. 

Data in the contemporary literature indicate an ESE in the region of about 0.5 to 1 R (Laughlin et al. 

1957; Feldman et al. 1958; Moeller, Terrill and Ingraham 1953). Measurements at the Hanford site 

indicated that for a 60 mAs photofluorography exposure at 100 kVp, the ESE was 1.5 R, which is 

likely an upper limit value based on a large patient and is consistent with an ESE of about 600-700 

mR for a 24-30 mAs exposure at somewhat lower kVp. Thus, although the Hanford measured value 

is likely an upper limit and hence an overstatement of the actual exposure from photofluorography to 

the average patient, this 1.5 R ESE value was chosen for the SRS organ dose calculations for 

claimant favorability.

2.3.1

The organ doses for photofluorography chest radiography were determined using an ESE value of 1.5 

R for 70 mm photofluorography. Organ doses were calculated using the exposure expressed as ESE 

assuming 1 R equals1 rad and were made in accordance with conversion coefficients in Tables A2 

through A8 of ICRP Publication 34 (ICRP 1982). These tables list organ doses based on an SID of 

183 cm for chest radiography. Photofluorography was accomplished with an SID of 40 inches or 102 

cm. Given the geometric considerations regarding divergence of the beam as discussed in ICRP 

Publication 34 (pp. 23ff), this difference in distance would have a negligible effect on organ doses 

and, therefore, the values listed in Tables A2 to A8 are appropriate for chest photofluorography. 

Because absorbed organ dose for radiographs are a function of beam quality, an HVL of 2.5 mm Al 

was used for dose determination. 

Organ Dose Calculation Methods – Chest Photofluorography 

The following method was used to calculate the dose in Gy for the period from 1951 to 1957 using the 

dose of 1.5 R. 
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1951 to 1957 70 mm chest photofluorographic x-rays with 2.5-mm Al filter and a measured ESE 

of 1.5 R. Absorbed dose in Gy was determined by 1500 mR/100 mR per mGy = 

15.00 mGy and 15.00 mGy/1000 mGy per 1 Gy = 0.0150 Gy. This value in Gy is 

used with Table B-3 dose conversion factors (DFs) (mGy per Gy air kerma for a 

beam quality for 2.5 mm Al HVL). These DFs were used with Table B-3 values to 

determine doses for the listed organs in ICRP Publication 34. The doses for the 

IREP listed organs that are not ICRP 34 (1982) listed were determined using 

analogues organs (Table 2.3-1). Organ doses are shown in Table B-4.

Table 2.3-1. Analog for IREP organs not included in ICRP 34.

Anatomical 

location

ICRP 34 

reference organ IREP organ analogues

Thorax Lung Thymus

Esophagus

Stomach

Bone surface

Liver, gall bladder, spleen

Remainder organs

Abdomen Ovaries Urinary/bladder

Colon/rectum

Head and neck Thyroid Eye/brain

This Table is copied from Table 3.1-1 Analogs for IREP organs not included in ICRP Publication 34 

(1982), (Kathren et. al. 2003).

2.4 UNCERTAINTY

Error, defined as deviation from the correct, true or conventionally accepted value of a quantity, and 

uncertainty, defined in terms of the potential range of a stated, measured, assumed or otherwise 

determined value of a quantity, provides an indication of the confidence of the dose estimates. Error 

implies knowledge of what the correct or actual value is, which is, of course, not known. Hence a 

more appropriate term is uncertainty, which is expressed in terms of a confidence level, e.g. 99% (i.e. 

that the correct or true value, although not actually known, has a 99% probability of falling within the 

range cited) and includes both precision or reproducibility of the measurement and accuracy, or how 

close the measurement or estimate of dose comes to the actual or correct value.

Although in theory a large number of factors can introduce uncertainties or affect the x-ray machine 

output intensity and dose to the patient, in practice only four factors can be reasonably considered to 

have an impact on dose uncertainty. These are 1) variation in applied kilovoltage, 2) variation in 

beam current,3) variation in exposure time, and 4) distance from the patient to the source of the xrays (SSD). The influence of such other factors as use of screens, grids, reciprocity failure, film 

speed and development, while potentially variable, would not affect the beam output intensity.

For a given set of machine settings and parameters, x-ray output should theoretically be constant and 

unvarying. However, this is not true in practice; although output is essentially constant unless focal 

spot loading occurs such as might be the case when the power rating of the machine is exceeded. It 

is unlikely that power ratings were ever exceeded since so doing would be difficult to achieve in 

practice and would result in damage to the x-ray tube. However, even with the use of so-called 

constant voltage transformers to control line voltages, slight variations may occur in line voltage input 

or other internal voltages, which in turn could alter the kVp of the output beam. In general, for a given 

kVp setting, variation in kVp falls within ± 5% of the machine setting. Since as noted above beam 
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intensity is approximately proportional to the 1.7 power of the kilovoltage, this translates to an 

uncertainty of approximately ± 8.7%with respect to output beam intensity in the 100 to 120 kVp range 

used for diagnostic radiographs at the Savannah River Site . For conservatism, this is rounded up to 

± 9%.

Similarly slight variations in tube current are normal; as a tube ages, or heats up from usage, tube 

current may change and typically will drop. Hence, all other factors remaining constant, beam 

intensity will be reduced, and in direct proportion to the change in tube current. Typically, the 

reduction in beam output from current variation is not more than a few per cent under normal 

operating conditions; large decreases in beam output will be readily detected and result in 

maintenance on the machine to restore the output, or, as a temporary stopgap measure, increase in 

the current or kVp to provide the necessary intensity for proper radiography. There is no evidence to 

suggest that these stopgap measures were ever necessary or applied at the Savannah River Site. 

For a given kVp setting, output of the beam is a function of the tube current, which in turn is measured 

by a milliammeter on the machine and measures average tube current. The measurement is subject 

to uncertainties, and in addition there may be minor changes in output as the tube heats up from 

normal usage. These variations are typically small, and hence uncertainty in beam output attributable 

to current variation has been estimated at ± 5%.

Another parameter that has potential to affect the dose, perhaps significantly, from a diagnostic 

radiograph relates to the time of exposure. This can be readily understood by noting that a full wave 

rectified machine produces 120 pulses per second of x-rays. For an exposure time of 1/20 of a 

second, only six pulses would result. A small error in the timer that resulted in a change of only ± 1 

pulse would correspondingly affect the output by ± 17%; for an exposure time of 1/30 of a second, the 

change in output corresponding to a deviation of ±1 pulse is ± 25%. Early mechanical timers were 

notoriously inaccurate, although timer accuracy improved significantly with the introduction of 

electronic timers. However, once again for conservatism, uncertainty in beam output attributable to 

timers will be assumed to have an upper limit of + 25%. 

The final factor that is likely to affect patient dose relates to distance from the source of the x-rays, 

which is a determinant of the entrance skin exposure. For a given individual, the SSD will be 

determined largely by the thickness of the patient, and how accurate the positioning is. For a typical 

patient, this variation in SSD is estimated at no more than a few centimeters, with an upper limit of 

perhaps 7.5 cm. Using inverse square law, this indicates an uncertainty of ± 10 % from this source.

There are two approaches to determination of the combined uncertainty from the above four potential 

sources of uncertainty. The first, and most conservative in that it gives the greatest range, would be 

to assume that the uncertainties are additive, which would give an uncertainty range of up to 9 

+5+25+10 = 49. However, a more reasonable approach would be to assume that the uncertainties are 

in fact random, and to compute the statistical root mean square (RMS) value. The RMS value is 

simply the square root of the sum of the squares, and computes as ± 28.7%. Thus, for any individual 

entrance skin exposure (ESE) or derived organ dose, an uncertainty of ± 30 % at the 99 % 

confidence level may be assumed; for further conservatism it may be appropriate to assume that 

errors are all positive, and only the + 30% should be used. 

2.5 DOSE RECONSTRUCTOR INSTRUCTIONS

The information given below summarizes instructions given to the dose reconstructors in determining 

organ doses from occupational medical (x-ray) procedures. For the purpose of evaluating probability 

of causation, x-ray doses are always considered acute, and are considered to be photons with energy 

E=30-250 keV.
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2.5.1

Organ doses are provided for ICRP Publication 34 (1982) listed organs and for IREP organs not listed 

in ICRP 34. The dose tables in Attachment B give specific values for organ doses from PA, Lateral 

and photofluorography procedures. The ICRP Publication 34 (1982) doses conversion factors are 

given in Tables B-1 and B-3. The analog organs are listed in Table 2.3.1 and the doses for these 

IREP organs that are not listed in ICRP 34 are also given in Tables B- 2 and B-4. 

Assignment of Organ Doses from X-ray Procedures: Maximizing Approach for Dose 

Reconstructions

For organ doses resulting in a Probability of Causation (POC) of less than 50%, the x-ray doses 

should be multiplied by a factor of 1.3 to insure that the highest doses to any organ in the Tables 

includes an uncertainty of +30% and hence are claimant favorable. Multiplying the organ doses by a 

factor of 1.3 accounts for uncertainty and maximizes the annual x-ray doses that can be used in dose 

reconstructions that are unlikely to be compensated. For organ doses that result in a POC of greater 

than 50%, each x-ray exam needs to be carefully scrutinized to assure that the assigned dose that 

contributes to the POC is reasonable and does not contain erroneous assumptions or gross 

overestimates. These x-ray doses presented are treated as a point estimate (constant) for the 

purpose of calculating probability of causation.

One annual x-ray procedure for each full or partial year of employment is assumed. However, if the 

dose reconstructor determines that more frequent procedures occurred or may have occurred, the 

annual organ doses can be raised accordingly.

2.5.2

The “best estimate” approach for dose reconstructions is appropriate in certain instances, and 

particularly in those cases where the POC of 50% would be exceeded if the maximizing dose method 

is used. To obtain the best estimate dose, the values obtained by the method above are divided by 

1.3, and the normal distribution is applied with a coefficient of variation (i.e. standard deviation 

expressed as a percentage) of 30%. The dose reconstructor should use the actual organ doses listed 

in Tables B-2 and B-4 and may use a frequency other than annual if the actual frequency of other 

exams are known

Assignment of Organ Doses from X-Ray Procedures: Best Estimate Approach for 

Dose Reconstructions

2.6 RECONSTRUCTION OF ORGAN DOSE FROM RADIOGRAPHY EQUIPMENT USING 

MINIMAL COLLIMATION

Prior to about 1970, x-ray measurement data, techniques, or beam port information may not be 

available to estimate the collimation of the x-ray beam. Several papers in the literature have 

considered the effects of cone size and centering on organ doses, and concluded that filtration, kVp, 

and the smallest possible cone size were most important to reduce these doses. Due to the reported 

variation in the literature and measurement data on the effects of collimation, it is claimant favorable 

to assume minimal or no additional external collimation was used when measurement data, 

technique, or other information to describe the collimation are not available for x-ray procedures 

performed prior to 1970. 

With minimal collimation, organs normally outside of the primary beam are exposed to the primary 

beam. This necessitates the use of dose conversion factors from ICRP 34 (ICRP1982) other than 

normal dose conversion factors for a PA or lateral (LAT) chest x-ray, since ICRP 34 (1982) dose 

conversion factors are based on properly collimated beams. For minimally collimated beams used 

prior to 1970, the substitute dose conversion factors in Table 2.6 below were used:
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Table 2.6. Substitute dose conversion factors.

Organ of interest

Substitute view and organ for which the dose conversion 

to use for minimally collimated beams

Thyroid AP Cervical spine corrected for depth by a factor of 0.2 (NCRP 102, Table B-8, P. 103)

Lat Cervical spine

Eye/brain PA and lat Skull, or PA chest, whichever is larger.

Ovaries and analogues, 

testes, and uterus

PA and lat Abdomen
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3.0

The occupational environmental dose refers to the dose received by workers outside of the facilities. 

These doses can be internal and external depending on the characteristics of the individual 

radionuclides. While most radionuclides when inhaled would represent a dose to various organs in the 

body, the noble gases would only represent an external dose. These radionuclides are addressed in 

the following sections.

OCCUPATIONAL ENVIRONMENTAL DOSE

3.1 INTERNAL DOSE FROM ON-SITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

The internal dose for workers outside of the facilities is determined from the air concentrations 

resulting from the individual facility releases, from ground level releases and from the resuspension of 

radioactive materials in the soil.

Unmonitored workers can be exposed to occupational doses internally from on-site releases into the 

air, from the resuspension of radioactive materials in soil, and externally from ambient radiation and 

releases of radioactive noble gases into air.

3.1.1

Although the annual and semiannual environmental reports contain air monitoring data, these data 

were determined to be of limited value because of the confounding effects of nuclear weapons testing, 

lack of spatial and radionuclide resolution, and insensitivity of sampling methods to radionuclides 

important to dose. Therefore, methods were developed to use well- documented source terms 

developed by others (Cummins 1991; CDC 2001), coupled with historical meteorology, to estimate 

radionuclide-specific airborne concentrations for 3

H, 131I, 41Ar, 238Pu/239Pu/240Pu, and 234U/235U/238U. 

These radionuclides were determined to account for more than 95 percent of the potential missed 

dose from inhalation and submersion pathways. Air concentration calculations included spatial and 

release height considerations for source terms from two processing areas (F- and H-Areas), the 

heavy water extraction area (D-Area), and five reactor areas (C-, K-, L-, P-, and R-Areas). The 50th 

and 95th-percentile air concentrations were estimated for the eight source term areas, seven other 

populated areas (shown in Figure 3.1.1-1), and a site average. These calculated air concentrations 

were the used to derive annual intakes based on an assumed individual ventilation rate of 2,400 cubic 

meters per year. In addition, the geometric standard deviations (GSD) of these intakes were 

estimated based on the 95th-percentile source terms (Rollins 2003) using the following relationship.

On-Site Releases to Air

(3–1)

These annual intakes and GSDs, by year of operation, are presented in Attachment C in Tables C-1 

through C-17. 

Atmospheric dispersion was modeled by developing atmospheric dispersion (X/Q) values using 5-

year average joint frequency distributions presented in the SRS environmental reports between 1975 

and 1999 (see reference section). These data were converted to X/Q tables for elevated and groundlevel releases, using the Multimedia Environmental Pollutant Assessment System (MEPAS) 4.0 

computer code (PNNL 1996) and plotted using an Excel® spreadsheet. To assure that atmospheric 

dispersion would not be overestimated and resultant air concentrations underestimated, the highest 

X/Q values in any direction over the 25-year period were chosen as representative for all years of 

( ) ( ) 1.65585

1

50 th percentile

95 th percentile GSD =
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Figure 3.1.1-1. The SRS showing the reactor (C-, K-, L-, P-. and R-Reactors), and 

processing areas (F Area and H Area) in a rough circle toward the center of the Site,

and D Area and A and M Areas near the Site perimeter. Also shown are DWPF (S 

Area), Saltstone (Z Area), Central Shops, Heavy Water Extraction (D Area), U.S. 

Forest Service Headquarters, and the Railroad Classification Yard.

operation. Studies performed at the SRS have indicated that the use of the highest X/Q values would 

not significantly over or under estimate actual atmospheric dispersion for all years of operation 

(Harvey 2001, Hamby 1991). Expressions were developed for ground level and elevated releases 

and used to estimate atmospheric dispersion based on distance from the source. These expressions 

were developed using the Excel® spreadsheet for all source-term locations (reactor and separations 

areas) from 500 m out to a distance of 80 km.

For ground level releases (5 m):

Y = 1.0146 X-1.8809 (3-2)
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For elevated releases (60 m):

Y = 0.0013X-1.1739 (3-3)

Where

Y = Atmospheric dispersion coefficient, s per m3

X = Distance from source in meters.

Distances between the populated areas were estimated using geographical information provided by 

Garmin (2001) and are presented in Table 3.1.1-1. These distances were used in the X/Q 

expressions and coupled with the adjusted source terms to estimate cumulative air concentrations for 

all 15 population areas evaluated (Rollins 2003 - TBD). 

Table 3.1.1-1. Distance, in meters, from source areas to population centers at SRS. Source

location

A

Area

C

Area

D

Area

F

Area

E

Area

H

Area

S

Area

Z

Area

R

Area

RR

Yard

P

Area

L

Area

K

Area

Forest

Service CS

C Area 11,500 500 7,500 3,800 4,100 5,400 6,600 7,500 9,400 12,500 9,400 6,600 4,300 13,500 2,200

F Area 8,100 4,000 10,600 500 1,600 3,500 3,900 4,700 9,100 13,900 11,300 9,700 8,300 9,700 6,500

H Area 10,500 5,500 12,900 3,500 1,900 500 1,200 2,000 5,700 11,100 8,900 8,800 8,900 8,800 6,400

K Area 15,800 4,300 6,300 8,200 8,200 8,800 10,100 10,800 10,500 11,300 8,200 3,900 500 17,500 6,900

L Area 17,800 6,600 10,150 9,600 8,900 8,600 10,100 10,500 8,100 7,400 4,400 500 3,900 17,600 7,600

P Area 19,100 9,400 14,400 11,100 9,900 8,800 9,900 10,100 5,300 3,100 500 4,400 8,100 16,900 9,600

R Area 15,900 9,400 16,300 9,100 7,500 5,700 6,200 5,900 500 6,100 5,300 8,300 10,500 12,000 9,400

For radionuclides potentially important for internal dose reconstruction, the average and maximum 

annual intakes and respective GSDs for all populated areas are provided in Tables C-16 and C-17, 

respectively. Area-specific annual intakes with respective GSDs are provided in Attachment C 

(Tables C-1 through C-15) for use where more detailed information is required. The values in these 

tables were derived from calculated concentrations for the major populated site locations based on 

historical on-site meteorology, distance from release points, and emissions data. 

The values in Tables C-1 through C-15 are estimates of annual intakes derived from the calculated 

outdoor air concentrations at SRS locations and, as such, may not be appropriate for calculating 

organ doses for workers located inside buildings. However, these values likely represent the upper 

bound of potential intakes for an individual working in outside areas where individual monitoring was 

not required. It is assumed that these intakes occurred uniformly in the year listed, so for partial years 

of employment, the intake may be proportionally adjusted by the fraction of the year worked. In 

addition, these intakes were derived based on an assumed ventilation rate of 2,400 cubic meters per 

year. 

For purposes of dose estimation, radioactive decay was not considered. This is considered a 

claimant-favorable assumption because it would tend to overestimate the actual air concentrations 

and, thus, the resultant radiation dose (see Section 3.3 for other uncertainty discussions).

Three situations are considered when reconstructing environmental inhalation doses for unmonitored 

individuals:

1. A worker spent significant time in a defined area for which data in Tables C-1 through C-15 

may be used to estimate intakes. These tables show the maximum annual intakes at the 50thpercentile and respective GSDs for 3

H, 131I, Pu, and U, in each SRS area by year.
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2. A worker was associated with multiple areas on the site best represented by the site average 

annual intakes in Table C-16.

3. A specific work location cannot be identified, so the maximum annual intakes in Table C-17 

(derived from Tables C-1 through C-15 in Attachment C) for each year of employment are

used to estimate the dose to the organ of interest.

It has been determined that defining which worker falls in which of the above categories is a very 

time-consuming and somewhat subjective part of the individual dose reconstruction process. 

Because the dose from environmental releases to air is generally fairly small and because of the 

uncertainty associated with applying these intakes to an individual, dose reconstructors are instructed 

to use only Table C-17, the maximum annual intakes, to assign environmental inhalation intakes. 

This is a claimant-favorable and reconstruction-expediting instruction. When there is indication that 

an individual may have inhaled radioactive material at SRS prior to 1955, the 1955 intake values from 

table C-17 are used to provide a claimant favorable estimate of intakes in earlier years. 

Based on currently available documentation, it does not appear likely that any internal fission product 

or plutonium exposure occurred prior to 1954. Records indicate that a tritium production facility 

became operational in 1952 and that uranium operations were occurring as far back as 1952. 

Because of the uncertainty regarding the start of environmental releases, it is assumed that intakes at 

the 1955 levels occurred as early as 1952. 

The following sections discuss details of specific radionuclides of importance for calculations of 

airborne concentrations.

Tritium

Tritium is one of the principal nuclear materials produced at the SRS. It is the heaviest and only 

radioactive isotope of hydrogen with a physical half-life of 12.5 years. Tritium processing operations 

have been performed at the SRS facility since October 1955, beginning in Building 232-F. In July 

1957, the process was moved to Building 232-H and operations doubled by 1958. 

Tritium is produced naturally by cosmic ray interactions and as a fission product in nuclear reactors. 

The nuclear reactions that produce tritium are:

• Fission of lithium by neutron irradiation of targets

• Reaction of neutrons with the heavy water moderator used in reactors

• Ternary fission of transuranic elements in the reactor fuel and targets.

At SRS, tritium was originally produced as a by-product of producing plutonium in the reactors. 

Lithium-aluminum control rods were used in R-Reactor and other reactors from the first day of 

operation to produce tritium and control reactor power. When additional tritium was needed beginning 

in 1955, enriched uranium was used in the reactors, which allowed for increased loading of lithiumaluminum targets for tritium production.

Gaseous tritium at room temperature tends to form elemental tritium (HT) by reaction with gaseous 

hydrogen. Tritium closely follows the reactions of ordinary hydrogen when it exists as tritiated water or 

vapor (T2O); tritium oxide [HTO]) is taken into the body quite easily, resulting in a higher factor of risk. 

Tritium is released from SRS in both elemental and oxide forms.
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The key processes that have led to tritium releases at the SRS are:

• Reactor operations (100 Areas)

• Recovery of transuranic elements in the separations facilities (200 Areas)

• Recovery of tritium in the tritium processing facilities (200 Areas)

• Laboratory research area (700 Area)

• Heavy water rework facility. (400 Area)

The majority of the tritium releases came from the reactor facilities (100 Areas) and the tritium 

processing facilities (200 Areas). For purposes of on-site dose reconstruction, only releases from 

these areas are considered because these releases account for the vast majority of the total tritium 

released to the atmosphere. Additional information, including histories and descriptions of operations 

at these facilities and the reconstructed atmospheric tritium source terms from 1955 through 1992, 

can be found in Savannah River Site Dose Reconstruction Project Phase II: Source Term Calculation 

and Ingestion Pathway Data Retrieval, Evaluation of Materials Released from the Savannah River 

Site, RAC Report No. 1-CDC-SRS-1999 (CDC 2001, Chapter 4.1)(hereinafter referred to as the RAC 

Report). Source terms from 1993 through 2001 were obtained from the SRS annual environmental 

reports (see reference section for citations). The 50th-percentile values were assumed to be those 

presented in the annual environmental reports. The 95th-percentile values were estimated by dividing 

the 50th-percentile yearly value by the smallest value of the ratio of the 50th to the 95th-percentile 

values for all of the years reported in the RAC Report (CDC 2001). This method provides reasonable 

assurance that the 95th-percentile values for 1993 through 2001 are not underestimated (see Rollins 

(2003) for details).

The uncertainty analysis applied to reported releases is described in Attachment E of the RAC Report 

(CDC 2001). The adjusted tritium source terms and uncertainty analyses (CDC 2001, Chapter 4.1) 

used to calculate atmospheric concentrations attempted to account for the insensitivity of the historical 

sampling methods and their inability to remove tritium oxide and elemental tritium effectively from the 

atmosphere. Table 3.1.1-2 presents a description of the monitoring equipment used to monitor 

atmospheric tritium releases.

Table 3.1.1-2. Overview of tritium monitors at SRS (Table E-1, CDC 2001, p. E-2).

Monitor Facility Period in use

Dehumidifier Reactors 1954–shutdown

Silica gel Reactors and disassembly areas 1954–1958 or until shutdown in some cases

Kanne ionization chambers Tritium facilities 1954–present

Stack tritium monitor (STM) Reactors 1970–1988

Berthold tritium monitor (BTM) Reactors 1988–present

Stack monitor integrator (SMI) Tritium facilities 1974–present

FORMS (STM with dehumidifier) Tritium facilities and reactors 1985–present

In the reactor areas, the primary atmospheric release points are the 200-foot stacks and ground-level 

evaporation from the disassembly basin water. An additional source of atmospheric tritium release is 

the ground-level evaporation of water purged from the disassembly basin to a seepage basin. Tritium 

in the form of DTO (from the moderator) is released to the ventilation system and, therefore, to the 

stack by evaporation of process water exposed to air flowing through the process area. During reactor 

operations, D2O containing DTO is evaporated from small leaks in pipe flanges, valves, and from the 

D2O process water exposed in the control rod guide tubes. To account for the ground-level releases, 

the conservative assumption was made (based on The RAC Report (CDC 2001, p. 4.1-8)) that 8% of 

the total tritium released from the reactor areas was from ground level. This additional amount was 

added to the adjusted source terms reported in the RAC Report (CDC 2001) and modeled as ground-
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level release. All releases from the processing facilities were modeled as elevated (200-foot) 

releases.

Although between 30% and 50% of all tritium released from SRS is thought to be of elemental form 

(RAC Report (CDC 2001, pp. 4.1-15, 16)), the annual intakes presented in Tables C-1 through C-15 

did not attempt to account for this. Therefore, the dose reconstructors should assume that the annual 

intakes consist entirely of the more biologically effective form of tritium oxide.

Radioiodine

This section discusses the releases of the most important radioiodine isotopes (131I and 129I) to the 

atmosphere from facilities at the SRS. The historic measurements at the facilities were interpreted in 

the context of current knowledge of radioiodine chemistry and the effectiveness of various sampling 

techniques (RAC Report (CDC 2001, Chapter 4.2-2)). These analyses that considered the chemical 

form, sample line losses, and efficiency of sampling methods resulted in revised estimates of 

radioiodine releases. These revised estimates, or radioiodine source-terms, were used to calculate 

air concentrations and annual intakes at 15 populated areas within the SRS. A detailed discussion of 

methods used to develop the revised source terms and the uncertainty analyses used to develop the 

95th-percentile values is provided in the RAC Report (CDC 2001, Chapter 4.2). These 95thpercentile values were used to derive the geometric standard deviations shown in Tables C-1 through 

C-17. An overview of the radioiodine sampling and monitoring methods used at the SRS is provided in 

Table 3.1.1-3.

Table 3.1.1-3. Overview of radioiodine sampling and monitoring methods [RAC Report (CDC 2001, 

Table 4.2-2)].

Period

Method of

sample collection Comments on reported results

Dec 1954–Oct 1956 NaOH scrubber Only reactive forms were measured; system was not 

considered reliable; no corrections for line loss were made

Oct 1956–Sept 1961 AgNO3-impregnated 

filters

Only reactive forms were measured; unknown variability in 

collection efficiency; no corrections for line loss were made

Sept 1961–Sept 1965 Charcoal cartridges All chemical forms were measured; attention paid to 

collection efficiency variations; no corrections for line loss 

were made

After Sept 1965 Charcoal cartridges New monitoring system with higher flow rate; all chemical 

forms were measured; attention paid to performance; no 

corrections for line loss were made

Revised radionuclide source terms were developed for both organic and inorganic chemical forms. 

Although organic radioiodines comprised a large fraction of the total radioiodine released to the 

atmosphere, for purposes of on-site dose reconstruction, the calculated annual intakes presented in 

Tables C-1 through C-17 are based on total radioiodine released without distinction to chemical form. 

This assumption is deemed appropriate because the major differences in the biological effectiveness 

of the chemical form occur primarily in the air-grass-cow-milk pathway, which is not considered for onsite dose reconstruction.

The fuel processing facilities in F Area and H Area were the sources of the highest releases of 

radioiodines, which were discharged through tall (60-m) stacks. However, for completeness, elevated 

releases of radioiodines from the reactor areas were also included in the air concentration calculations 

and resultant annual intakes. The largest reported releases occurred in 1956. Releases during the 

years 1955–1961 were significantly higher than the releases that occurred between 1962 and 1971. 

Those latter releases were, in turn, markedly greater than releases in the later 1970s and 1980s.
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The revised source terms from the processing facilities were provided for the combined releases from 

F and H Areas (RAC Report (CDC 2001, Chapter 4.2)). To improve spatial resolution of the 

calculated annual intakes, these combined source terms were partitioned into individual area releases 

by comparing each area’s uranium through-put to the total uranium through-put (RAC Report (CDC 

2001, Chapter 4.2 data files)) for both reprocessing areas. 

Similarly, revised source terms were provided for combined releases from the five reactor areas. 

Since there were no quantitative measurements of iodine releases to the atmosphere from the reactor 

areas prior to 1972, revised estimates were developed based on radioiodine measurements in reactor 

liquid effluents and later measurement of ratios of atmospheric to liquid releases. To improve spatial 

resolution of the on-site calculated air concentrations and resultant annual intakes, individual reactor 

atmospheric releases were estimated based on ratios of reactor-specific radioiodine liquid releases to 

the combined source terms for all reactors.

Annual intakes of 129I were not calculated because the quantity of 129I released to the atmosphere was 

several orders of magnitude lower than that of 131I and therefore not deemed important to on-site dose 

reconstruction. 

Plutonium and Uranium

Plutonium and uranium releases to the atmosphere represent the greatest potential radiological risk of 

all the alpha-emitting radionuclides in the SRS environment. Although releases of these radionuclides 

were reported for the reactor areas, M Area facilities, and the Administration Area (A Area), the 

majority of the releases were reported for the two chemical separations areas canyon facilities (221-F 

and 221-H). Because releases of plutonium and uranium from F- and H- Areas represent more than 

97% of the total atmospheric releases of all alpha-emitting radionuclides, air concentrations and 

resultant annual intakes were calculated only for uranium and plutonium released from the chemical 

separations areas. The slight underestimation of air concentrations and potential doses resulting from 

this simplifying assumption is considered to be well within other identified uncertainties.

In general, from 1955 through 1989, the source terms reported by Cummins (1991) for plutonium and 

uranium were determined to be the best available data (RAC Report 2001 (CDC 2001)). 

Uncertainties were estimated (RAC Report (CDC 2001, Chapter 4.4)) based on sample collection and 

counting procedures, and estimated possible sample losses during transmission through sampling 

probes and lines. These reported values and their estimated uncertainties were used to calculate the 

50th-percentile annual intakes and geometric standard deviations presented in Tables C-1 through C17. 

Source terms from 1990 through 2001 were obtained from the SRS annual environmental reports 

(see reference section for citations). The 50th-percentile values were assumed to be those presented 

in the annual environmental reports. The 95th-percentile values were estimated by dividing the 50thpercentile yearly value by the smallest value of the ratio of the 50th to the 95th-percentile values for all 

of the years reported in the RAC Report (CDC 2001). This method provides reasonable assurance 

that the 95th-percentile values for 1989 through 2001 are not underestimated.

To simplify the analysis and because isotopic data are limited, calculated annual intakes were 

developed for total releases of uranium and plutonium. Isotopes assumed to be present in the mix 

include 238Pu/239Pu/240Pu and 234U/235U/238U. For purposes of dose reconstruction, the isotope of 

plutonium and uranium contained in the mix that results in the maximum organ dose should be 

selected and assumed to be present at 100%. This assumption will result in a small, but claimantfavorable, overestimation of the actual dose.
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3.1.2

Soil sampling and analysis were not routinely performed at the Savannah River Site during the period 

of greatest atmospheric releases (from 1955 through the late 1960s). However, several special 

surveys were performed between 1958 and 1970, but methods for soil sampling and analysis were 

not standardized throughout the DOE weapons complex until the early 1970s (RAC Report (CDC 

2001, Chapter 12.2, p. 12.2-1)). As of 1973, a laboratory had been dedicated to soils analysis and the 

necessary field and laboratory equipment had been purchased or fabricated. 

Annual Intakes from Resuspension

Soil analyses conducted in the 1970s and 1980s indicated no measurable SRS-released plutonium at 

off-site locations. However, elevated levels of certain radionuclides, including plutonium, were 

detected at several locations near the release points in both F- and H-Areas. Careful review of these 

data over several years concluded that the measured soil concentrations in F and H Areas were due 

almost entirely to deposition during the years of the greatest releases. Since the great majority of 

239Pu was released from F Area in a single year, 1955, and the majority of 238Pu was released from H 

Area in 1969, the measured soil concentrations for 239Pu could be conservatively assumed to have 

persisted in F Area since 1955 and for 238Pu in H Area since 1969 (RAC Report (CDC 2001, p. 12.2-

7)). Because the atmospheric releases of nonvolatile radionuclides from other on-site locations were 

much less than those from F and H Areas, only the soil measurements in F and H Area were deemed 

important for purposes of on-site dose reconstruction. 

For purposes of reconstructing potential missed or unmonitored dose, the annual intakes provided in 

Table C-18 should be used only for individuals whose employment records indicate that they spent 

considerable time in either F or H Areas. The 239Pu annual intakes should be assumed to have 

persisted since 1955, and the 238Pu annual intakes since 1969. Annual intakes resulting from 

concentrations of 90Sr and 137Cs should be conservatively assumed to have persisted essentially 

unchanged since 1955. For workers assigned to both areas, the area that results in the larger organ 

dose will be used to represent the entire period of exposure.

Section 1.3 above discusses available information regarding chemical forms of radioactive materials 

in use at SRS for major facilities and processes. These chemical forms, and the SRS assumptions 

regarding ICRP 30 inhalation solubility classes, provide information to assist the internal dose 

reconstructor in determining appropriate solubility parameters for use with the IMBA program. 

Additionally, Attachment A contains a number of tables providing information related to radionuclide 

mixtures and dosimetric quantities assumed to be applicable for those radionuclides as taken from 

SRS program documentation. It is important to note that these dosimetric parameters were based on 

ICRP 30 recommendations. Internal doses reconstructed in support of EEOICPA are derived using 

more recent ICRP 68 (1994) recommended methodology than was in use at SRS. Consequently, 

inhalation solubility classes, dose conversion factors, and annual dose fractions discussed in Section 

1.3 and Attachment A of this document should guide the selection of similar quantities used for 

EEOICPA internal dose reconstructions.

Annual intakes resulting from resuspension of radionuclides in soil in concentrations given in Table 

3.1.2-1 are presented in Table C-18 and were calculated using the following formula:

CA = SD × CS × SF × RF (3-3)

where

CA = Air concentration, pCi m-3
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SD = Soil density = 1600 kg m-3 (PNL-10550, 1995) 

CS = Radionuclide soil concentration = pCi kg1

SF = Surface factor = 0.08 m (Arnett 1993)

RF = Resuspension factor = 1.00E-09 m-1 (Till and Meyers, 1983, p. 5-32)

Table 3.1.2-1. Radionuclide soil concentrations, pCi g-1, (1 sigma) dry weight (0-8 cm 

depth).

a

Location Sr-90 σ Cs-137 σ Pu-238 σ Pu-239 σ

F-Area 0.0216 0.0182 1.05 0.05 0.403 0.014 0.531 0.014

H-Area 0.0289 0.0183 1.32 0.05 0.0213 0.0016 0.0554 0.0023

a. Source: CDC 2001, Table 12.2-8, on p. 12.2-8, Arnett 1993.

It has been assumed that workers may have ingested foodstuffs (fruits and vegetables) that have 

been grown in radioactively contaminated soil on the SRS. To account for the potential doses 

received by workers from this source the doses have been bounded by the use of claimant favorable 

assumptions related to soil concentrations, soil-to-vegetation transfer factors and, in the case of 

tritium, maximum vegetation concentrations. The 50 year equivalent organ dose (mrem) has been 

calculated based on the consumption of 60 kgm (wet mass) of foodstuffs consumed per year. These

values are presented in Attachment 3.C. The doses were calculated for 90Sr, 137Cs, 239Pu, 238Pu, and 

HTO since these radionuclides would provide 95% of the dose potential.

3.1.3 Annual Doses from the Ingestion of Foodstuffs

3.2 OCCUPATIONAL EXTERNAL DOSE

Workers are subjected to external doses from the ambient radiation levels and from noble gases. 

Prior to 1966, ambient radiation levels were measured with pocket ionization chambers and GM 

detectors. After 1966, the measurements were made with ion chambers and thermoluminescent 

dosimeters (TLDs). Concentrations of the noble gas 41Ar were calculated and can be used to 

determine the external dose due to submersion.

3.2.1

The environmental radiological profile has been developed for the Savannah River Site for use by 

dose reconstructors when personal dosimetry or bioassay program participation was not required. 

Site annual environmental reports were reviewed for data that would be useful in reconstructing 

ambient radiation levels and airborne radionuclide concentrations. Data in these historical documents 

(see reference section for citations) included ambient ion chamber and thermoluminescent dosimetry 

(TLD) radiation measurements, air sampling results, and residual soil activity. The historical ion 

chamber and TLD results with uncertainty estimates (geometric standard deviations), where available, 

were tabulated and are presented in Table C-19.

Ambient Radiation

The ambient radiation levels presented in Table C-19 were taken directly from the SRS annual and 

semiannual environmental reports (see reference section) and include contributions from the SRS 

releases and background radiation. Until 1966, ambient radiation measurements were performed 

using either pocket ionization chambers or GM detectors, and results were reported in units of millirep 

(mrep), milliroentgen (mR), or millirad (mrad) per 24 hours without any estimates of uncertainty. In 

1966, the Savannah River Site began performing ambient radiation measurements with TLDs. For 

purposes of this report, mrep, mR, and mrad were assumed to be equivalent to mrem. Although 

ambient radiation results were typically reported in 24 hour increments, measurements were usually 
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performed over one-week periods resulting in between 600 and 800 measurements per year from all 

locations. The values in Table C-19 are for 2,000 hours of exposure per year. 

In 1974, the environmental reports began presenting a quantitative estimate of the uncertainty 

associated with the maximum and minimum radiation measurements. Along with the maximum and 

minimum radiation values, the reports also provided a value for two sigma. The 50th-percentile 

values presented in the tables are generally the average reported values for a given area. The 

geometric standard deviation values in Table C-19 were based on the reported two sigma value. It 

should also be noted that all of these values include the contributions from natural background 

radiation and nuclear weapons fallout. Therefore, use of these data to account for potential missed or 

unmonitored dose should in all cases represent an overestimation of the worker’s actual dose.

3.2.2

In addition to radionuclides important to internal dose reconstruction, 41Ar was released in relatively 

large quantities during reactor operation. The radionuclide 41Ar is a noble (nonreactive) gas with a 

half-life of 1.8 hours and is produced in the air surrounding the reactor. Because this air volume 

communicated with air in the reactor building, most of the 41Ar produced in the reactor was released 

with ventilation air exiting the 200-foot reactor stack. The five reactors were responsible for the vast 

majority of the atmospheric releases of 41Ar. Neutrons interacting with the naturally occurring 40Ar in 

the air surrounding the reactor vessel created 41Ar which quickly mixed with reactor building air and 

was exhausted through the main stack. Once out of the stack, reactor gases moved and were 

dispersed with local air flows and both on-site and off-site individuals nearby were potentially exposed 

to beta and gamma radiation emitted during the 41Ar decay. 

Releases of Noble Gases

The most reliable, reactor-specific, source term estimates, according to CDC (2001, p, 4.3-1), were 

provided by Cummins (1991). These source terms were used to calculate the 50th-percentile annual 

whole body and organ doses presented in Tables C-20 and C-21. Because Cummins (2001) did not 

include estimates of uncertainty, the 95th-percentile source terms were estimated by assuming that 

the 95th-percentile release was simply 25 percent greater than the 50th-percentile release.

This potential dose resulting from submersion in the plume of this noble gas is considered to be 

entirely external because the gas is not readily assimilated by the body. Using methodology identical 

to that for the radionuclides important to internal dose, calculated air concentrations at the populated 

areas of the SRS were estimated. These concentrations were then converted to organ and whole 

body dose using submersion dose conversion factors published in Federal Guidance Report No. 12 

(EPA 1993). The annual doses (assuming 2000 hours per year of exposure) have been tabulated 

based on 50th-percentile source terms presented in Tables C-20 and C-21. The geometric standard 

deviations provided in these tables were based on the estimated 95th-percentile source terms using 

the relationship provided in Section 3.1.1 above.

For purposes of external dose reconstruction, 41Ar doses to the skin, bone surfaces, whole body, and 

other underlying organs (Tables C-20 and C-21) should be assessed and should be assumed to be 

100% from >250 keV photons.

3.3 UNCERTAINTY

As discussed in the previous sections, estimates of annual intakes employed conservative (i.e., 

claimant favorable) assumptions. For example, the use of maximum, 25-year, sector X/Q values to 

estimate atmospheric radionuclide concentrations. Rollins (2003) validated that these calculated air 

concentrations would likely overestimate the actual radionuclide concentrations by comparison to 
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actual measured data taken from the SRS annual environmental reports. It should be noted that the 

estimated annual intakes were based on annual average meteorology and that the actual 

instantaneous meteorology on any given day could vary from these averages by factors (plus or 

minus) of several orders of magnitude. However, the intake values given in Attachment C should 

represent a reasonable upper bound of the actual intakes that could have occurred.

In instances where more detailed information is known about a particular individual or job 

classification, other modifying factors should be employed. For example, if, for a particular job 

classification, there is reason to believe that the actual ventilation rate for the claimant may vary 

markedly from the average of 2,400 cubic meters per 2,000 hours of exposure, the dose reconstructor 

should use professional judgment to adjust the estimated intakes as necessary whether the individual 

is engaged in light or heavy work. The respiration rate for light work is 1.2 m3

/h and 1.7 m3/h for heavy 

work. In these cases, to estimate annual intake, the product of the fractional annual period for each, 

job dependent, level of work and the corresponding ventilation rate, are summed to determine the 

total ventilation volume for the year m3

. The annual intake is then determined by multiplying the 

annual ventilation volume by the annual average concentration in Tables C-1 through C-17 for the 

year and location of interest.

In general, the annual intake values presented in tables C-1 through C-17 reflect adjusted source 

terms and uncertainty analyses provided in CDC (2001). For purposes of this document, no other 

attempt has been made to quantify other uncertainties. 

3.4 DOSE RECONSTRUCTOR INSTRUCTIONS

The information given below summarizes instructions given to the dose reconstructors in determining 

doses from environmental (on-site ambient) radiation. For the purpose of evaluating probability of 

causation, on-site ambient doses are always considered chronic, and are considered to be photons 

with energy E=30-250 keV.

3.4.1

Organ doses from on-site ambient radiation have been adapted from the information in Table C-19. 

For convenience and to ensure claimant favorability, the maximum dose listed for any area on the site 

has been selected to represent the assigned dose for that year, regardless of the area(s) the 

employee actually worked. For the year 1953, even though the site was not in operation for the full 

year, the value for 1954 has been applied as a claimant favorable assumption. Also, the value for 

2001 has been applied to 2002 and 2003 to ensure completeness.

Assignment of Environmental (On-Site Ambient) Doses: Maximizing Approach for 

Dose Reconstructions

The doses listed in Table C-19 were multiplied by a factor of 1.25 to account for the possibility of 50 

hours of work per week rather than 40 hours per week as assumed in the development of the table. 

The resulting maximum annual environmental (on-site ambient) doses that may be applied in dose 

reconstructions for likely non-compensable cases are shown in Table 3.4-1.

The values shown are applied for each full or partial year of employment. Because the values are 

based on an assumed 50-hours per week, if the dose reconstructor determines that a higher number 

of hours is appropriate, the annual doses can be raised accordingly.

For simplicity, the organ dose conversion factor is assumed to be 1 for all organs. The doses are 

treated as a point estimate (constant) for the purpose of calculating probability of causation.
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The combination of assumptions described above ensures that on-site ambient doses assigned are 

claimant favorable. 

3.4.2

A lognormal distribution is applied to the above values with a GSD of 1.3. The relevant organ dose 

conversion factor is applied based on an isotropic exposure geometry and a photon energy of 30-250 

keV.

Assignment of Environmental (On-Site Ambient) Doses: Best Estimate Approach 

for Dose Reconstructions
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Table 3.4-1. On-site ambient organ doses (rem).

Year Dose

1951 0.000

1952 0.000

1953 0.096

1954 0.096

1955 0.105

1956 0.113

1957 0.134

1958 0.150

1959 0.109

1960 0.228

1961 0.071

1962 0.080

1963 0.101

1964 0.094

1965 0.078

1966 0.131

1967 0.111

1968 0.059

Year Dose

1969 0.084

1970 0.101

1971 0.084

1972 0.095

1973 0.099

1974 0.089

1975 0.085

1976 0.068

1977 0.090

1978 0.061

1979 0.065

1980 0.044

1981 0.074

1982 0.065

1983 0.069

1984 0.093

1985 0.114

1986 0.059

Year Dose

1987 0.053

1988 0.050

1989 0.050

1990 0.046

1991 0.028

1992 0.021

1993 0.031

1994 0.031

1995 0.034

1996 0.031

1997 0.033

1998 0.038

1999 0.035

2000 0.036

2001 0.034

2002 0.034

2003 0.034

The dose reconstructor has the option of using the Table C-19 doses for the actual area(s) in which 

the employee worked rather than the maximum values shown above. The dose reconstructor also 

has the option of reducing the yearly doses to account for partial years of employment. The dose 

reconstructor also has the option of not assigning environmental dose if it can be determined that an 

elevated background dose was not subtracted from the dosimeter readings for reporting purposes.
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4.0

Radionuclides of concern listed in the SRS Internal Dosimetry Technical Basis Manual (WSRC-IM-90-

139, 31 December 2001), chapter 5, include the following elements and isotopes:

OCCUPATIONAL INTERNAL DOSE

• tritium (3H)

• manganese (54Mn)

• cobalt (60Co)

• zinc (65Zn)

• strontium (90Sr)

• ruthenium (106Ru)

• antimony (125Sb)

• cesium (134Cs, 137Cs)

• cerium (144Ce)

• europium (152Eu, 154Eu)

• thorium (228Th, 232Th)

• uranium (234U, 235U, 236U, 238U, and mixtures)

• neptunium (237Np)

• plutonium (238Pu, 239Pu/240Pu, 241Pu)

• americium (241Am, 243Am)

• curium (244Cm)

• californium (252Cf, including 248Cm daughter)

These radionuclides were those identified by SRS as the most significant at delivering doses or that 

were effective tracers for significant dose-delivering radionuclides. A summary of the data in this 

section is presented in Attachment D.

4.1 IN VITRO MINIMUM DETECTABLE ACTIVITIES (MDAS), COUNTING METHODS

From WSRC-IM-90-139, the decision level for urinalysis is calculated as:

TREVA

C

D

BACKGROUND

L

* 3

2

4.65

+

= (4–1)

where

DL is in dpm/L

CBACKGROUND = background counts in the region of interest

T = count time

R = recovery

E = average detector efficiency

V = sample volume (usually 1.5 L), and

A = the alpha abundance for the radionuclide in question.

The decision level is the level at which activity is considered present in a sample with a 95% 

confidence level.
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Similarly, the MDA is calculated as:

TREVA

C

MDA

4.65 * BACKGROUND + 3 = (4–2)

The reporting level is distinguished from the MDA. As described in WSRC-IM-90-139, the reporting 

level is the minimally acceptable decision level; the bioassay laboratory was required to achieve 

decision levels below the reporting level. SRS technical documentation provided no quantitative 

relationship between the reporting level and MDA.

4.1.1

Since 1990 excreta analysis records have been maintained in an electronic database. The database 

identifies baseline (new hire), routine, special, and termination measurements. Prior to the electronic 

database, excreta results were kept on employee bioassay cards

In Vitro Urine Analysis

MDA values since about 1988 are 2 times the average decision level for the analysis. These MDAs 

apply to both routine urinalyses and to special urinalyses. Since about 1990, routine analysis results 

(except tritium) that exceed the decision level are recounted for twice the normal count time. The 

recounted result is placed in the database instead of the original count, so the MDA for a recount 

result will be approximately 0.7 times the values shown for routine analyses. Some samples are 

“rerun,” meaning a second aliquot is processed. In these cases the database will show two results 

with different sample numbers for the same collection date and same sample volume. 

Activity fractions for depleted, recycled, and high enriched uranium are listed in the glossary section.

4.1.2

The following section discusses the in vitro methods for specific radionuclides.

In Vitro Methods for Individual Radionuclides

In vitro bioassay for trivalent actinides (americium, curium, and californium)

Records showing urinalysis for trivalent actinides date back at least to the mid 1960s. The early 

reporting levels varied from 1 to 3 dpm/1.5 L. In 1971 the reporting level using gross alpha counting 

on a solid state detector was listed as 0.3 dpm per 1.5 liters. In 1990, a change in radiochemical 

processing resulted in an MDA of 0.1 dpm per liter. Alpha spectrometry has been used since 1994 

with MDAs of 0.064 dpm/L for 241Am and 237Np and 0.047 dpm/L for 244Cm and 252Cf. 

Use of a gross alpha count prior to 1994 means that the measurement itself does not indicate which 

radionuclide or combination of radionuclides is being measured. Location or other information 

concerning an intake may allow determination of the principal radionuclide. If not, it is claimantfavorable to assume 241Am.

In vitro bioassay for plutonium

From the beginning of the plutonium urinalysis program in 1954 to approximately 1959, urine samples 

were radiochemically processed using bismuth phosphate and lanthanum fluoride coprecipitation and 

electroplated, and activities were determined by gross alpha track analysis of exposed NTA 

emulsions. In 1959 nitric acid/hydrogen peroxide dissolution and ion exchange replaced the bismuth 

phosphate method. This was faster and used less volume of urine but had essentially the same MDA. 

The reporting level did not change. Results were recorded as Pu or sometimes as Pu238/Pu239. 

From around 1964 to 1988, counting for gross alpha activity was performed using a solid state surface 

barrier alpha detector. Tri-isooctlyamine (TIOA) liquid extraction replaced the ion exchange chemistry 
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in 1966. This method also used direct evaporation on planchets instead of electrodeposition. This 

method also allowed separation of neptunium and uranium from the same sample. Sensitivity was 

stated at 0.1 dpm/1.5L, which was consistent with the reporting level already in use. In or about 1981, 

a new coprecipitation technique was introduced for routine samples along with alpha spectrometry. 

Sample-specific determination of plutonium recover by use of a 242Pu tracer was also introduced at 

that time. Results for 238Pu and 239Pu were reported separately. The TIOA method with gross alpha 

counting continued to be used on special samples until 1988. An electronic database was introduced 

in 1990 and results were thereafter reported as per liter. Electrodeposition was reinstated in 1994. 

Separation of plutonium+neptunium, actinides, uranium, and strontium from a single sample using 

TEVA and TRU resins began in 2001. Alpha-emitting plutonium and neptunium isotopes are 

electrodeposited and counted by alpha spectrometry on a single planchet. Reporting levels for the 

various radiochemical and counting methods are listed in Tables 4.1.1-1 and 4.1.1-2. Note: MDAs 

are assumed to be more or as sensitive as the reporting levels. 

Table 4.1.1-1. Pu-239/240 – in vitro urinalysis.

Period Reporting levela Method

1954–1959 0.05 dpm /1.5L Bismuth phosphate, lanthanum fluoride 

coprecipitation; gross alpha for plutonium; 

alpha track counting

1959–1962 0.05 dpm/1.5L Ion exchange, gross alpha for plutonium, 

alpha track counting (until 1964)

1963–1966 0.1 dpm/1.5L (1963) Solid state alpha counting (1964)

1966–1981 0.1 dpm/1.5L TIOA extraction; gross alpha for plutonium on 

solid state detector 

1981–1988 0.07 dpm/1.5L for routine samples Alpha spectroscopy for 239/240Pu (used 

primarily for routine urine samples) 

1988–present 0.029 pCi/L (0.06 dpm/L) Alpha spectroscopy for 239/240Pu

Table 4.1.1-2. Pu-238 – in vitro urinalysis.

Period MDA/reporting level – urinalysis Method

1954–1981 Reporting levels same as in Table 4.1.1-2, 

238Pu and 239Pu combined into single result

Methods same as in Table 4.1.1-2

1981–1988 0.05 dpm/1.5 L [reporting level] Alpha spectroscopy for 238Pu (used primarily 

for routine urine samples)

1988–present 0.030 pCi/L (~0.07 dpm/L) [MDA] Alpha spectroscopy for 238Pu

Because plutonium urinalysis results (fecal results as well) prior to 1981 (1988 for special samples) 

measured the total activity from alpha-emitting isotopes whereas 238Pu and 239+240Pu results were 

reported separately from 1981 to the present, the two sets of data cannot be entered into the intake 

code as reported. If, for a specific case, all the bioassay data are prior to 1981, the total plutonium 

alpha data can be entered as 239Pu. The calculated intake will have to be partitioned into 238Pu and 

239Pu per the mixture ratios applicable to the intake (if known). However, it is unlikely that for the older 

intakes, intake-specific isotopic mixture ratios were established so the default mixture ratios will have 

to be used. The same mixture ratio will then be used to determine the intake of 241Pu and 241Am. 

(Note: in the chemistry techniques described above, the gross alpha count would not have included 

any 241Am present in the bioassay sample.)

Plutonium existed in mixtures of 238Pu, 239Pu, 240Pu, and 241Pu. The activity of 242Pu was always 

insignificant dosimetrically. The relative activities of these isotopes depended on the nature of the 

irradiation of the uranium fuel (referred to a burn-up) and the time between the end of irradiation and 

the intake. Processes at SRS did not perturb the relative activities of the isotopes. Generally, SRS 

created plutonium mixtures ranging from about 3% per weight 240Pu to about 12% by weight 240Pu. 
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Generally, the plutonium mixtures were blended to produce a final product with about 6% by weight 

240Pu, which is referred to as a weapons-grade mixture. The 241Pu, a beta-emitter with a 14-yr halflife, undergoes noticeable decay to 241Am during the lifetime of the facilities so that the isotopic ratios 

of contamination will be different than the ratios in newly irradiated fuel (or new feed to the Head End 

Stream). 241Am builds up from near zero at end of irradiation; however, it is removed during 

production of the plutonium product, then begins to build up again as the 241Pu remaining in the 

product decays. So the ratio of 241Pu to, say, 239+240Pu decreases from time of end of irradiation; 

whereas the ratio of 241Am to 239+240

Pu increases from time of last separation of the 241Am from the 

plutonium. 

If for a given intake the activity of 241Am was established, through chest counting, trivalent actinide 

excreta analysis, or analysis of the source of the contamination, then that activity should be used. If 

unknown, the default ratios provided in Table 4.1.1-3 can be used as best fits the nature of the intake. 

Ten-year old 12% plutonium mix is the most claimant-favorable. 

Table 4.1.1-3. Activity composition for reference 6% and 12% 240Pu mixtures.

a

Isotopic component

Reference 6% Pu mix Reference 12% Pu mix

Freshb Agedc 5 years Aged 10 years Fresh Aged 5 years Aged 10 years

Specific activity in mixture, Ci/g

Pu-238 8.56E-3 8.23E-3 7.91E-3 1.71E-2 1.64E-2 1.58E-2

Pu-239+240 7.13E-2 7.13E-2 7.13E-2 7.98E-2 7.98E-2 7.98E-2

Pu-241 8.24E-1 6.48E-1 5.09E-1 3.09 2.43 1.91

Am-241 0 5.83E-3 1.04E-2 0 2.19E-2 3.89E-2

Pu alpha 7.99E-2 7.95E-2 7.92E-2 9.69E-2 9.62E-2 9.56E-2

Total alpha 7.99E-2 8.53E-2 8.96E-2 9.69E-2 1.18E-1 1.35E-1

Activity ratios

Pu239+240:Am-241 NA 1.22E1 6.87 NA 3.64 2.05

Pu-alpha:Am-241 NA 1.36E1 7.61 NA 4.39 2.46

Pu-241:Pu-239+240 1.16E1 9.09 7.14 3.87E1 3.05E1 2.39E1

Pu-241:Pu-alpha 1.03E1 8.15 6.43 3.18E1 2.53E1 2.00E1

Pu-238:Pu 239+240 0.120 0.115 0.111 0.214 0.206 0.198

Pu-238:Pu-alpha 0.107 0.104 0.999 0.176 0.170 0.165

a. From Methods and Models of the Hanford Internal Dosimetry Program (Carbaugh, 2003). Also found in part in Guide for Good 

Practices for Occupational Radiological Protection in Plutonium Facilities (USDOE 1998).

b. Two weeks after end of irradiation.

c. Time since separation of the 241Am.

If some of the bioassay data were obtained prior to 1981 and some after, then the data will have to be 

made consistent. This can be done in either of two ways. 1) If all the 238Pu and 239Pu results are 

above their respective decision levels, then the 238Pu and 239Pu results for each sample can be 

summed to get a total plutonium alpha result. This result can be equated to the pre-1981 results and 

treated in the manner described in the preceding paragraph. 2) If some of the 238Pu or 239Pu results 

are below their decision levels, then it may be best to use the isotope that has the most robust data 

set and adjust each of the pre-1981 results to reflect just the portion from that chosen isotope. The 

adjustment will be made using the default isotopic mixture ratios unless the intake-specific ratio is 

known. For instance a gross alpha result of 2.1 dpm/1.5L would be converted to 1.9 dpm/1.5L of 

239Pu for 6% Pu or 1.8 dpm/1.5L for 12% Pu before being entered into the intake code.

In vitro bioassay for tritium

From startup until 1958, HTO in urine was analyzed by passing hydrogen evolved from the urine 

sample through an ionization chamber; the reported MDA for this method was 1 μCi/L. In 1958, liquid 

scintillation counting was initiated, which is still the method in use. A History of Personnel Radiation 

Dosimetry at the Savannah River Site (WSRC-RP-95-234) reports that the MDA consistently 
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improved to the current level of 20,000 pCi/L (or 0.02 μCi/L). The reporting level remained at the 

value of 1 μCi L1, and was later reduced to 0.5 μCi L-1

, on or about 1981, and then to the current 

value of 0.1 μCi L1

. (an exact date was not determined at least by 1989). During the 1980s, although 

the reporting level of 0.5 μCi L-1 was generally used, some results below 0.5 are listed directly, e.g. 

0.4, 0.3. Considering that the true MDA was probably well below the reporting level, these results 

should be considered as real. 

It should be noted that for current analyses, tritium results of 0.05 μCi L-1 or less are reported as 

“<0.1μCi L-1

,” and results between 0.05 μCi L-1 and 0.1 μCi L-1 are reported as “0.1 μCi L-1.” Results 

greater than 0.1 μCi L-1 are reported as measured (to 1 significant figure).

Tritium analyses are listed as “T” on the employee bioassay cards. Tritium may also be listed as “P10,” especially in the 1950s. Tritium results in the 1990s were listed on the same summary form as 

external dose monitoring results. They are referred to as sample results with dates and analysis 

results, but the word “tritium” or any other radionuclide identifier is not mentioned per se. 

For tritium results, the denominator used for reporting purposes has always been per liter of urine. 

(The denominator of 1.5 liters was never used for tritium as it was for other radionuclides.)

Only tritium bioassay results greater than 5 μCi L-1 were evaluated prior to computer evaluation of 

data.

In vitro bioassay for uranium

A variety of methods have been used historically to analyze uranium at SRS. These methods and the 

associated detection capabilities are summarized in Table 4.1.1-4.

Table 4.1.1-4. Uranium urinalysis.

Time period Uranium mixture Reporting level Urine analysis method

Startup to mid-1960s Enriched uranium 0.15 dpm/1.5 L Gross alpha for uranium, alpha track counting

Mid-1960s to 1982 Enriched uranium 1 dpm/1.5 L Gross alpha for uranium on solid state detector

1954–1982 Depleted uranium 1-5 μg/La Fluorophotometric analysis

1982–1986 U-235 0.14 ng Delayed neutron analysis (DNA)

U-nat 1 μg/L DNA analysis for U-235

Enriched uranium 1 dpm/L DNA analysis for U-235

1986–1990 Enriched uranium 1 dpm/1.5 L Gross alpha for uranium on solid state detector

1986–1994 Depleted uranium Not found, use 1 μ/L Kinetic phosphorimetry analysis (KPA)

1990–present (enriched 

uranium), 1994–

present (depleted 

uranium)

DU Use U-238 and isotopic 

ratios from glossary

Alpha spectroscopy for specific uranium 

isotopes

RU Use U-238 and isotopic 

ratios from glossary

HEU Use U-235 and isotopic 

ratios from glossary 

U-234 0.032 pCi/L

U-235 0.036 pCi/L

U-238 0.032 pCi/L

a. Use value as shown in the individual’s records if available or otherwise use 5 μg/L.

Activity fractions for depleted, recycled, and highly enriched uranium are listed in the glossary section. 

SRS technical documentation indicates that for earlier monitoring periods, the designations “enriched” 

and “depleted” analysis for uranium referred to analysis performed by alpha counting or chemical 

measurement, respectively, and was not necessarily indicative of the degree of uranium enrichment. 

EU was the code used on employee bioassay cards for the gross alpha count method, and DU was 

used to designate the fluorophotometric method. 
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Because of excretion of uranium from natural sources, the reporting levels for natural or depleted 

uranium listed in Table 4.1.1-4 should not be used to determine if a urinalysis result greater than the 

reporting level resulted from occupational intakes. Based on a probit analysis of 1728 worker 

samples in 1993-1994 (WSRC-IM-90-139, rev. 8), an upper limit at the 99% confidence level for 

analytical noise and presumably some effect of natural background was set at 0.15 dpm/L. The SRS 

policy is that a sample result must exceed both its decision level and 0.15 dpm/L to be considered a 

detection. Lacking other information, i.e., similar studies at different times throughout the history of 

the SRS, it is reasonable to apply the 0.15 dpm/L criterion to all historical samples. The background 

excretion in terms of μg/L was not established, but almost certainly was below the 1-μg/L reporting 

level. Therefore, for the mass uranium analyses, results greater than the reporting level in Table 

4.1.1-4 can be associated with an occupational intake. For missed dose calculations, the reporting 

levels in Table 4.1.1-4 can be used.

If isotopic data are available, they can also be used to determine if a uranium urinalysis result 

represents an occupational intake. The following rules apply:

• If 235U or 236U are detected, then 234U must also be detected.

• If both 234U and 238U are detected, the uranium should be considered from occupational 

sources if the 234U/238U ratio is outside of a range from 1 to 3. 

• If 238U is detected, the uranium should be considered from an occupational source if the ratio 

of the 234U critical level to the 238U result is less than 1.

• If 234U is detected, the uranium should be considered from an occupational source if the ratio 

of the 234U result to the 238U critical level is greater than 3. 

In vitro analysis for radiostrontium

The start date for urinalysis for radioisotopes of strontium was not specifically determined, other than 

“late 1950s.” From start until 1969 strontium was separated by alkaline earth phosphate 

coprecipitation followed by beta counting on a GM counter. This analysis was also called fission 

product analysis (see discussion below). Both 89Sr and 90Sr would have been counted (and 

potentially radioisotopes of cerium and promethium), but it is claimant-favorable to assume the result 

is all 90Sr. Yttrium-91, grown in from 91Sr, might also be present at intake if the source of intake was 

recent fission products (up to about a year post irradiation). If the intake prior to 1969 occurred in 

reactor buildings or the 221-F canyon facility and an intake of 90Sr is indicated, then an equal intake of 

91Y should be assumed. This is a claimant-favorable assumption because after about a year post 

irradiation, 91Y activity would be negligible. The reporting level is assumed to be the same as for the 

beta component of the fission product analyses. 

From 1969 to 2001 strontium was analyzed by liquid ion-exchange that separates the yttrium progeny, 

followed by beta proportional counting. Yttrium-91 would be included as a possible interference but 

89Sr would not. For 90Sr intakes in 1969 to 1989 in reactor buildings, an intake of 89Sr equal to the 

intake of 90Sr should be assumed. 

Since 2001, 90Sr has been separated as part of the use of TRU and TEVA resins for plutonium, 

uranium, actinides, and strontium. The present MDA is 2.9 pCi/L.

In vitro fission product analysis

Prior to whole body counts, intakes of gamma-emitting fission products were monitored by urinalysis. 

The record is not clear exactly what procedure was used; both gamma-spectrometry analyses and 
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gross beta counting are mentioned as being possible. Nor has it been determined if any separation 

chemistry was done prior to counting, though the reporting levels that were used seem to imply that 

40K was removed or reduced prior to counting. These analyses were coded FP, FPIA, or IA under 

element on the employee bioassay cards. The IA code can have two meanings depending on the 

context. An IA code with a result prior to 1960 most likely refers to the fission product analysis; an IA 

code with no result means inconclusive analysis, that is, the lab was unable to produce a valid result. 

The employee bioassay cards show less-than values that varied over the years, presumably 

indicating different reporting levels and probably indicating a switch from beta counting to photon 

counting; for instance, 1955-1960: <30 d/m/750 ml; 1960-65: <100 d/m/1.5 L; 1966-1971; <1 

nCi/1.5L. Some records in the 1960-65 period show two results for the same sample: one for beta 

(<100 dpm/1.5L) and one for gamma (<500 dpm/1.5L). This would imply they did a gamma scan after 

removal of 40K, at least. The records also show some fission product urinalyses as late as 1984 with 

the result listed as <1. Presumably the units are nCi/1.5L. 

A procedure for fission product urinalysis at Hanford, dated May 1950, has been discovered 

(Thorburn, 1950). It is likely that the Savannah River procedure was similar; at least through 1960. 

The procedure separated and counted radionuclides of strontium, yttrium, cerium, and promethium. It 

did not account for radionuclides of ruthenium or cesium. 

To interpret results from the fission product urinalysis, it is noted that the strontium, barium, europium, 

zirconium, and niobium radionuclides concentrate primarily in the bone, with 90Sr providing the largest 

dose rate. Cerium, lanthanum, and promethium concentrate primarily in the liver, with some 

concentration in the bone, with 144Ce providing the largest dose rate. Cesium and ruthenium are 

assumed to be uniformly distributed in the whole body. 

Hence, based on what little is known, fission product urinalysis results should be interpreted as 

follows:

Note: If the claimant had a whole body count during any year and had a detectable fission product 

urinalysis result, the intakes of 137Cs, 106Ru, or any other gamma-emitting fission products should be 

determined from the whole body count. The fission product result could be assumed to represent 90Sr 

in total, or if more accuracy is desired, the activity of the gamma-emitting fission products could be 

subtracted from the fission product result with the remainder being 90Sr. This requires determining the 

activity of the radionuclides found in the whole body count that would be excreted in the urine sample.

• 100% Type F 90Sr if the claimant’s disease is related to bone or bone marrow plus an equal 

intake of 137Cs (Type F) and 3 fold intake of 106Ru (Type F). 

• 100% Type M 144Ce if the claimant’s disease is related to the liver plus an equal intake of 106Ru 

(Type F) and a 20% intake of 137Cs (Type F). 

• 100% Type M 144Ce if the claimant’s disease is related to the respiratory tract or digestive 

tract, plus an equal intake of Type S 106Ru and a 20% intake of 137 Cs (Type F). 

• For other disease locations, use the conditions in the first bullet. 

Fission product urinalysis was not available in 1954 when the reactors were first operating. Because 

tritium was the dominant source of intake at the reactors, if there was evidence that a worker had a 

tritium intake in 1954, then assume an intake of one or more of the fission products listed above 

based on an MDA bioassay from a one-year-old intake.
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In vitro analysis for 237Np

The Savannah River Technical Basis Manual, Rev. 8, says that neptunium analyses started in the 

1950s, perhaps 1959, using coprecipitation with calcium-magnesium ammonium phosphate and ion 

exchange, followed by autoradiography on NTA emulsion for 10,000 minutes. The MDA was not 

given but was mostly likely similar to the MDA for plutonium alpha autoradiography method used at 

the same time, i.e., about 0.035 dpm/L. The results were recorded as S.P. (special product), and the 

reporting level was 0.05 dpm/1.5 L, same as plutonium. Analysis by TIOA extraction and 

electrodeposition occurred at the same time as for plutonium analyses, about 1964, which was a 

faster method but the reporting level stayed the same. In 1993-1996, samples were analyzed by an 

offsite, commercial laboratory using extraction chromatography resin. In 1996 analysis returned to 

SRS and alpha spectrometry was introduced. Since 2001, TEVA and TRU resins are used to 

separate neptunium from actinides and uranium, and alpha spectrometry has continued. The MDA is 

still listed as 0.035 dpm/L; however, from personal communication with the SRS lead internal 

dosimetrist, an MDA of 0.4 dpm/L was considered more applicable in practice. If results are reported 

below 0.4 dpm/L, they can be used as given. 

4.1.3

At SRS, routine urine samples are considered partial-day samples. Urinalysis results are recorded as 

per 1.5L or per L. The code used to determine intake from bioassay data requires the period of time 

represented by the sample and the result of the sample in activity per day. If the volume of the 

sample is known, the result can be adjusted to activity per sample. Then if the collection time of the 

sample is known, the activity can be adjusted to activity per day by ratioing the collection hours to 24 

hours. (This is often the case for special samples collected since about 1990.) If either the collection 

period or the volume is not known, then a result in activity per L should be normalized to activity per 

day using the reference volume of 1.4 L per day. Activity expressed per 1.5L can be assumed to be a 

day’s excretion. 

Correcting for Urinalysis Volume

4.1.4

SRS technical documentation indicates that fecal sample analysis has been performed periodically 

since the 1950s. These samples were always for special, rather than routine, sampling purposes, and 

were always performed in conjunction with urine bioassay sampling. Fecal results will be found on 

separate reports, not routine employee bioassay cards.

Fecal Sample Analysis

Fecal samples are screened by photon counting with the following typical MDAs: 7 pCi (241Am), 300 

pCi (238Pu), and 600 pCi (weapons grade plutonium). However, if results could not be quantified by 

photon counting, that is, if nothing was detected, then samples may have been analyzed by normal 

separation chemistry and alpha spectrometry to improve the detection capability. This protocol 

applies mostly to samples from 1994 to present, and probably was not applied to samples prior to 

1994. Because only an aliquot of the fecal sample is used for wet radiochemistry, the MDAs are 

typically 1.5 times the MDAs for routine urine samples. 

4.2 IN VIVO MDAS, COUNTING METHODS, AND REPORTING PRACTICES

Whole body counting has been performed since 1960 when the SRS Whole Body Counting facility 

was completed. Chest counting was initiated in the early 1970s using phoswich detectors. In vivo 

count records are not kept in a central electronic database; each count is maintained as a single hardcopy record. A whole body count record and a chest count record on the same day are listed on a 

single hard-copy record. In vivo bioassay methods and detection capabilities are reviewed below.
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4.2.1

Prior to whole body counting, fission products intakes were monitored using urinalysis.

Whole Body Counting

Whole body counting began in approximately 1960, using a 4” high by 8” thick diameter NaI detector. 

(Detection energy range was 100 keV to 2000 keV.) The monitored individual sat in a reclining chair 

positioned in an arc around the detector; this was referred to as the “40-cm arc geometry” in bioassay 

monitoring reports. Bioassay via this method was not used for plutonium and americium due to their 

low energy emissions. It is reported this detector was in service until September 1995. Reported 

MDAs for various radionuclides are shown in Tables 4.2.1-1 and 4.2.1-2. MDAs and reporting levels 

were the same. In the early 1970s, whole body counting using an array of NaI detectors was 

implemented. 

Table 4.2.1-1. Whole body counting.

Period Nuclide MDA, nCi MDA method

~1960 - ?

Ce-144 29

~3X standard deviation 

of expected count rate 

in energy region

Ua 62

I-131 1.4

Ru-106 6.1

Cs-137 1.0

Zr/Nb-95 2.2

Zn-65 5.1

Ba/La-140 9.3

a. Listed as U, but based on measurement of U-235. A result greater 

than the MDA would certainly be a false positive unless it was 

associated with a major intake. Other counts and urinalyses would be 

expected.

Table 4.2.1-2. Current MDAs for whole body counting.

Radionuclide MDA (nCi) Radionuclide MDA (nCi) Radionuclide MDA (nCi)

Mn-54 3.4 Sb-125 14 Eu-152 18

Co-60 2.9 Cs-134 3.8 Eu-154 8.4

Zn-65 6.1 Cs-137 4.1 U-235 14

Ru-106 36 Ce-144 69 Np-237 14

(WSRC-IM-90-139 indicates five 4” × 4” detectors were used, while WSRC-RP-95-234 says that four 

4” × 5” detectors were used.) Additionally, in the mid-1980s, whole body counting using stand-up 

geometry and large (4” × 4” × 16” or 5” × 3” × 16”) NaI detectors was implemented. MDAs for these 

counting systems were generated individually for counts by processing software. Many forms for 

reporting the results of whole body counts were used throughout the years. Some had the MDAs 

listed on the form; others just reported “detected” radionuclides. If MDAs are not shown, use the ones 

in Table 4.2.1-1.

One of the forms has columns headings: nuclide, energy, B.G. c/m, 40 cm Arc net c/m, and results. 

The first column and the results column are the columns of concern; the other columns involve 

preliminary steps. The 137Cs results are reported as ncCs, meaning nCi of 137Cs, and pcCs/gmK, 

apparently meaning pCi of 137Cs per gram of potassium calculated from the 40K result. The 

importance for the latter result has not been determined yet, but the dose reconstructor can just be 

concerned with the ncCs value. The letters A, B, C, D under the nuclide column appear to be 

placeholders for that energy region of the spectrum, apparently because there was not a single 

radionuclide of interest that had a peak in that region (perhaps there were none or more than one). 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 73 of 248

Hopefully no peaks show up in those regions or a note will be provided under results explaining what 

radionuclide the peak resulted from. 

Another one of the in vivo counting forms has two main blocks of data, one under WBC (NAI) referring 

to the whole body count results, the other under Chest Count (NAI) referring to the results from the 

chest counter. The columns list the nuclide and other intermediate steps in the calculation and the 

last column gives the MDA for the nuclide, meaning the count did not detect the nuclide above the 

MDA. If there is no value in the last column, then there should be a special listing just below the 

columns. This is where the result in nCi is shown for a nuclide that is detected. One should expect 

40K to be listed in this location as a minimum. 

For special investigational counts, the MDAs are stated to be approximately 0.71 times these values 

for whole body counting.

Fallout affected everyone in North America, and body burdens of 137Cs measurable in whole body 

counters were common in the 1960s and 1970s. NCRP Report No. 94 (NCRP 1987) provides mean 

body burdens of 137Cs for the United States for the years most likely to produce interference with 

occupational whole body count results. Those values are listed in Table 4.2.1-3. If whole body count 

results show detection of only 40K and 137Cs and the 137Cs result is less than the values given in Table 

4.2.1-3, the 137Cs results can be assumed to be due to fallout. If other fission or activation product 

radionuclides are present in either the whole body count or a recent urinalysis for 90Sr, then it is 

claimant-favorable to assume the 137Cs is from occupational sources. 

Table 4.2.1-3. Mean body burdens of 137Cs from fallout in 

the United States.a

Year Body burden (nCi) Year Body burden (nCi)

1953 0.27 1966 9.7

1954 1.1 1967 5.6

1955 2.2 1968 3.5

1956 4.3 1969 2.7

1957 5.1 1970 2.7

1958 6.5 1971 2.7

1959 8.1 1972 2.7

1960 6.8 1973 2.7

1961 4.6 1974 1.6

1962 6.0 1975 1.1

1963 11 1976 1.6

1964 19 1977 1.1

1965 16

a. From NCRP Report No. 94.

4.2.2

Chest counting was initially performed using a 5” diameter thin NaI detector from 1966 to around 

1971. No MDA information could be located in SRS technical documentation for this method of 

analysis. 

Chest Counting

In 1971, a phoswich detection system was implemented that used a NaI/CsI sandwiched detector. 

This system was modified to a dual phoswich system in 1972, which remained in use until around 

1987 or 1988. MDAs were listed on results forms that varied from count to count but were in the 

neighborhood of 0.2 nCi for 241Am, 5 nCi for 238Pu, 10 nCi for 239Pu, and 5 nCi for 244Cm; however, 

based on operating experience over the years, the values for all but the 241Am seem overly optimistic 
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due to noise in the 1-20 keV part of the spectrum. The method of accounting for chest wall thickness 

may also be a factor in the difference between stated MDAs and present experience. However, 

phoswich systems had better low-energy background and better detection efficiency than germanium 

systems. Phoswich systems had poor resolution and could not discriminate between 238Pu, 239Pu, or 

244Cm. 

In general chest counting MDAs should not be used to determine potential undetected intakes of 

plutonium if urinalysis data are available.

The current six detector solid state planar germanium array was implemented in 1987 or 1988.

MDAs for later periods have been generated individually for counts by processing software, with 

typical values shown in Table 4.2.2-1, based on a chest wall thickness of 2.5 cm. However, MDAs 

are often listed for each count as part of the measurement results and the count-specific MDAs should 

be used when available. The uncertainty in the MDA calculation has been estimated, according to 

WSRC-IM-90-139, rev. 8, as ±46%. The confidence level was not stated but would appear to be one 

standard deviation. However, uncertainty in chest wall thickness can have a large impact on MDA 

determination. Using weight and height measurements on 1100 male workers at Hanford in 2002-

2003, and using the chest wall thickness formula in the SRS technical basis manual, the 50th 

percentile chest wall thickness was 3.7 cm and the 95th-percentile chest wall thickness was 5.9 cm. 

The MDAs for radionuclides with low-energy photons affected by variations in chest wall thickness for 

the 50th and 95th-percentile male workers are shown in Table 4.2.2-2. 

Table 4.2.2-1. Current MDAs for chest counting (2.5 cm chest wall thickness).

Radionuclide MDA (nCi) Radionuclide MDA (nCi) Radionuclide MDA (nCi)

Ce-144 0.31 U-235 0.10 6% Pu 96

Eu-152 0.056 U-236 89 12% Pu 70

Th-228 3.2 U-238 1.1 Am-241 0.10

Th-232 31 Np-237 0.31 Am-243 0.12

DU 1.2 Pu-238 58 Cm-242 28

RU 8.3 Pu-239 130 Cm-244 37

HEU 5.2 Pu-240 47 Cf-252 32

U-234 30 3% Pu 110

For special investigational chest counts, the MDAs are stated to be approximately 0.77 times these 

values.

For 241Am results associated with an intake of 6% to 12% 240Pu mixture, the dose reconstructor can 

assume that the plutonium isotopes behave the same in the lung 

Table 4.2.2-2. Current MDAs for chest 

counting 50th-percentile and 95thpercentile male workers.

as the 241Am. This assumption does 

not hold for systemic americium, however. 

Radionuclide

MDA (nCi)

50th

MDA (nCi)

95th

Am-241 0.13 0.20

Cm-244 78 310

Pu-238 130 580

Pu-239 290 1,300

Pu-240 110 470

Cf-252 85 500
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4.3 PERSONAL AIR SAMPLING DATA AS APPLICABLE

The objective of breathing zone air monitoring is to measure the concentration of radioactive material 

in the air that a worker breathes for the purpose of calculating exposure [in Derived Air Concentrationhours (DAC-h)]. Intakes calculated from air monitoring were assigned only if bioassay data are 

unavailable or inadequate. Breathing zone air monitoring must sample air that is representative of the 

air in the worker's breathing zone. This was accomplished by placing the sampler as close as possible 

to the worker's nose and mouth. Fixed breathing zone air samplers are typically placed at face height 

in front of hoods and glove boxes. Portable breathing zone air samplers, often referred to as personal 

air samplers (PAS), are attached to the worker's clothing in the neck region. PAS have been used 

when it was not practical to use regional air samplers (RAS) (such as work outside of buildings), 

inside plastic suits used in highly contaminated environments, and for operational studies.

Air sample records were considered workplace monitoring records as opposed to intake monitoring 

records and hence were not maintained in the same records collection as bioassay records. Air 

sample records were associated with each facility separately and were stored as facility records. 

Boxes containing workplace monitoring records can be retrieved from storage at the federal 

repository; however, locating a specific set of air sample records would be very labor consuming and 

should be considered only as a last resort. Correlation between air sample concentrations in given 

rooms or work locations and a specific person would be difficult. 

4.4 INTERFERENCES, UNCERTAINTY, AND USE OF REPORTING LEVELS

4.4.1

Excreta samples were collected in the workplace. Because the levels of activity that are significant in 

excreta samples, especially urine samples, were generally below detectability on workplace personnel 

detectors, contamination of samples from the worker’s hands or clothing is a possibility. Hanford 

found a decrease in detectable plutonium bioassay results after switching to home collection. 

Laboratory contamination and mix-up of samples in the laboratory are also a possibility, although 

laboratory Quality Control procedures and performance of test samples was designed to minimize this 

source of contamination. 

Contamination of Samples

It is likely that a contaminated sample will show up as an obvious outlier in the dataset for a given 

worker. If the dataset shows an unusually high urinalysis result for a radionuclide other than tritium, 

and if follow-up samples were collected that were not consistent with the high result, then the high 

result may be considered an outlier. However, if the result is not obviously an outlier, then it is 

claimant-favorable to assume the result is real. 

For in vivo measurements, contamination can occur as external to the body or, in the case of chest 

counting, as external to the lung. If a follow-up in vivo count is obtained the same day or within a few 

days that shows a dramatic decrease in activity or no detectable activity, then external contamination 

can be assumed. Radon progeny and medical diagnostic or therapeutic procedures involving 

radionuclides can cause interferences to in vivo measurements, especially for NaI detectors. 

However, unless the count was invalidated or noted as being influenced by such interferences, the 

results should be used as recorded. 
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4.4.2

Special consideration for plutonium intakes

Use of Reporting Levels and Less-Than Values

Because plutonium stays in the body for a very long time, and because the urine excreta curve 

(activity per day excreta versus days after intake) has a small slope beyond the first year, an intake of 

plutonium that might have been missed in the 1950s or 60s because of poor detection capability or 

other reasons can still be confirmed or otherwise by urinalysis obtained years later. Additionally, lowlevel chronic or frequent-intermittent intakes not originally detectable, if continued for many years, will 

lead to increasing rates of excreta over time. This condition has implications for the determination of 

potential missed intakes.

Table 4.4.2-1 provides the maximum potentially missed chronic intakes of 239Pu based on a urinalysis 

MDA of 0.1 dpm/d (0.07 dpm/L) for various periods of chronic intake. The urine sample is assumed to 

be collected on the last day of the period (under Duration of Intake). This approach provides the 

maximum potentially missed chronic intake regardless if there were missed samples or samples with 

high MDAs during the period. Once the 239Pu intake is obtained, standard isotopic ratios (e.g., 6% 

240Pu or 12% 240Pu) can be applied to get the total intakes. The daily intake rates and cumulative 

intakes are linear with the applicable MDA so the values in Table 4.4.3-1 can be adjusted for different 

MDAs (or reporting levels) by the ratio of the applicable MDA to 0.1 dpm/d. 

Chronic intakes for radionuclides associated with in vitro monitoring

A similar approach can be made for 237Np, Table 4.4.2-2, and Type S uranium, Table 4.4.2-3. The 

uranium table also provides the daily intake rate and cumulative intakes for Type F and M uranium. 

For the latter two inhalation types, equilibrium between daily intake and excretion is reached within a 

year or two. 

Note: There is an important difference in interpretation of the results of Tables 4.4.2-1 through 4.4.2-

5 and 4.4.2.8 through 4.4.2-11 depending on how quickly equilibrium is established between daily

intake and daily excretion. For type M and S plutonium and transuranics and type S uranium, the 

excretion rate builds slowly and the MDAs of the last monitoring periods are generally most important 

and can be used as is. 

The phrase bioassay source term is used here to describe those source terms, for which not all of the 

components are measured, but some are inferred from the measurement of others. To address 

calculations of missed dose from different bioassay source terms in a consistent and efficient manner, 

the default assumption is that the worst-case identified source term was encountered during the 

assumed exposure period. For plutonium at SRS, this means that the default mixture consists of 10-y 

aged, 12% Pu and that the chosen absorption type is the one that causes the largest dose to the 

target organ. 

Missed dose from radionuclides not routinely considered by the bioassay method would not be 

reconstructed based on workplace location only. For example at SRS, bioassay result interpretation 

of plutonium routinely includes consideration of americium-241, but does routinely include 

Table 4.4.2-1. 239Pu chronic inhalation intake assessment based on MDA urinalysis on last day of the 

period.

Solubility: Type M Type S

Particle size: 5 5

Analytical MDA: 0.1 dpm/d 0.1 dpm/d

Duration of intake Daily intake rate Cumulative intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

1 365 36.7 1.34E+04 6.03 1536 5.61E+05 253
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2 730 25.6 1.87E+04 8.42 778 5.68E+05 256

3 1095 20.6 2.26E+04 10.2 518 5.67E+05 256

4 1461 17.8 2.60E+04 11.7 390 5.70E+05 257

5 1825 15.9 2.89E+04 13.0 315 5.75E+05 259

6 2190 14.5 3.18E+04 14.3 267 5.85E+05 263

7 2556 13.4 3.43E+04 15.4 233 5.96E+05 268

8 2922 12.6 3.68E+04 16.6 211 6.17E+05 278

9 3287 11.9 3.91E+04 17.6 189 6.21E+05 280

10 3650 11.4 4.15E+04 18.7 174 6.35E+05 286

12 4383 10.45 4.58E+04 20.6 151 6.62E+05 298

14 5113 9.73 4.97E+04 22.4 134 6.85E+05 309

15 5475 9.43 5.16E+04 23.3 128 7.01E+05 316

16 5844 9.15 5.35E+04 24.1 122 7.13E+05 321

18 6574 8.66 5.69E+04 25.6 112 7.36E+05 332

20 7300 8.24 6.02E+04 27.1 104 7.60E+05 342

25 9125 7.41 6.76E+04 30.5 88.8 8.11E+05 365

30 10950 6.78 7.43E+04 33.5 78.4 8.59E+05 387

35 12775 6.29 8.04E+04 36.2 70.6 9.03E+05 407

40 14600 5.89 8.61E+04 38.8 64.7 9.35E+05 421

45 16425 5.56 9.14E+04 41.2 59.9 9.85E+05 443

50 18250 5.28 9.64E+04 43.4 56.1 1.02E+06 462

Table 4.4.2-2. 237Np chronic inhalation intake assessment 

based on MDA urinalysis on last day of the period.

Solubility: Type M

Particle size: 5

Analytical MDA: 0.1 dpm/d (0.1dpm/1.5L)

Duration of intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi)

1 365 4.61 1.68E+03 7.58E-01

2 730 4.05 2.96E+03 1.33E+00

3 1095 3.77 4.13E+03 1.86E+00

4 1461 3.57 5.22E+03 2.35E+00

5 1826 3.41 6.23E+03 2.80E+00

6 2190 3.28 7.18E+03 3.24E+00

7 2556 3.17 8.10E+03 3.65E+00

8 2922 3.08 9.00E+03 4.05E+00

9 3287 3.00 9.86E+03 4.44E+00

10 3652 2.93 1.07E+04 4.82E+00

12 4383 2.82 1.24E+04 5.57E+00

14 5113 2.73 1.40E+04 6.29E+00

16 5844 2.65 1.55E+04 6.98E+00

18 6574 2.59 1.70E+04 7.67E+00

20 7305 2.53 1.85E+04 8.33E+00

25 9131 2.42 2.21E+04 9.95E+00

30 10958 2.33 2.55E+04 1.15E+01

35 12784 2.26 2.89E+04 1.30E+01

40 14610 2.20 3.21E+04 1.45E+01

45 16436 2.15 3.53E+04 1.59E+01

50 18263 2.11 3.85E+04 1.74E+01

Table 4.4.2-3. 234U chronic inhalation intake assessment based on MDA urinalysis on last day of the 

period.a

Solubility: Type F Type M

Particle size: 5 5

Analytical MDA: 1 dpm/d (1dpm/1.5L) 1 dpm/d (1 dpm/1.5L)

Duration of intake Daily intake rate Cumulative intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

1 365 3.73 1.36E+03 6.13E-01 15.9 5.80E+03 2.61E+00
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2 730 3.71 2.71E+03 1.22E+00 15.1 1.10E+04 4.97E+00

3 1095 3.70 4.05E+03 1.83E+00 15.0 1.64E+04 7.40E+00

4 1461 3.69 5.39E+03 2.43E+00 14.9 2.18E+04 9.81E+00

5 1826 3.68 6.72E+03 3.03E+00 14.9 2.72E+04 1.22E+01

6 2190 3.68 8.06E+03 3.63E+00 14.8 3.25E+04 1.46E+01

7 2556 3.67 9.38E+03 4.23E+00 14.8 3.79E+04 1.70E+01

8 2922 3.66 1.07E+04 4.82E+00 14.8 4.32E+04 1.95E+01

9 3287 3.66 1.20E+04 5.42E+00 14.8 4.85E+04 2.19E+01

10 3652 3.65 1.33E+04 6.00E+00 14.8 5.39E+04 2.43E+01

12 4383 3.65 1.60E+04 7.21E+00 14.7 6.45E+04 2.90E+01

14 5113 3.64 1.86E+04 8.38E+00 14.7 7.52E+04 3.39E+01

16 5844 3.64 2.13E+04 9.58E+00 14.7 8.57E+04 3.86E+01

18 6574 3.63 2.39E+04 1.07E+01 14.65 9.63E+04 4.34E+01

20 7305 3.63 2.65E+04 1.19E+01 14.6 1.07E+05 4.81E+01

25 9131 3.62 3.31E+04 1.49E+01 14.6 1.33E+05 6.01E+01

30 10958 3.62 3.97E+04 1.79E+01 14.6 1.60E+05 7.20E+01

35 12784 3.61 4.62E+04 2.08E+01 14.6 1.86E+05 8.39E+01

40 14610 3.61 5.27E+04 2.38E+01 14.6 2.13E+05 9.58E+01

45 16436 3.60 5.92E+04 2.67E+01 14.5 2.39E+05 1.08E+02

50 18263 3.60 6.57E+04 2.96E+01 14.5 2.65E+05 1.20E+02

Solubility: Type S

Particle size: 5

Analytical MDA: 1 dpm/d (1 dpm/1.5L)

Duration of intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi)

1 365 432 1.58E+05 7.10E+01

2 730 319 2.33E+05 1.05E+02

3 1095 268 2.93E+05 1.32E+02

4 1461 240 3.51E+05 1.58E+02

5 1826 221 4.04E+05 1.82E+02

6 2190 208 4.56E+05 2.05E+02

7 2556 199 5.09E+05 2.29E+02

8 2922 192 5.61E+05 2.53E+02

9 3287 186 6.11E+05 2.75E+02

10 3652 181 6.61E+05 2.98E+02

12 4383 174 7.63E+05 3.44E+02

14 5113 168 8.59E+05 3.87E+02

16 5844 164 9.58E+05 4.32E+02

18 6574 160 1.05E+06 4.74E+02

20 7305 157 1.15E+06 5.17E+02

25 9131 151 1.38E+06 6.21E+02

30 10958 147 1.61E+06 7.26E+02

35 12784 144 1.84E+06 8.29E+02

40 14610 142 2.07E+06 9.35E+02

45 16436 140 2.30E+06 1.04E+03

50 18263 139 2.54E+06 1.14E+03

a. Same results can be used for U-235, U-238, or gross uranium alpha results. Total intake of uranium isotopes will have 

to be adjusted depending on the enrichment of the uranium (e.g., depleted, natural, enriched) and the result from the 

analytical method (specific isotope, gross alpha, or total mass).

consideration of other radionuclides. Am-241 would be included in the plutonium missed dose 

assessment. Cs-137 is not considered in SRS plutonium bioassay result interpretation, so Cs-137 

would not be included in the plutonium missed dose calculation even if it was listed as a radionuclide 

in the workplace. However, if there is other evidence of Cs-137 exposure (FP urine data, WBC, 

incident data, etc.), missed dose for Cs-137 (or the applicable radionuclide) would be calculated 

based on the additional data.
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Using the worst-case identified bioassay source term, the minimum detectable dose due to 

undetected or unreported radioactivity in in vitro bioassay results is calculated from the detection 

threshold (the MDA, the recording or the reporting level, whichever is the applicable value in the 

record of interest). One half of the detection threshold is selected to calculate the mode value of 

missed dose. It is also assumed that it’s possible that there is no dose received when in vitro

bioassay results are all less than the detection threshold, although the probability associated with this 

possibility is assumed to be low, because of the fact that the worker was included in the in vitro

bioassay program. 

Based on the above, a triangular distribution is used to assign missed dose from in vitro bioassays 

that are all below their associated detection thresholds. The minimum dose is zero, the mode dose is 

one half of the maximum dose and the maximum dose is based on a chronic exposure where the 

bioassay result(s) are equal to the detection threshold for the radionuclide of interest. 

Alternative methods for assigning missed dose have been developed for tritium urinalysis and the in 

vivo programs.

For radionuclides and inhalation types that reach equilibrium quickly, if the reporting level changes 

throughout the history of a worker’s monitoring, the daily intake rate and cumulative intake will have to 

be calculated separately for each period of different reporting levels, and the total intake becomes the 

sum of the intakes for the various periods. 

Table 4.4.2-4 provides the same information for 90Sr, which can be applied to the radiostrontium 

analysis. 

Because, as discussed in 4.1.1, fission product analysis might be interpreted as 144Ce, Table 4.4.2-5 

provides the information for 144Ce.

Tritium reaches equilibrium very rapidly (within 2 months) so a table is not needed. An intake rate of 

6.09 × 106 dpm/d (2.74 μCi/d) will produce a daily urinary excretion of 1 μCi/L (ICRP 1997, table 

A.1.10). 

Special consideration for fission/activation product urinalysis

Fission product urinalysis at SRS appeared to be fairly sensitive prior to 1960 with a reporting level of 

30 dpm/750 ml. The reporting level increased in 1960 and again in 1965 when it appeared to be 1 

nCi/1.5 L. (2200 dpm/1.5 L). It is not clear whether fission product urinalyses were scheduled on a 

routine basis or were only incident driven. In either case, if no information is available regarding the 

likely time or period of exposure, the calculated missed intakes and doses for extended periods from 

the urinalysis results alone are unbelievably large (these assumed intakes would have begun to effect 

radiation survey instrument readings as the activity in an employee’s body increased over the years). 

Although beginning in the 1960s, in vivo counting was employed for detecting photon-emitting 

fission/activation products, it does not appear that all workers were monitored. In addition, the in vivo

results do not appear to be sufficiently frequent for most workers to provide more than a bounding 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 80 of 248

Table 4.4.2-4. 90Sr chronic inhalation intake assessment based on MDA urinalysis on last day of the 

period.

Solubility: Type F

Particle size: 5

Analytical MDA: 100 dpm/d (100dpm/1.5L)

Duration of intake Daily intake rate Cumulative intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

1 365 441 1.61E+05 7.25E+01

2 730 433 3.16E+05 1.42E+02

3 1095 428 4.69E+05 2.11E+02

4 1461 425 6.21E+05 2.80E+02

5 1826 421 7.69E+05 3.46E+02

6 2190 419 9.18E+05 4.13E+02

7 2556 417 1.07E+06 4.80E+02

8 2922 415 1.21E+06 5.46E+02

9 3287 413 1.36E+06 6.12E+02

10 3652 412 1.50E+06 6.78E+02

12 4383 410 1.80E+06 8.09E+02

14 5113 410 2.10E+06 9.44E+02

16 5844 406 2.37E+06 1.07E+03

18 6574 405 2.66E+06 1.20E+03

20 7305 404 2.95E+06 1.33E+03

25 9131 403 3.68E+06 1.66E+03

30 10958 402 4.41E+06 1.98E+03

35 12784 401 5.13E+06 2.31E+03

40 14610 400 5.84E+06 2.63E+03

45 16436 400 6.57E+06 2.96E+03

50 18263 400 7.31E+06 3.29E+03

Table 4.4.2-5. 144Ce chronic inhalation intake assessment based on MDA urinalysis on last day of the 

period.

Solubility: Type M Type S

Particle size: 5 5

Analytical MDA: 100 dpm/d (100dpm/1.5L) 100 dpm/d (100 dpm/1.5L)

Duration of intake Daily intake rate Cumulative intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

1 365 4.62E+05 1.68E+08 7.59E+04 1.93E+07 7.05E+09 3.17E+06

2 730 3.09E+05 2.26E+08 1.02E+05 1.10E+07 8.02E+09 3.61E+06

3 1095 2.74E+05 3.00E+08 1.35E+05 8.96E+06 9.81E+09 4.42E+06

4 1461 2.62E+05 3.83E+08 1.73E+05 8.25E+06 1.21E+10 5.43E+06

5 1826 2.58E+05 4.71E+08 2.12E+05 7.97E+06 1.46E+10 6.56E+06

6 2190 2.57E+05 5.62E+08 2.53E+05 7.86E+06 1.72E+10 7.76E+06

7 2556 2.56E+05 6.54E+08 2.95E+05 7.82E+06 2.00E+10 9.00E+06

8 2922 2.56E+05 7.47E+08 3.37E+05 7.80E+06 2.28E+10 1.03E+07

9 3287 2.56E+05 8.40E+08 3.78E+05 7.79E+06 2.56E+10 1.15E+07

10 3652 2.56E+05 9.33E+08 4.20E+05 7.79E+06 2.84E+10 1.28E+07

12 4383 2.56E+05 1.12E+09 5.05E+05 7.79E+06 3.41E+10 1.54E+07

14 5113 2.56E+05 1.31E+09 5.89E+05 7.79E+06 3.98E+10 1.79E+07

16 5844 2.56E+05 1.49E+09 6.73E+05 7.79E+06 4.55E+10 2.05E+07

18 6574 2.56E+05 1.68E+09 7.57E+05 7.79E+06 5.12E+10 2.31E+07

20 7305 2.56E+05 1.87E+09 8.41E+05 7.79E+06 5.69E+10 2.56E+07

25 9131 2.56E+05 2.33E+09 1.05E+06 7.79E+06 7.11E+10 3.20E+07

30 10958 2.56E+05 2.80E+09 1.26E+06 7.79E+06 8.53E+10 3.84E+07

35 12784 2.56E+05 3.27E+09 1.47E+06 7.79E+06 9.96E+10 4.48E+07

40 14610 2.56E+05 3.73E+09 1.68E+06 7.79E+06 1.14E+11 5.13E+07

45 16436 2.56E+05 4.20E+09 1.89E+06 7.79E+06 1.28E+11 5.77E+07

50 18263 2.56E+05 4.67E+09 2.10E+06 7.79E+06 1.42E+11 6.41E+07
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 81 of 248

estimate of what might have occurred prior to the late 1980s when the Canberra FASTSCANTM 

counter became available at SRS.

For employees, whose records indicate exposure to fission/activation products, but have no positive 

bioassay results a missed dose is calculated. When it is necessary to estimate missed 

fission/activation product dose, the approach to estimating worker dose from exposure to fission 

products in the SRS nonreactor areas is to calculate a worst-case, chronic intake based on 

descriptions of the air monitoring and radiation safety programs. This worst case takes into account 

the fact that there was a robust radiation safety program and that area surveys and air monitoring for 

radioactivity were routine at SRS. SRS air monitoring was reportedly able to detect levels at 10% of 

their MOL (maximum occupational limit) or RCG (radioactivity concentration guide). For unknown 

beta emitters this limit appears to have been initially set at 2 × 10-10 uCi/cc and later raised to 3 × 10-10

uCi/cc. The “SRP Health Physics Procedure Manual for the 200 Areas," circa 1960, stated:

Respiratory protection is required whenever air sample results indicate activity above the 

maximum breathing air limit or wherever air activity is suspected due to the nature of the 

work, such as welding of contaminated equipment, decontamination of spills, etc.

Assault masks were required when airborne radioactivity exceeded the RCG and self-contained air 

masks were required with airborne radioactivity exceeded 100 times the RCG and may have been 

used at even lower air concentrations.

Based on the above, it is conceivable in the 200 Areas that exposures to fission products could have 

occurred at levels up to the RCG without respiratory protection. While it is unlikely that workers were 

chronically exposed all day, every day without respiratory protection, this is exactly what is assumed 

for this bounding missed intake calculation. This missed intake calculation is assumed to apply from 

1954 through 1988. The first SRS reactor start up was on December 28, 1953; prior to this there was 

no report of handling fission products at SRS. DOE Order 5480.11 became effective in 1989, and this 

required more emphasis to be placed on controlling and monitoring internal doses. The internal 

monitoring data after 1988 should be more robust and missed dose for this later period is better 

addressed with any available bioassay data from this later period.

The annual maximum missed chronic intake (I) of fission/activation products is estimated from the 

RCG and the assumed breathing rate in a work year (BR), which is 2400 cubic meters per year. The 

larger RCG is used, because it is not clear exactly when the lower value applied.

I = (RCG)(BR) = (3E − 10uCi / cc)(2.4E9cc / year ) = 0.72uCi / year

This is equal to a missed intake rate of about 4400 dpm/day. Although it is not credible to have an 

intake that consists of only a single fission or activation product in most areas at SRS, a bounding 

organ dose can be calculated from the information in Table 4.4.2-6. The calculated organ dose is an 

upper bound (constant dose distribution type for IREP). 

For a more realistic bound of the missed dose reconstruction, the total intake is apportioned to a more 

realistic mix of fission/activation products. Table 4.4.2-7 provides a summary of assumed intakes for 

more typical mixes of fission/activation products. The default absorption type is listed after the 

radionuclide. The intake rate assignments in this table would be more reasonable for long-term 

exposures to airborne fission/activation products. The intakes are based on the assumption that from 

a dose perspective Co-60, Sr-90, Ru-106, Cs-137, Cs-137 and Ce-144 dominate the source term. It 

is also assumed that Co-60 would be unlikely to contribute to more than 15% of the long term intake 

from airborne activity and that no radionuclide would contribute to more than 50% of intake from this 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 82 of 248

Table 4.4.2-6. Parameters for simplified upper bound dose from fission/activation products.

For these organs/tissues:

Assume 4,400 

dpm/day intake of:

Adrenals, breast, heart wall, thymus, extrathoracic, extrathoracic2, lymph nodes 

(extrathoracic), lymph nodes (thoracic), esophagus

Co-60 S

Urinary bladder, brain, gall bladder, kidneys, muscle, ovaries, pancreas, testes, thyroid, 

stomach, small intestine, upper large intestine, skin, spleen, uterus, gonads

Ru-106 F

Colon Ru-106 M

Lower large intestine, lung Ru-106 S

Liver, extrathoracic1 Ce-144 M

Bone surface, red bone marrow Sr-90 F

Table 4.4.2-7. Parameters for reconstructing upper bound dose from fission/activation products.

For these organs/tissues:

Intake assumptions (in dpm/day)

2200 880 660 660

Adrenals, brain, breast, gall bladder, heart wall, kidneys, 

muscle, ovaries, pancreas, thyroid, stomach, small intestine, 

skin, spleen, thymus, uterus, esophagus, gonads

Ru-106 F Cs-137 F Ce-144 M Co-60 S

Lymph nodes (extrathoracic) Ru-106 S Cs-137 F Ce-144 M Co-60 S

Urinary bladder, testes, Ru-106 F Cs-137 F Ce-144 M Sr-90 F

Upper large intestine Ru-106 F Ce-144 M Cs-137 F Co-60 S

Extrathoracic, extrathoracic2, lung, lymph nodes (thoracic) Ru-106 S Ce-144 M Cs-137 F Co-60 S

Colon, lower large intestine Ru-106 S Ce-144 M Cs-137 F Sr-90 F

Liver Ce-144 M Ru-106 F Cs-137 F Co-60 S

Extrathoracic1 Ce-144 M Sr-90 F Cs-137 F Ru-106 F

Red bone marrow Sr-90 F Ru-106 F Cs-137 F Ce-144 M

Bone surface Sr-90 F Ce-144 M Ru-106 F Cs-137 F

long term exposure. This more realistic intake source term also provides an upper bound organ dose 

(constant dose distribution type for IREP).

This method of missed dose assignment is not meant to restrict the user from assessing missed dose 

with in vivo count data when it is available or when organs other than the red bone marrow or bone 

surface are involved. The results of the in vivo missed intake calculations for the radionuclides listed 

in Table 4.4.2-6 are similar in magnitude to the missed intakes based on the RCG.

For the reactor areas assign annual dose(s) equal to the assigned tritium dose(s) and enter it into the 

IREP spreadsheet as “electrons >15 keV” to account for missed fission/activation product doses. 

The basis for this is that tritium dominated the SRS internal radiation doses in the reactor areas.

Chronic intakes for radionuclides common in whole body counts

Because most fission and activation product radionuclides emit gamma rays, whole body counters 

have the potential for measuring many of them at once. This is wonderful when the emphasis is on 

detecting and quantifying intakes of mixed fission and activation products, but provides a special 

challenge when the emphasis is on estimating intakes that were not measured. The chronic intake 

rates that would produce a body burden of the major fission and activation products just below each 

radionuclide’s approximate MDA are provided in Tables 4.4.2-8 through 4.4.2-11. (MDAs are from 

Table 4.2.1-1 which were applicable for most of the years; these MDAs are approximate because 

actual MDAs were calculated separately for each person and each count. The intakes given in the 

tables can be linearly adjusted for changes in the MDA applicable at the time; i.e., the adjusted intake 

equals the intake in the table times the ratio of new MDA to table MDA.)
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 83 of 248

Table 4.4.2-8. 137Cs chronic inhalation intake 

assessment based on MDA whole body count on 

last day of the period. Solubility: Type F

Particle size: 5

Analytical MDA: 1 nCi

Duration of intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi)

1 365 35.7 1.30E+04 5.87E+00

2 730 32.6 2.38E+04 1.07E+01

3 1095 32.3 3.54E+04 1.59E+01

4 1461 32.2 4.70E+04 2.12E+01

5 1826 32.2 5.88E+04 2.65E+01

6 2190 32.2 7.05E+04 3.18E+01

7 2556 32.2 8.23E+04 3.71E+01

8 2922 32.2 9.41E+04 4.24E+01

9 3287 32.2 1.06E+05 4.77E+01

10 3652 32.2 1.18E+05 5.30E+01

12 4383 32.2 1.41E+05 6.36E+01

14 5113 32.2 1.65E+05 7.42E+01

16 5844 32.2 1.88E+05 8.48E+01

18 6574 32.2 2.12E+05 9.54E+01

20 7305 32.2 2.35E+05 1.06E+02

25 9131 32.2 2.94E+05 1.32E+02

30 10958 32.2 3.53E+05 1.59E+02

35 12784 32.2 4.12E+05 1.85E+02

40 14610 32.2 4.70E+05 2.12E+02

45 16436 32.2 5.29E+05 2.38E+02

50 18263 32.2 5.88E+05 2.65E+02

Table 4.4.2-9. 144Ce chronic inhalation intake 

assessment based on MDA whole body count on 

last day of the period. Solubility: Type M

Particle size: 5

Analytical MDA: 29 nCi

Duration of intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi)

1 365 3.80E+03 1.39E+06 6.25E+02

2 730 2.85E+03 2.08E+06 9.37E+02

3 1095 2.60E+03 2.85E+06 1.28E+03

4 1461 2.52E+03 3.68E+06 1.66E+03

5 1826 2.49E+03 4.55E+06 2.05E+03

6 2190 2.47E+03 5.41E+06 2.44E+03

7 2556 2.47E+03 6.31E+06 2.84E+03

8 2922 2.47E+03 7.22E+06 3.25E+03

9 3287 2.47E+03 8.12E+06 3.66E+03

10 3652 2.47E+03 9.02E+06 4.06E+03

12 4383 2.47E+03 1.08E+07 4.88E+03

14 5113 2.47E+03 1.26E+07 5.69E+03

16 5844 2.47E+03 1.44E+07 6.50E+03

18 6574 2.47E+03 1.62E+07 7.31E+03

20 7305 2.47E+03 1.80E+07 8.13E+03

25 9131 2.47E+03 2.26E+07 1.02E+04

30 10958 2.47E+03 2.71E+07 1.22E+04

35 12784 2.47E+03 3.16E+07 1.42E+04

40 14610 2.47E+03 3.61E+07 1.63E+04

45 16436 2.47E+03 4.06E+07 1.83E+04

50 18263 2.47E+03 4.51E+07 2.03E+04

Table 4.4.2-10. 106Ru chronic inhalation intake assessment based on MDA whole body count on last 

day of the period.

Solubility: Type S Type F

Particle size: 5 5

Analytical MDA: 6.1 nCi 6.1 nCi

Duration of intake Daily intake rate Cumulative intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

1 365 1.10E+03 4.02E+05 1.81E+02 6.83E+02 2.49E+05 1.12E+02

2 730 8.60E+02 6.28E+05 2.83E+02 5.37E+02 3.92E+05 1.77E+02

3 1095 7.93E+02 8.68E+05 3.91E+02 5.00E+02 5.48E+05 2.47E+02

4 1461 7.69E+02 1.12E+06 5.06E+02 4.81E+02 7.03E+05 3.17E+02

5 1826 7.62E+02 1.39E+06 6.27E+02 4.76E+02 8.69E+05 3.92E+02

6 2190 7.57E+02 1.66E+06 7.47E+02 4.74E+02 1.04E+06 4.68E+02

7 2556 7.56E+02 1.93E+06 8.70E+02 4.73E+02 1.21E+06 5.45E+02

8 2922 7.56E+02 2.21E+06 9.95E+02 4.73E+02 1.38E+06 6.23E+02

9 3287 7.56E+02 2.48E+06 1.12E+03 4.72E+02 1.55E+06 6.99E+02

10 3652 7.56E+02 2.76E+06 1.24E+03 4.72E+02 1.72E+06 7.76E+02

12 4383 7.56E+02 3.31E+06 1.49E+03 4.72E+02 2.07E+06 9.32E+02

14 5113 7.56E+02 3.87E+06 1.74E+03 4.72E+02 2.41E+06 1.09E+03

16 5844 7.56E+02 4.42E+06 1.99E+03 4.72E+02 2.76E+06 1.24E+03

18 6574 7.56E+02 4.97E+06 2.24E+03 4.72E+02 3.10E+06 1.40E+03

20 7305 7.56E+02 5.52E+06 2.49E+03 4.72E+02 3.45E+06 1.55E+03

25 9131 7.56E+02 6.90E+06 3.11E+03 4.72E+02 4.31E+06 1.94E+03

30 10958 7.56E+02 8.28E+06 3.73E+03 4.72E+02 5.17E+06 2.33E+03

35 12784 7.56E+02 9.66E+06 4.35E+03 4.72E+02 6.03E+06 2.72E+03

40 14610 7.56E+02 1.10E+07 4.98E+03 4.72E+02 6.90E+06 3.11E+03

45 16436 7.56E+02 1.24E+07 5.60E+03 4.72E+02 7.76E+06 3.49E+03

50 18263 7.56E+02 1.38E+07 6.22E+03 4.72E+02 8.62E+06 3.88E+03
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 84 of 248

Table 4.4.2-11. 60Co chronic inhalation intake assessment based on MDA whole body count on last 

day of the period.

Solubility: Type M Type S

Particle size: 5 5

Analytical MDA: 1 nCia 1 nCia

Duration of intake Daily intake rate Cumulative intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

1 365 2.10E+02 7.67E+04 3.45E+01 1.57E+02 5.73E+04 2.58E+01

2 730 1.69E+02 1.23E+05 5.56E+01 1.04E+02 7.59E+04 3.42E+01

3 1095 1.53E+02 1.68E+05 7.55E+01 8.48E+01 9.29E+04 4.18E+01

4 1461 1.45E+02 2.12E+05 9.54E+01 7.59E+01 1.11E+05 5.00E+01

5 1826 1.40E+02 2.56E+05 1.15E+02 7.08E+01 1.29E+05 5.82E+01

6 2190 1.37E+02 2.91E+05 1.31E+02 6.78E+01 1.31E+05 5.91E+01

7 2556 1.35E+02 3.45E+05 1.55E+02 6.58E+01 1.68E+05 7.58E+01

8 2922 1.34E+02 3.92E+05 1.76E+02 6.44E+01 1.88E+05 8.48E+01

9 3287 1.33E+02 4.37E+05 1.97E+02 6.34E+01 2.08E+05 9.39E+01

10 3652 1.33E+02 4.86E+05 2.19E+02 6.26E+01 2.29E+05 1.03E+02

12 4383 1.33E+02 5.83E+05 2.63E+02 6.17E+01 2.70E+05 1.22E+02

14 5113 1.33E+02 6.80E+05 3.06E+02 6.11E+01 3.12E+05 1.41E+02

16 5844 1.33E+02 7.77E+05 3.50E+02 6.07E+01 3.55E+05 1.60E+02

18 6574 1.33E+02 8.74E+05 3.94E+02 6.04E+01 3.97E+05 1.79E+02

20 7305 1.33E+02 9.72E+05 4.38E+02 6.03E+01 4.40E+05 1.98E+02

25 9131 1.33E+02 1.21E+06 5.47E+02 6.01E+01 5.49E+05 2.47E+02

30 10958 1.33E+02 1.46E+06 6.56E+02 6.00E+01 6.57E+05 2.96E+02

35 12784 1.33E+02 1.70E+06 7.66E+02 6.00E+01 7.67E+05 3.46E+02

40 14610 1.33E+02 1.94E+06 8.75E+02 5.99E+01 8.75E+05 3.94E+02

45 16436 1.33E+02 2.19E+06 9.85E+02 5.99E+01 9.85E+05 4.43E+02

50 18263 1.33E+02 2.43E+06 1.09E+03 5.99E+01 1.09E+06 4.93E+02

a. Based on the assumption that the MDA for Co-60 was at least as good as the MDA for Cs-137.

Similar tables for other fission and activation product radionuclides are provided in Attachment D 

(D-3).

However, it is unreasonable to assume that a worker is exposed to all the possible radionuclides, 

each at an intake rate that results in a measurement just below its respective MDA. 

The mix of radionuclides in fission and activation products depends on the location of intakes and age 

of the contamination dating from the time of removal from the reactor. If a claimant’s work location is 

known and consistent throughout the employment period, then one of the radionuclides above can be 

used as an indicator radionuclide and intakes of the other radionuclides can be estimated using the 

activity ratios in the tables in Attachment A. For instance, for the canyon buildings, Ce-144

predominates among gamma-emitting fission products and also has a high whole body count MDA. A 

chronic intake for the appropriate employment period can be obtained from Table 4.4.2-9, adjusted for 

possible changes in MDA; then the intakes of other fission product radionuclides would be obtained 

by ratioing to the 144Ce activity fraction in Attachment A. The reasonableness of the calculation could 

be checked by comparing the body burden of another radionuclide (106Ru would be a good choice) 

determined in this manner to its MDA. The predicted result should be less than the MDA. 

Use one half of the MDA in these calculations if the worker had many whole body counts with all 

results below the MDAs for the various radionuclides (40K excepted). 

However, the above approach does not work well when the worker’s work location is unknown or was 

variable. In that case it is claimant-favorable to pick the fission or activation product that produces the 

highest dose to the organ of interest rather than using an indicator radionuclide. This approach may 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 85 of 248

also be used to expedite reconstruction. To assist with this determination, tables listing 50-year 

committed doses to various organs from chronic intakes producing MDA-level body burdens are 

provided in Attachment D (D 4). Because the major fission and activation products have similar 

effective half-lives in the body, the relative doses are similar for integration periods less than 50 years.

4.4.3

SRS began producing and handling radioactive material years after routine bioassay programs were 

developed and implemented at some major AEC sites, e.g. Hanford and Los Alamos. SRS has 

always had a robust radiation protection program. Consequently it is plausible that SRS did not have 

numbers of workers exposed to potential intakes of radioactive material while bioassay programs 

were being developed or were allowed significant exposure to airborne contamination without 

monitoring. This would be especially true for exposure to airborne plutonium, neptunium, trivalent 

actinides, and fission products. “Significance,” of course, would be defined by standard radiation 

protection policies of the time. Exceptions might occur for short-term workers that left employment 

without getting scheduled for or providing the required bioassay, or for workers that occasionally 

entered areas requiring monitoring but somehow did not get into the scheduling system. Thousands 

of bioassay results on monitored workers show, though, that detectable intakes were rare, with the 

possible exception of tritium and small intakes of fission products.

Unmonitored Worker

Under certain conditions, airborne effluents from one facility became air intakes for other facilities. 

Also workers were exposed to diluted effluents when walking between buildings or parking lots or 

while driving on the site. So workers in buildings who did not enter contaminated or airborne areas 

could have incurred environmental level intakes. 

These considerations lead to the following reasonable yet claimant-favorable assumptions for 

unmonitored workers at SRS.

• If a worker’s record shows no bioassay and no evidence of having worn a dosimeter, the 

unrecorded internal dose should be based on environmental intake only. If the worker’s file, 

including the CATI, indicates the plausible possibility of exposure to radioactive contamination 

or greater-than-ambient airborne concentrations (even if the worker did not wear a dosimetry, 

then the worker’s intakes can be determined by the information in the file or by the criteria in 

the following bullet. 

• If a worker was monitored for external dose, but had no bioassay, for each year of the 

assumed exposure period assign:

– annual missed dose(s) from tritium,

– annual missed dose(s) equal to the tritium dose(s) and enter it into the IREP spreadsheet 

as “electrons >15 keV” to account for fission products, and

– environmental doses from uranium, plutonium and 131I. 

In the latter case, it is reasonable to pick the fission or activation product associated with whole body 

counts that produces the largest dose to the organ of interest rather than using an indicator 

radionuclide. A spreadsheet has been developed to assist the reconstructor in making the choice of 

radionuclide and absorption type to estimate an upper bound of possible unrecorded intakes. 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 86 of 248

4.5 MAXIMIZING DOSE

For cases where the radionuclide(s) of interest do not concentrate in the organ(s) of interest and 

therefore are likely to be non-compensable, a worst-case internal dose may be assigned.

4.5.1

As described in the Technical Information Bulletin: Maximum Internal Dose Estimates for Savannah 

River Site (SRS) Claims (ORAUT-OTIB-0001) for likely non-compensable cases a maximizing 

radionuclide intake scenario was developed to ensure that claimant-favorable doses to organs in 

which the radionuclides of concern at SRS do not concentrate (referred to as “non-metabolic” for the 

purpose of this discussion) and digestive tract organs could be assigned without a detailed analysis of 

bioassay results. For simplicity and to ensure claimant favorability, the doses given in Table 4.5.1-1 

are assigned to all non-metabolic and digestive tract organs based on an assumed intake on the first 

day of the first year of employment, continuing to the last day of the year of diagnosis. These doses 

are based on the dose to the colon, which is the organ with the highest calculated doses among the 

organs considered “non-metabolic” or part of the digestive tract.

Assignment of Internal Doses (from Radionuclides Other than Tritium): Maximizing 

Approach for Dose Reconstructions

The Year 1 doses are applied for the first full or partial year of employment, and the remaining doses 

are applied for each successive year until the date of cancer diagnosis. Internal doses are not applied 

prior to 1953 because the site was not in operation.

Before the doses listed can be applied, all available bioassay results are reviewed to ensure that 

these doses are claimant favorable. If positive bioassay results could possibly lead to higher doses, 

the values given in this technical basis document are not used, and doses calculated based on the 

bioassay results are used in their place. Additionally, for non-digestive tract organs where one or 

more radionuclides of interest concentrate, the doses in Table 4.5.1-1 are not used; instead, the 

doses applied are either 1) those calculated to the specific organ based on the maximizing intake 

scenario, or 2) those calculated based on the bioassay results.

Table 4.5.1-1. Doses (rem) to non-metabolic and digestive tract organs.

Year Alpha Electrons

1 0.138 0.046

2 0.100 0.003

3 0.090 0.001

4 0.082 4.53E-04

5 0.076 3.28E-04

6 0.072 2.53E-04

7 0.069 2.00E-04

8 0.067 1.61E-04

9 0.065 1.32E-04

10 0.063 1.09E-04

11 0.063 9.18E-05

12 0.062 7.76E-05

13 0.062 6.62E-05

14 0.061 5.68E-05

15 0.061 4.92E-05

16 0.061 4.26E-05

17 0.061 3.72E-05

18 0.061 3.26E-05

Year Alpha Electrons

19 0.060 2.88E-05

20 0.060 2.54E-05

21 0.060 2.25E-05

22 0.060 2.01E-05

23 0.060 1.80E-05

24 0.060 1.62E-05

25 0.060 1.46E-05

26 0.060 1.32E-05

27 0.060 1.20E-05

28 0.060 1.09E-05

29 0.060 9.99E-06

30 0.059 9.17E-06

31 0.059 8.45E-06

32 0.059 7.77E-06

33 0.059 7.19E-06

34 0.059 6.66E-06

35 0.059 6.21E-06

36 0.059 5.76E-06

Year Alpha Electrons

37 0.058 5.37E-06

38 0.058 5.02E-06

39 0.058 4.71E-06

40 0.058 4.40E-06

41 0.058 4.13E-06

42 0.058 3.88E-06

43 0.057 3.66E-06

44 0.057 3.43E-06

45 0.057 3.24E-06

46 0.057 3.05E-06

47 0.057 2.89E-06

48 0.057 2.72E-06

49 0.056 2.57E-06

50 0.056 2.42E-06

51 0.056 2.30E-06

52 0.056 2.17E-06

53 0.056 2.05E-06
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For the purpose of calculating probability of causation, both the alpha and electron internal doses are 

assumed to be chronic, which is consistent with the nature of internal radiation exposures. The 

electron doses are assumed to be of energy E>15 keV. The doses are treated as a point estimate 

(constant).

4.5.2

As described in ORAUT-OTIB-0001, potential missed doses from tritium are assigned to employees 

as a maximizing assumption for likely non-compensable cases. The following doses (or the reported 

annual tritium doses, whichever are higher) are assigned to all organs.

Assignment of Internal Doses from Tritium: Maximizing Approach for Dose 

Reconstructions

Table 4.5.2-1. Tritium dose (rem).

Year Dose

1951 0.000

1952 0.000

1953 0.355

1954 0.355

1955 0.355

1956 0.355

1957 0.355

1958 0.355

1959 0.355

1960 0.355

1961 0.355

1962 0.355

1963 0.355

1964 0.355

Year Dose

1965 0.355

1966 0.355

1967 0.355

1968 0.355

1969 0.355

1970 0.355

1971 0.355

1972 0.355

1973 0.355

1974 0.355

1975 0.355

1976 0.355

1977 0.355

1978 0.355

Year Dose

1979 0.355

1980 0.355

1981 0.355

1982 0.355

1983 0.355

1984 0.071

1985 0.071

1986 0.071

1987 0.071

1988 0.071

1989 0.071

1990 0.071

1991 0.071

1992 0.007

Year Dose

1993 0.007

1994 0.007

1995 0.007

1996 0.007

1997 0.007

1998 0.007

1999 0.007

2000 0.007

2001 0.007

2002 0.007

2003 0.007

For the purpose of calculating probability of causation, doses from tritium are assumed to be chronic, 

and are assumed to be from electrons with energy E<15 keV. The doses are treated as a point 

estimate (constant).
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5.0

5.1 INTRODUCTION

OCCUPATIONAL EXTERNAL DOSIMETRY

SRS operations involved several processes of the nuclear weapons development cycle (DOE 1997) 

and represented a significant role within the US nuclear weapons program. These processes include 

nuclear fuel fabrication; nuclear reactor operations; radiochemical separations; radionuclide 

production; recycling uranium; refining, finishing and storing plutonium; and handling the associated 

radioactive waste. 

SRS workers, especially those employed during the peak production decades of the 1950s and 

1960s, have been exposed to radiation types and energies associated with the respective nuclear 

weapon development processes. SRS utilized facility and individual worker monitoring methods to 

measure and control radiation exposure to workers. There are numerous SRS records concerning 

facility monitoring, safety evaluations, investigations, etc. However, it is time consuming to locate and 

evaluate these records for SRS facilities and processes since 1951. Evaluations of the extensive 

scope of facility, process, and worker information relevant to an individual worker’s potential dose, 

many years or even decades after employment are difficult. Records of radiation dose to individual 

workers from personnel dosimeters worn by the worker and co-workers are available for SRS 

operations beginning in 1951. Dose from these dosimeters was recorded at the time of 

measurement, routinely reviewed by SRS operations and radiation safety staff for compliance with 

radiation control limits, and routinely available to individual workers. The NIOSH External Dosimetry 

Implementation Guide (NIOSH 2002a) has identified these records to represent the highest quality 

record to retrospectively assess dose. Information in this section pertains to analyzing these records 

and does not address parameters regarding skin and testicular and/or breast radiation dose that may 

result from acute beta (electron) radiation exposure in generally non-routine workplace exposure 

profiles. 

Radiation dosimetry practices were initially based on experience gained during several decades of 

radium and x-ray medical diagnostic and therapy applications. These methods were generally well 

advanced at the start of the Manhattan Engineering District (MED) program to develop nuclear 

weapons beginning about 1940. The primary difficulties encountered by MED operations to measure 

worker dose to external radiation involved:

• Large quantities of high level radioactivity

• Mixed radiation fields involving beta, photon (gamma and x-ray), and neutron radiation with 

low, intermediate, and high energies.

• Neutron radiation.

5.2 BASIS OF COMPARISON 

Historically, since the initiation of the MED in the early 1940s, various radiation dose concepts and 

quantities have been used to measure and record occupational dose. The basis of comparison for 

reconstruction of dose, as described in “External Dose Reconstruction Implementation Guideline” 

(NIOSH 2002a), is the Personal Dose Equivalent, Hp(d), where d identifies the depth (in mm) and 

represents the point of reference for dose in tissue. For weakly penetrating radiation of significance to 

skin dose, d = 0.07 mm and is noted as Hp(0.07). For penetrating radiation of significance to “whole 

body” dose, d = 10 mm and is noted as Hp(10). Both Hp(0.07) and Hp(10) are the radiation quantities 

recommended for use as the operational quantity to be recorded for radiological protection purposes 
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by the International Commission on Radiological Units and Measurements (ICRU 1993). In addition, 

Hp(0.07) and Hp(10) are the radiation quantities used in the DOE Laboratory Accreditation Program 

(DOELAP) used to accredit DOE personnel dosimetry systems since the 1980s (DOE 1986). The 

International Agency for Research on Cancer (IARC) Three Country Combined Study (Fix et al 1997) 

and IARC Collaborative Study (Thierry-Chef et al 2002) selected Hp(10) as the quantity to assess 

error in historical recorded “whole body” dose for workers in the IARC nuclear worker epidemiologic 

studies. The basis for comparison for neutron radiation is more complicated since historically the 

calibration of dosimeters to measure neutron dose was based on different dose quantities such as 

First Collision Dose, Multiple Collision Dose, Dose Equivalent Index, etc. The numerical difference in 

using these dose quantities compared to the Hp(10) dose used in current DOELAP performance 

testing could be evaluated using the relative values of the dose conversion factors for the respective 

dose quantities in conjunction with characteristics of the respective SRS neutron dosimeter response 

characteristics and workplace radiation fields. 

5.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of the beta, photon (x-ray, gamma ray), and neutron radiation types energies and 

geometries of exposure, and the characteristics of the respective SRS dosimeter response are crucial 

to the assessment of bias and uncertainty of the original recorded dose in relation to the radiation 

quantity Hp(10). The bias and uncertainty for current dosimetry systems is typically well documented 

for Hp(10) (SRS 1993). Often the performance of current dosimeters can be compared with 

performance characteristics of historical dosimetry systems in the same, or highly similar, facilities or 

workplaces. In addition, current performance testing techniques can be applied to earlier dosimetry 

systems to achieve a consistent evaluation of all dosimetry systems. Dosimeter response 

characteristics for radiation types and energies in the workplace are crucial to the overall analysis of 

error in recorded dose.

Overall, accuracy and precision of the original recorded individual worker doses and their 

comparability to be considered in using NIOSH (2002) guidelines depend on (Fix et al 1997):

• Administrative practices adopted by facilities to calculate and record personnel dose based 

on technical, administrative, and statutory compliance considerations.

• Dosimetry technology includes the physical capabilities of the dosimetry system, such as the 

response to different types and energies of radiation, in particular in mixed radiation fields.

• Calibration of the respective monitoring systems and similarity of the methods of calibration to 

sources of exposure in the workplace.

• Workplace radiation fields that may include mixed types of radiation, variations in exposure 

geometries, environmental conditions.

An evaluation of the original recorded doses based on these parameters is expected to provide the 

best estimate of Hp(0.07), as necessary, and Hp(10) for individual workers with the least relative 

overall uncertainty. 

5.3.1

Historically, SRS had an extensive radiation safety monitoring program to measure radiation exposure 

in the workplace using portable radiation instruments, contamination surveys, zone controls, and 

personnel dosimeters. The results from the personnel dosimeters were used to measure and record 

SRS Historical Administrative Practices
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dose from external radiation exposure to SRS workers throughout the history of SRS operations 

(Taylor et al 1995). These dosimeters include one or more of the following:

• Personnel Whole Body (WB) beta/photon dosimeters 

• Pocket Ionization Chamber (PIC) dosimeters 

• Personnel extremity dosimeters

• Personnel neutron dosimeters. 

SRS began operations in 1951 using dosimeter and processing technical support provided by the Oak 

Ridge National Laboratory (ORNL). A summary of SRS personnel beta/photon and neutron 

(respectively, dosimeter type, exchange, Minimum Detection Level (MDL) and potential missed 

annual dose) is shown in Table 5.3.1-1. ORNL, then the Clinton Laboratory, had implemented their 

dosimetry methods based on the personnel beta/photon dosimeter design developed at the 

Metallurgical Laboratory at the University of Chicago (Pardue et al 1944). ORNL provided SRS with 

beta/photon film dosimeters and neutron nuclear track, type A (NTA) emulsion dosimeters at the 

beginning of SRS operations until SRS implemented their in-house personnel dosimetry capabilities 

about one year later. The precise detection levels shown in Table 5.3.1-1 are difficult to estimate, 

particularly for the older systems. The current SRS commercial TLD system minimum detection 

levels (MDLs) are identified in SRS external dosimetry technical basis documentation (SRS 1993) 

based on a DOE Laboratory Accreditation Program (DOELAP) laboratory testing protocol (DOE 

1986). During earlier years, the MDLs are subject to additional uncertainty since factors involving the 

radiation field and film type, as well as the processing, developing and reading system cannot be 

tested. The film dosimeter MDLs shown in Table 5.3.1-1 were estimated based on information from 

NIOSH (1993), NRC (1989), and Wilson et al (1990). The TLD MDLs were obtained from Taylor et al 

(1995) and SRS (1993).

Dose reconstruction parameters concerning SRS administrative practices significant to dose 

reconstruction involve issues as follows:

• Policies to assign dosimeters to workers

• Policies to exchange dosimeters

• Policies to record notional dose (i.e., some identified value for lower dosed workers often 

based on a small fraction of the regulatory limit) 

• Policies to estimate dose for missing or damaged dosimeters

• Policies to replace any destroyed or missing records

• Policies to evaluate and record dose for incidents

• Policies to obtain and record occupational dose to workers for other employer exposure.

SRS policies appear to have been in place for all of these parameters. SRS routine practices appear 

to have required assigning dosimeters to all workers who entered a controlled radiation area. 

Dosimeters were exchanged on a routine schedule. All dosimeters were processed and the 

measured results recorded and used to estimate dose. There appears to be no use of recorded 

notional doses, although there are issues of recording dose for low-dose dosimeters (See section on 

“missed dose”). Figure 5.3.1-1 illustrates the number of SRS monitored workers and the collective 
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Table 5.3.1-1. SRS dosimeter type, period of use, exchange frequency, MDL and potential annual 

missed dose.

Dosimeter Period of use

Exchange 

frequency

Laboratory 

MDL 

(mSv)a

Max. annual 

missed dose

(mSv)b

Beta/photon dosimeters

ORNL two-element film 

dosimeter

Prior to 3/52 Weekly (n=52) 0.4 20.8

SRS two-element film 

dosimeter

After 3/52 through 9/30/1957 Weekly (n=52) 0.4 20.8

Beginning 10/1/1957 through 

11/8/1959

Biweekly (n=26) 0.4 10.4 

SRS multi-element film 

dosimeter

Beginning 11/9/1959 through 

12/31/1964

Biweekly (n=26) 0.4 10.4

Beginning 1/1/1965

through 12/31/1965

4-week (n=13) 0.4 5.2

Beginning 1/1/1966 

through 3/31/1970

Monthly (n=12) 0.4 4.8

SRS TLD Beginning 4/1/1970 through 

6/30/1983

Monthly (n=12) 0.15 1.8

Panasonic TLD Beginning 7/1/1983 to 2003 

(ongoing)

Monthly (n=12) 0.05 0.6

Neutron dosimeters

ORNL NTA Prior to 8/3/1953 Weekly (n = 52) ~0.5 13c

SRS NTA Beginning 8/3/1953 through 

7/13/1960

Weekly (n = 52) ~0.5 13c

SRS NTA Beginning 7/14/1960 through 

12/31/1970

Biweekly (n=26) ~0.5 6.5c

SRS Hoy TLND Beginning 1/1/1971 through 

12/31/1994

Monthly (n=12) 0.2 1.2

Panasonic TLND Beginning 1/1/1995 to 2003 

(ongoing)

Monthly (n=12) 0.15 0.9

a. Estimated film dosimeter detection levels based on NIOSH (1993), NRC (1989), and Wilson et al (1990). TLD detection 

levels from Taylor et al (1995) and SRS (1993). 

b. Maximum annual missed dose (NIOSH 2002)

c. The potential annual missed dose based on laboratory irradiations is not applicable to workplace missed neutron dose. 

“whole body” dose of record for SRS workers since 1951 (Taylor et al 1995). Administrative practices 

are generally available in the SRS procedures manual (DuPont 1954, 1957, 1961a, 1961b), and 

detailed information for each worker in the respective SRS exposure history file documentation. 

Based on the SRS documentation and the reasonable trend in these figures, there does not appear to 

be any significant administrative practice that would affect the integrity of the recorded dose of record 

for SRS workers with the exception that SRS hard-copy radiation records must be consulted to 

separate the photon, neutron and tritium dose for workers who terminated prior to 1979. Apparently, 

only the total whole body dose (i.e., photon + neutron + tritium) is available from the second quarter of 

1963 through December 31, 1972. The missing dose components of the total whole body dose could 

be estimated from prior year and subsequent year dose results according to the methods described 

by Watson et al (1994) based on examination of continuity in the worker’s job and work activities. 

5.3.2

SRS dosimetry methods evolved during the years as improved technology was developed and the 

complex radiation fields were better understood. The adequacy of the respective dosimetry methods 

to accurately measure radiation dose is determined from the radiation type, energy, exposure 

geometry, etc., as described in later sections. The dosimeter exchange frequency was gradually 

SRS Dosimetry Technology
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0

5000

10000

15000

20000

25000

1940 1950 1960 1970 1980 1990 2000

Year

Normalized Values

Monitored Workers

Whole Body Dose

Figure 5.3.1-1. SRS monitored workers and collective dose.

lengthened, generally corresponding to the time period of the regulatory dose controls. At the 

beginning of SRS operations, a weekly dose control of 3 mSv was in effect. This was changed to an 

annual limit of 50 mSv in the latter 1950s. Table 5.3.2-1 is a summary of the major historical events in 

the SRS personnel dosimetry program. 

Operations at SRS occurred about 7 to 8 years after the MED nuclear weapons facilities began 

operations at Oak Ridge National Laboratory (ORNL), Los Alamos National Laboratory (LANL), and 

Hanford. These AEC sites implemented, in the beginning, essentially the same beta/photon and 

neutron dosimeter methods. There were intercomparisons of dose interpretations among these AEC 

sites through the years as reported in Wilson et al (1990). 

5.3.2.1 Beta/Photon Dosimeters

In the early 1950s, SRS used personnel dosimeter services from ORNL (Taylor et al 1995). ORNL 

had implemented in the mid-1940s the beta/photon film dosimeter design, with perhaps minor 

differences, that was developed originally at the Metallurgical Laboratory at the University of Chicago 

(Pardue et al 1944). Within about one year, SRS implemented their own in-house dosimetry 

capabilities. SRS followed a process of dosimetry research and development leading to gradual 

upgrades in dosimetry capabilities for the complex radiation fields encountered in SRS reactor, 

reprocessing and plutonium handling facilities (Taylor et al 1995). This evolution in dosimetry 

capabilities was followed by other DOE sites leading to site-specific multi-element film and 

thermoluminescent dosimetry systems. SRS dosimeters always included a comparatively “open 

window” to measure significant beta radiation and to distinguish the dosimeter response between beta 

and photon radiation. The photon energy response characteristics of the SRS beta/photon 

dosimeters are illustrated in Figure 5.3.2.1-1 based on the essentially identical two-element film 

dosimeter used at SRS, Hanford and ORNL (and likely other MED sites) designed at the University of 

Chicago (Pardue et al 1944). Figure 5.3.2.1-1 also illustrates the Hp(10) response. The improved 

energy response of multi-element film and thermoluminescent dosimeters is shown in Figure 5.3.2.1-1 

using available Hanford published information (Wilson et al 1990). The general trend in the energy 

response curves for these dosimeters, relative to Hp(10), provides insight into the expected 
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Table 5.3.2-1. SRS historical dosimetry events (Taylor et al 1995).

Date Description

1951 Pocket Ionization Chambers (PICs) were used to measure SRS worker exposure. PICs were 

used in addition to film dosimeters.

11/12/1951 ORNL two-element (i.e., open window and 1 mm silver filter) beta/photon dosimeters issued to 

SRS workers. Processing was provided by ORNL. Routine dosimeter exchange period was 

weekly.

1951 Nuclear Track Emulsion (NTA) dosimeters were issued to selected SRS workers to measure 

neutron radiation. Capability was provided by ORNL. NTA film was placed, facing the worker’s 

body, in two-element beta/photon dosimeter holders. NTA film exchange was also weekly.

3/3/1952 SRS began onsite dosimetry program with the same two-element film dosimeter design as 

ORNL. 

3/1953 SRS began processing film using the ORNL film badge dosimeter. The results were reported in 

mrem/week, manually recorded on individual worker file folders. 

1953 Extremity film dosimeter use and processing began at SRS.

8/3/1953 SRS assumed processing of NTA films.

4/1954 Indium foil placed in dosimeter holder to respond to neutron radiation.

10/1/57 SRS beta/photon dosimeter exchange changed to biweekly.

11/9/1959 SRS designed multi-element film dosimeter use was implemented. This dosimeter design 

permitted analysis of beta, gamma and x-ray exposure to personnel. 

7/14/1960 NTA exchange changed to biweekly.

1963 PICs were replaced with self-reading dosimeters to provide a real-time indication of exposure.

1965 Four week beta/photon dosimeter exchange implemented.

1966 Monthly beta/photon dosimeter exchange implemented. 

4/1/1970 SRS TLD replaced film as the principal means of measuring beta/photon dose. The minimum 

recorded dose was about 0.15 mSv.

1/1/1971 SRS Hoy Thermoluminescent Neutron Dosimeter (TLND) implemented.

1980 Extremity dose routinely measured with TLD chip.

7/1/1982 Commercial Panasonic beta/photon TLDs implemented. 

mid-1980s SRS implemented changes such as on-phantom calibration and 137Cs calibration source to 

participate in new DOE Laboratory Accreditation Program (DOELAP) implemented to 

performance testing. The overall change in recorded dose with new practices, as described in 

Taylor et al (1995), was determined to be an equivalent 11.9% increase compared to earlier 

dose estimates.

1989 SRS TLD is DOELAP accredited. Lower limit of detection, based on DOELAP protocol, was 0.2 

and 0.05 mSv, respectively, for the shallow and deep dose components. 

1/1/1995 Panasonic neutron dosimeter replaced SRS TLND. Routine dosimeter exchange for the 

Panasonic beta/photon dosimeter is quarterly and for the Panasonic TLND, monthly.

performance of these dosimeters in SRS workplace radiation fields. The original two-element 

dosimeter shows a significant under-response to photon energies less-than about 100 keV. 

Evaluations of the two-element film dosimeter that was used at SRS, ORNL, LANL, and Hanford sites 

have been performed by the International Agency for Research on Cancer (IARC) Collaborative Study 

researchers (Thierry-Chef et al 2002) and by Fix et al (1994) to several photon energies. IARC 

conducted a dosimeter intercomparison study to higher-energy (i.e., >100 keV) photons of ten 

commonly used historical dosimetry systems throughout the world. Three of the dosimeter designs 

were from the US. These included the two-element dosimeter used at Hanford (identified as US-2), 

multi-element film dosimeter used at Hanford (identified as US-8), and the currently used SRS 

Panasonic dosimeter system (identified as US-22). The IARC Study considered that exposure to 

dosimeters worn by workers could be characterized as a combination of anterior-posterior (A-P), 

rotational, and isotropic irradiation geometries. Dosimeter response to selected photon energies was 

measured using two phantoms. These phantoms were used to simulate the effect of the workers 

body on the measured dosimeter response. The first phantom is the International Standards 
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Figure 5.3.2.1-1. Estimated SRS dosimeter photon response 

characteristics.

Organization water filled slab phantom with polymethyl methacrylate walls measuring 30 cm wide by 

30 cm tall and 15 cm in depth that is widely used for dosimeter calibration and performance testing. 

The second phantom used was an anthropomorphic Alderson Rando Phantom. This phantom is 

constructed from a natural human skeleton cast inside material that has a tissue equivalent response. 

The results of this testing, for US dosimeters only, are presented in Table 5.3.2.1-1. As noted in 

Table 5.3.2.1-1, the two-element dosimeter design significantly overestimated Hp(10) for all irradiation 

geometries and increasingly for the lower energy photon energy at 118 keV. Similar anthropomorphic 

phantom laboratory irradiations using anterior-posterior and rotational geometries to several photon 

energies have been done at Hanford (Wilson et al 1990) with similar results. 

Table 5.3.2.1-1. IARC testing results for US beta/photon dosimeters.

Geometry Phantom

118 keV 208 keV 662 keV

Meana SD/ Mean Meana SD/ Mean Meana SD/ Mean

US-2 (Two-element film dosimeter)

A-P Slab 3.0 2.1 1.3 1 1.0 0.8

A-P Anthropomorphic 3.0 4.2 1.2 1.9 1.0 1.8

Rotational Anthropomorphic 2.2 2 1.4 3 1.2 3.2

Isotropic Anthropomorphic 1.5 4.4 1.1 1.6 1.0 2.7

US-8 (Multi-element film dosimeter)

A-P Slab 1.0 1.5 1.0 0.8 0.8 1.7

A-P Anthropomorphic 0.8 9.5 0.9 6 0.8 1.8

Rotational Anthropomorphic 1.2 1.9 1.2 17 1.1 1.8

Isotropic Anthropomorphic 1.0 3 1.2 9 1.0 2.3

US-22 (Multi-element thermoluminescent dosimeter)

A-P Slab 0.9 4.4 0.9 3.9 0.9 3.5

A-P Anthropomorphic 0.8 3.1 0.9 2.1 0.9 3.9

Rotational Anthropomorphic 1.1 3.1 1.2 1.5 1.0 4.1

Isotropic Anthropomorphic 0.9 0.3 1.0 2.5 0.9 1.6

a. Ratio of recorded dose to Hp(10).

To evaluate the dosimeter response to lower-energy (i.e., < 100 keV) photons that are significant in 

plutonium facilities, Hanford conducted intercomparison testing of all Hanford historical dosimeter film 

designs (Wilson et al 1990) using A-P irradiations only. Although there are differences in films and 
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particularly the filters used in the multi-element dosimeters, good comparison in energy response for 

SRS and Hanford dosimeters is expected. The results of this testing is summarized in Table 5.3.2.1-

2. The dosimeter results for energies greater than 100 keV, are consistent with the IARC results 

showing an over-estimate of Hp(10) for the two-element dosimeter. For energies less than 100 keV, 

the two-element dosimeter underestimates the photon dose. This under-response is also illustrated in 

the original University of Chicago two-element dosimeter energy response curve (Pardue et al 1944). 

Table 5.3.2.1-2. Testing results for Hanford two-element and multi-element film dosimeters for energy 

and angular response.

a,b

Beam

(energy, 

keV)

Anterior-posterior (AP) exposure Rotational exposure

Film dosimeters TLD,

1972-

present

Film dosimeters TLD,

1972-

93

Two-element

1944-56

Multi-element,

1957-71

Two-element 

1944-56

Multi-element,

1957-71

16b 0.1 0.9

59b 0.5 1.1

M150(70) 0.7 0.70 0.95 1.31 1.31 1.77

H150(120) 1.6 0.64 0.87 3.00 1.20 1.64

137Cs(662) 1.0 1.0 1.0 1.46 1.46 1.46

a. Divide recorded dose by table value to estimate Hp(10). 

b. Based on Wilson et al (1990). 

SRS adopted a conservative practice historically to calculate the deep dose for selected workers in 

plutonium facilities in specific instances when gonadal dose was considered significant. This practice 

involved adding the open window (i.e., skin dose), based on a 16 keV k-fluorescent calibration, to the 

deep dose as measured from the film response behind the 1 mm silver filter. Figure 5.2.3.1-1 

illustrates the potential significance of this correction to the routinely recorded dose. Based on this 

practice, routine SRS deep dose estimates are judged to be reasonably accurate compared to Hp(10) 

for all photon energies. SRS used uranium to routinely calibrate the dosimeter open window dose 

response.

5.3.2.2 Neutron Dosimeters 

Two general types of SRS neutron dosimeters have been used and these differ dramatically in their 

response to neutron radiation. 

NTA. The nuclear track emulsion, type A, (NTA) neutron dosimeter consisted of the NTA film 

enclosed in the same holder as used for SRS beta/photon dosimeters until November 1959 when a 

NTA specific dosimeter holder was implemented (Taylor et al 1995). 

TLD. The SRS TLD neutron (TLND) dosimeter was implemented on 1 January 1971. The TLND is a 

sphere design, using two polyethylene spheres covered on the curved surface with cadmium. The 

dosimeter is worn using a belt to minimize distance from the worker’s body (Hoy 1972). SRS 

implemented a commercial Panasonic neutron TLD system on January 1, 1995.

The AEC held a series of Personnel Neutron Dosimetry Workshops to address problems being 

experienced by the respective AEC sites concerning accurate measurement of neutron dose. The 

first workshop was held 23–24 September 1969 (Vallario et al 1969) with the stated concern: “...for 

intermediate energy ......neutron sources, NTA personnel neutron dosimeters cannot be effectively 

used. This leaves a gap in the personnel dosimetry program which at many installations may be quite 

serious.” The workshops were generally limited to representatives from sites with active personnel 

neutron dosimetry programs and continued for many years. The 11th DOE Workshop on Personnel 

Neutron Dosimetry was held in 1991 (Rabovsky et al 1991). The significance of the underestimated 
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neutron dose became evident with the studies leading to implementation of the thermoluminescent 

dosimeters (TLDs) and using simultaneous NTA and TLD (neutron) measurements of recorded 

worker dose. Hanford (Nichols et al 1972) conducted field studies with workers wearing both types of 

dosimeters. 

Response characteristics of the SRS NTA type dosimeter and the TLND neutron dosimeter in 

comparison with other DOE neutron dosimeters are presented in Brackenbush et al (1980). In 

general, the response of the NTA dosimeter decreases with decreasing neutron energy that are 

greater than a threshold energy estimated by SRS to be about 500 keV (Taylor et al 1995) and the 

TLD response increases with decreasing neutron energy. The response of both dosimeters is highly 

dependent upon the neutron energy spectra. Both dosimeter types require matching the laboratory 

calibration neutron spectra with the workplace spectra for accurate results. Results reported by Hoy 

during the first AEC Neutron Dosimetry Workshop in 1969 indicated that simultaneous laboratory2

5.3.3

measurements of the dose equivalent with NTA was about one-half to one-fourth that compared with 

other methods including the TLND (Vallario et al 1969). 

Potential error in recorded dose is dependent on the dosimetry technology response characteristics to 

each radiation type, energy, and geometry; the methodology used to calibrate the dosimetry system; 

and the extent of similarity between the radiation fields used for calibration and that present in the 

workplace. The potential error is much greater for dosimeters with significant variations in response 

such as film dosimeters for low energy photon radiation and both the NTA and TLND for neutron 

radiation. 

Calibration

5.3.3.1 SRS Beta/Photon Dosimeters

SRS dosimeters were originally calibrated using primarily uranium and 226Ra using in-air (i.e., no 

phantom) exposures to selected levels. K-fluorescent x-rays were used to develop dosimeter 

response characteristics for the lower energy photon fields in plutonium facilities. This practice is 

similar to other AEC sites. Taylor et al (1995) describes adjustments to SRS recorded dose to

estimate Hp(10) based on SRS preparations for DOELAP performance testing in the mid-1980s. At 

that time, it was concluded that:

• Prior to 1 January 1986 the recorded dose of record (i.e., photon) dose should be multiplied by 

a factor of 1.119 (11.9%). 

• Prior 1 January 1987 recorded dose of record (i.e., photon) should be multiplied by a factor of 

1.039 (3.9%).

These changes in the recorded dose should be made to arrive at an assured claimant-favorable 

treatment. Common sources of laboratory bias are shown in Table 5.3.3.1-1 for personnel 

beta/photon dosimeter calibration based on comparison of the recorded dose with Hp(10). 

5.3.3.2 SRS Neutron Dosimeters

Historical aspects of the calibration of SRS NTA and TLNDs are described by Taylor et al (1995) and 

in articles by Hoy that are included in several reports during Atomic Energy Commission workshops 

on personal neutron dosimetry such as the first during 23-24 September 1969 (Vallario et al 1969) 

 2. Based on June 19, 2003 teleconference between NIOSH, NIOSH project contractors and SRS technical staff.
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Table 5.3.3.2-1. Common sources of laboratory bias for neutron dosimeter calibration parameters.a

Parameter Historical description Anticipated laboratory biasb

Source energy 

spectraIn the 1980s, SRS began exposing calibration dosimeters on phantom (used to simulate 

worker’s body). As such, previous 

calibrations do not include response from 

radiation backscatter response. The beam 

was also degraded in an attempt to better 

represent the workplace.

The delivered dose was uncertain as noted in 

Brackenbush et al (1987). (see workplace 

radiation fields)

Radiation quantity Neutron dose quantities used to calibrate 

SRS neutron dosimeter systems have varied 

historically. For many years the neutron dose 

equivalent index was used.

The effects of the respective dose quantities 

used to calibrate SRS neutron dosimeters is 

uncertain and could be evaluated in comparison 

to the Hp(10) dose used in DOELAP 

performance testing.

Angular response SRS dosimeters calibrated using anteriorposterior (A-P) laboratory irradiations.

Recorded dose of record likely too low since 

dosimeter response is lower at angles other than 

A-P. Effect is highly dependent upon energy. 

Environmental 

stability

SRS NTA film dosimeter and TLD systems 

are subject to signal fade with time, heat, 

humidity, light, etc.

Recorded dose of record likely too low because 

of fading however this effect depends strongly 

upon when calibration dosimeters are irradiated. 

a. Judgment based on SRS dosimeter response characteristics.

b. Recorded dose compared to Hp(10).

and second during 8-9 July 1971 (Vallario et al 1971). Table 5.3.3.2-1 describes the common sources 

of laboratory bias for personnel neutron dosimeter calibration based on expected comparison of the 

recorded dose with Hp(10). 

Table 5.3.3.1-1. Common sources of laboratory bias for beta/photon dosimeter calibration 

parameters.

a

Parameter Historical description Anticipated laboratory biasb

In-air calibration In the 1980s, SRS began exposing calibration 

dosimeters on phantom (used to simulate 

worker’s body). As such, previous 

calibrations do not include response from 

radiation backscatter response.

Recorded dose of record too high. Backscatter 

radiation from worker’s body is highly dependent 

upon dosimeter design.

Radiation 

quantity

Prior to the 1980s, SRS dosimeter systems 

were typically calibrated to photon beam 

measured as exposure.

For higher energy 226Ra and 137Cs gamma 

radiation used to calibrate dosimeters this 

caused only a slight (about 3%) underresponse in recorded dose. 

Tissue depth of 

dose

Historically, SRS used selected depths of 1 or 

2 cm to estimate the deep dose. 

The numerical effect of this for photon radiation 

is comparatively low. SRS dosimeter designs 

had filtration density thickness of about 1,000 

mg/cm2 that would relate to the 1 cm depth in 

tissue. 

Angular response SRS dosimeter system is calibrated using 

anterior-posterior (A-P) laboratory irradiations.

Recorded dose of record likely too low.

Dosimeter response is lower at non A-P angles. 

Effect is highly dependent upon radiation type 

and energy. 

Environmental 

stability

SRS film dosimeter and TLD systems are 

subject to signal fade with time, heat, 

humidity, light, etc.

Recorded dose of record likely too low however 

this depends strongly upon when calibration 

dosimeters are irradiated. Mid-cycle calibration 

minimizes effects. 

a. Judgment based on SRS dosimeter response characteristics.

b. Recorded dose compared to Hp(10).
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5.3.4

SRS operations are characterized by significant complex beta, photon and neutron radiation fields in 

SRS reactor, irradiated fuel processing, plutonium handling, neutron source production, and 

radioactive waste facilities. Descriptions of processes and primary nuclides are presented in the 

respective SRS External Dosimetry Technical Basis documents (WSRC 2003).

Workplace Radiation Fields

5.3.4.1 SRS Workplace Beta/Photon Dosimeter Response

Essentially all SRS radiological work areas involved beta/photon radiation covering a wide range of 

energies characteristic of the radionuclides being handled within the respective facilities and 

processes. Radiation beta/photon fields characteristic of SRS facilities can be generally defined 

based on historical information as presented in Table 5.3.4.1-1. The expected performance of SRS 

dosimeters to these fields has been estimated in this TBD using the extensive evaluations of the very 

similar Hanford dosimeters (Wilson et al 1990, Fix et al 1994, 1997), the IARC intercomparison study 

results (Thierry-chef et al 2002) and available internal3

Common beta/photon personnel dosimeter parameters important to Hp(10) performance in the 

workplace are summarized in Table 5.3.4.1-2. Based on the collective information, SRS dosimeters 

are expected to reasonably measure the Hp(10) dose under all SRS workplace radiation fields. The 

SRS historical practice, for the two-element film dosimeter in plutonium facilities characterized by 

predominant photon energies <100 keV, to calculate the total whole body deep dose by summing the 

shallow dose from the open window film response based on a 16 keV fluorescent x-ray calibration and 

the deep dose from the 1 mm silver filtered film response based on a 226Ra calibration, will result in a 

over-estimate of the actual Hp(10) dose. In addition, laboratory irradiations of the two-element 

dosimeter have shown an over-response of the actual Hp(10) dose by about a factor of 2 to photons 

greater than 100 keV. A claimant-favorable approach is proposed to ignore this over-response 

because of the complexity of workplace photon energies and exposure geometries that tend to result 

in an under-estimate of the Hp(10) dose. The respective SRS dosimeters have filtration of 

approximately 1000 mg/cm2 (i.e., nearly equivalent to 1 cm depth in tissue) for those regions of the 

dosimeter used to measure the whole body dose. The response to beta radiation in SRS workplaces 

is limited because beta radiation cannot penetrate the filtration. As such, a reasonable estimate of 

deep dose, compared to Hp(10), is expected for SRS beta/photon workplace radiation.

SRS documentation.

5.3.4.2 SRS Workplace Neutron Dosimeter Response

SRS neutron dosimetry methods are similar to those used at other early MED sites. Table 5.3.4.2-1 

presents a summary of common workplace neutron dosimeter performance characteristics that may 

be important to SRS workplace measured dose. There are four main areas at SRS where there is a 

potential for neutron exposure: the plutonium facilities in the 200 area; the Calibration Facility (736-A) 

and the Cf-252 facility (773-A) in 700 area; the reactor areas (100) during operation; and the 321 

Building (Pu-Al alloys) in the 300 Area. Each of these facilities has different neutron exposure 

spectrum. These areas also have demonstrated different neutron to photon ratios. The following 

sections discuss the different energy spectra and neutron to photon ratios by facility. 

5.3.4.2.1 Plutonium Facilities

The majority of the neutron exposure at the Savannah River Site (SRS) is associated with the 

separation, finishing, and storage of plutonium. These operations are performed primarily on the 

 3. For example, letter, dated September 3, 1968 from W.R. Tyson, to J.D. Ellett entitled, “Personnel Radiation Exposures –

Building 235-P.”
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Table 5.3.4.1-1. Selection of beta and photon radiation energies and percentages for SRS facilities. Process/

buildings Description

Operations Radiation 

type

Energy 

Begin End selection Percent

Fuel fabrication Produced reactor fuel and target assemblies. 1953 1989 Beta > 15 keV 100%

300M Uranium fuel fabrication 1953 1989 Photon 30 – 250 keV 100%

Reactors During Operation: Highly dispersed fields of higher-energy photon radiation 

fields from fission process, activation/ fission product nuclides.. Potential for 

significant airborne nuclides and there may be significant higher-energy beta 

radiation. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50%

While not in Operation: Highly dispersed fields of higher-energy photon 

radiation fields from activation/fission product nuclides.. There may be 

significant higher-energy beta radiation during maintenance work resulting from 

fission products.

105C Heavy water moderated reactors 3/1955 6/1985

105K 10/1954 1988

1992 7/1992

105L 7/1954 1968

1985 6/1988

105P 2/1954 8/1988

105R 12/1953 6/1964

Processing plants Radiochemical Operations: Highly dispersed fields of higher-energy photon 

radiation fields from activation/fission product nuclides dominant to most 

exposure profiles. Potential for higher-energy beta radiation during sampling 

and, maintenance work resulting from fission products. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50% 200F Radiochemical processing plant 1954

200H Radiochemical processing plant 1955

utonium 

production

Plutonium Component Production: Plutonium is machined into weapon 

components using a glove box assembly process with predominant close 

anterior exposure to workers. Radiation characteristics in this area involve 

significant lower-energy photons and lower-energy neutron radiation.

Photon < 30 keV

30 – 250 keV

25%

75%

Plutonium Storage: Radiation characteristics in this area generally involve 

dispersed lower-energy neutron radiation and scattered photons to include 60 

keV Am-241 gamma ray.

221F, B-line Plutonium Finishing Process 1954

221H, B-line Plutonium Finishing Process 1955

772F Production Control Laboratory 1955

235F Plutonium Fuel Facility 1955

736A Calibration Facility 1953

adionuclide

production

Radiation characteristics are highly dependent upon the radionuclide being 

produced. SRS has produced several radionuclides including tritium, 238Pu, 239Pu, americium, curium, californium, etc. 

(Specific to 

radionuclide)

Photon, 

default

30 – 250 keV

> 250 keV

50%

50%

773A Savannah River Laboratory 1953

Waste handling Radiation characteristics are highly dependent upon the source of waste but 

typically fission product nuclides (90Sr/90Y, 137Cs) are dominant. Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50% 200F General waste handling, storage tanks, etc. 1953 Ongoing

200H General waste handling, storage tanks, etc 1953 Ongoing

200F B-line (221-F), the 200H HB-line (221-H), the Production Control Laboratory (772-F), and at the 

PuFF Facility (235-F). Neutron dose is also associated with process control analysis and storage of 

the product. The plutonium is introduced into the SRS plutonium finishing processes as plutonium 

fluoride. The plutonium fluoride is converted into plutonium oxide and made into buttons. Plutonium 

fluoride produces an (α,n) reaction resulting in a significant source of neutrons. Neutrons are also 

emitted spontaneously from 238Pu and 240Pu, which compose a large activity fraction of the plutonium 

buttons. Plutonium-238 is the primary isotope used for the production of heat sources (Bebbington, 

1990). 

5.3.4.2.1.1 Neutron Energy Spectra

Brackenbush et al (1987) measured neutron spectra at several locations across the Savannah River 

Site. Figure 5.3.4.2-1 provides the neutron spectra measured at glove boxes on the FB and HB lines 

respectively. It should be noted that the data in Brackenbush et al. (1987) were depicted as dose 
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Table 5.3.4.1-2. Common workplace beta/photon dosimeter Hp(10) performance.

a

Parameter Description Workplace biasb

Exposure 

geometry

SRS dosimeter system 

calibrated using anteriorposterior (A-P) laboratory 

irradiations.

Recorded dose of record likely too low since dosimeter 

response is lower at angles other than A-P. Effect is highly 

dependent upon radiation type and energy. 

Energy 

response

SRS film “deep dose” 

dosimeter response to 

photon radiation <100keV is 

too low and >100keV is too 

high.

Bias in recorded dose depends upon the photon spectra 

in the workplace. SRS practice to include the shallow 

dose based on a 16-keV calibration to the deep dose for 

selected workers in SRS plutonium facilities where gonadal 

dose was significant eliminated this common source of 

potential under-response. Reasonable estimate of Hp(10) 

dose is expected.

Mixed fields SRS dosimeters respond to 

beta and photon radiation.

Filtration of about 1,000 mg/cm2 over dosimeter component 

used to calculate deep dose will minimize dosimeter 

response to beta radiation. Reasonable estimate of Hp(10) 

dose is expected.

Missed dose Doses less than Minimum 

Detection Level recorded as 

zero dose.

Recorded dose of record likely too low and the relative 

impact is significant primarily in earlier years with frequent 

dosimeter exchange and film dosimeters with higher MDLs. 

Environmental 

effects

Workplace heat, humidity, 

etc., fades dosimeter signal. 

Recorded dose of record likely too low.

a. Judgment based on SRS dosimeter response characteristics and workplace radiation fields.

b. Recorded dose compared to Hp(10).

Table 5.3.4.2-1. Common workplace neutron dosimeter Hp(10) performance.a

Parameter Description Workplace biasb

Workplace 

neutron energy 

spectra

SRS NTA dosimeter response 

decreases and SRS TLD response 

increases with decreasing neutron 

energy.

Depends upon workplace neutron spectra. NTA 

recorded dose of record likely too low. Article 

by Hoy presented in Vallario et al (1969) 

identified ratios of 2 to 4 under-response. 

Exposure 

geometry

NTA Dosimeter response increases 

with increasing exposure angle and 

TLD response decreases with 

increasing exposure angle. 

NTA recorded dose likely too high since 

dosimeter response is higher angles other than 

A-P. TLD recorded dose of record likely too 

low since dosimeter response is lower at angles 

other than A-P. Effect is highly dependent upon 

neutron energy. 

Dosimeter 

placement

SRS apparently instructed workers to 

wear NTA dosimeter facing the 

workers body to enhance dosimeter 

response (i.e., to respond to 

backscatter neutron radiation).

Recorded dose of record likely too low. 

Reflected neutrons are expected to be less-than 

the NTA energy threshold identified as 0.5 MeV 

by SRS. 

Mixed low energy 

photon and 

neutron radiation

SRS NTA dosimeter responds 

significantly to lower-energy photons 

that can result in fogging of the NTA 

film.

NTA recorded dose of record likely too low. 

SRS NTA dosimeters selected to minimize 

photon response. SRS TLDs have capability to 

remove photon dose contribution. 

a. Judgment based on SRS dosimeter response characteristics and workplace radiation fields.

b. Recorded dose compared to Hp(10).

equivalent rates, however for simplicity of calculation, a one hour exposure was assumed in order to 

use dose equivalent. 

Although somewhat different neutron spectra were recorded for the different exposure areas (HB 

Line, FB Line, and storage facility), similarities were observed in the general shape of the spectra and 

the energy of the flux peak at approximately 1-2 MeV. The HB glovebox spectra in Figure 5.3.4.2-1 
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Neutron Energy (MeV)

10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101 102

Dose Equivalent (mrem) 10-3

10-2

10-1

100

101

102

0.355 mrem 0.168 mrem

36.0 mrem

0.557 mrem

FB Line Glovebox

Neutron Energy (MeV)

10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101 102

Dose Equivalent (mrem) 10-4

10-3

10-2

10-1

100

101

0.304 mrem 0.118 mrem

2.60 mrem 2.83 mrem

HB Line Glove Box 

Figure 5.3.4.2-1. Neutron energy spectra recorded at SRS plutonium facilities with neutron energy 

groups overlay.

demonstrated a higher peak energy which is possibly the result of more spontaneous fission from Pu238 but other spectra from the HB line indicated a spectra with neutron energies between the FB 

Glove box line spectra and the HB Glovebox spectra depicted. 

The fraction of the total dose in each neutron energy group was determined by subdividing the 

neutron spectra into the four lower neutron energy groups discussed in the External Dose 

Reconstruction Implementation Guideline (NIOSH 2002) (See overlays in Figure 5.3.4.2-1. The 

highest neutron energy group (>20 MeV) was not used since operations at SRS did not produce 

neutrons of this energy. The dose for each neutron energy group was calculated by multiplying the 

neutron flux (Ø) provided in Brackenbush et al (1987) by the corresponding flux to dose-rate 

conversion factors (DCF) found in NCRP 38. The neutron doses in each NCRP 38 energy interval 

are then summed to develop the four neutron group doses. The dose fraction (Df) for each neutron 

energy group (n) was then calculated by dividing the neutron group dose by the total dose (DT) (Eqn 

5.1). 

T

i i i

f n D

E DCF

D E

∑

=

( )

( )

φ

(5–1)

where:

Ø(Ei

) = Neutron flux of the ith energy bin

DCFi = NCRP 38 (1971) flux to dose-rate conversion factor for the ith energy bin

DT = Total dose

Tables 5.3.4.2.1.1-1 and 5.3.4.2.1.1-2 provide the neutron dose fractions by energy group using the 

data measured by Brackenbush et al (1

T

i i i

f n D

E DCF

D E

∑

=

( )

( )

φ

987).

Although Brackenbush et al. (1987) measured some dose from lower (<10 keV) and intermediate (10-

100 keV) energy neutrons, the contribution to the total dose was less than 5%, respectively. The 

Radiation Effectiveness Factors used in the Interactive RadioEpidemiological Program (IREP) to 
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calculate the Probability of Causation (PC) for these two neutron energy groups are less than the 

fission spectrum energy group (0.1 – 2 MeV) (Kocher et al., 2002). As a result combining the lower 

Table 5.3.4.2.1.1-1. Measured dose fractions at different locations in H 

Area.

Neutron 

energy group

HB-line 

glovebox 

HB-line in 

front of hood

Through glove 

port (extremity) Default

< 10 keV 0.041 0.003 0.028

10 – 100 keV 0.026 0.004 0.002

0.1 – 2 MeV 0.438 0.584 0.563

2 – 20 MeV 0.495 0.409 0.407

Claimant favorable dose fractions

0.1 – 2 MeV 0.5 0.6 0.6 0.6

2 – 20 MeV 0.5 0.4 0.4 0.4

Table 5.3.4.2.1.1-2. Measured dose fractions at 

different locations in F Area.

Neutron 

energy group

FB-line 

glovebox 

Storage 

area Default

< 10 keV 0.009 0.029

10 – 100 keV 0.007 0.017

0.1 – 2 MeV 0.966 0.816

2 – 20 MeV 0.018 0.138

Claimant favorable dose fractions

0.1 – 2 MeV 1.0 0.86 1.0

2 – 20 MeV 0.14

and intermediate energy neutron groups into the fission spectrum group is a reasonable and claimant 

favorable simplification of the dose calculation. In both table 5.3.4.2.1.1-1 and 5.3.4.2.1.1-2, claimant 

favorable default values are also provided when there is insufficient work history information to

determine the type of operation the energy employee conducted. 

5.3.4.2.1.2 Neutron to Photon Ratio

Trends in the annual SRS and Hanford neutron collective dose (Taylor et al 1995, Buschbom and 

Gilbert 1993, respectively), normalized to the annual plutonium production (DOE 1996), are illustrated 

in Figure 5.3.4.2-2. It is evident in this figure that the collective neutron dose was under-recorded 

prior to implementation on 1 January 1971 of the SRS TLND and 1 January 1972 for the Hanford 

TLD. The extent of the under-estimate is difficult to estimate. SRS and Hanford showed a significant 

increase in the ratio of the annual collective neutron dose to the annual plutonium production when 

the TLD neutron dosimeters were implemented. The continued increase in the ratio shown in Figure 

5.3.4.2-1 may be attributable to the increasingly lower handling of plutonium as done during 

production compared to increasing work activities in later years to store plutonium with the associated 

high scatter of neutrons from shielding. The TLND is well known to over-respond to lower energy 

neutrons (Scherpelz et al 2002). 

Subsequent to the implementation of the Hanford TLD on 1 January 1972, AEC headquarters staff 

conducted a detailed review of recorded neutron dose for Hanford personnel using a committee of 

technical experts from Hanford, SRS, and other AEC facilities (Biles 1972). Central to this 

investigation was the selection of 18 long-term Hanford workers for detailed evaluation. Fix et al 

(1996) conducted a later analysis of this information using additional dosimetry data recorded after 

1972. Three distinct periods of dose recording from 1950 through the present are identified in 
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Table 5.3.4.2.1.2-1 corresponding to the respective Hanford two-element, multi-element and 

thermoluminescent dosimeters that parallel similar dosimeters implemented at SRS. 

Year

1940 1950 1960 1970 1980 1990

Collective Neutron Dose / Pu Production

0

50

100

150

200

250

300

350

400

450

SRS 

Hanford 

TNLD use began

Hanford

SRS

Figure 5.3.4.2-2. Trends in SRS and Hanford collective neutron dose normalized to 

plutonium production.

Table 5.3.4.2.1.2-1. Recording periods for selected Hanford plutonium workers.

Category Period Description

A 1950-56 This period of record involved use of the original Hanford two-element dosimeter for 

nonpenetrating (shallow) and penetrating (deep) dose components and NTA film for 

neutron radiation. 

B 1957-71 This period of record involved use of the Hanford multi-element film dosimeter for 

nonpenetrating, x-ray and penetrating dose components, and NTA film for neutron 

radiation. 

C 1972-95 This period of record involved use of the Hanford TLD for beta, photon, and neutron 

dose components.

All of these workers had the Hanford plutonium facility as their primary work area at least during the 

1970s. Nine of the 18 workers have dose histories that extend from 1950 or earlier through 1980 or 

later. It is interesting to examine trends in this data. For example, the ratios of the shallow to deep 

dose and neutron to deep dose for the respective recording periods are summarized in Table 

5.3.4.2.1.2-2. These ratios exhibit the expected response characteristics of the respective dosimeters 

based on laboratory studies.

The shallow dose response for the two-element film dosimeter used at Hanford prior to 1957 is 

expected to be too high because of the significant over-response of this dosimeter to the low-energy 

photons prevalent in plutonium facilities (Figure 5.3.4.2-2). These data also show increasing levels of 

recorded neutron dose, relative to the deep dose, for each succeeding dosimeter design. 
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Table 5.3.4.2.1.2-2. Hanford ratio of recorded dose 

components.

Recording period

Ratio (range)

Shallow/deep Neutron/deep

1950-56 1.6 (1.1 - 3.7) 0.003 (0 - 0.06)

1957-71 1.2 (1.1 - 1.7) 0.4 (0.1 – 0.7)

1972-95 1.3 (1.1 – 1.5) 0.6 (0.1 – 1.6)

While no formal report has been published establishing a neutron to photon ratio for Savannah River 

Workers, a review has been conducted as part of this Technical Basis Document to validate the 

assumption that the Hanford plutonium worker’s neutron to photon ratio was similar to that 

experienced by Savannah River workers. Neutron to photon ratios were evaluated for nine plutonium 

workers (5 from F-Area and 4 from H-Area) totaling 38 monitoring years of data post 1971 (Table 

5.3.4.2.1.1-3). While the average neutron to photon ratio was similar to the ratio observed by Fix et al 

(1996), there was a significant difference in the upper bound of the HB-Line. This variation could be 

the result of the higher specific activity of Pu-238 in the HB line compared to Pu-239 in the FB-line. 

Table 5.3.4.2.1.1-3. Post-1971 SRS ratio of FB and HB Line recorded dose 

components.

Area

Neutron/deep ratio Ratio (average, range)

(GM, GSD, 95th%) SRS neutron/deep Hanford neutron/deep

FB Line 0.36, 2.52, 1.65 0.52 (0.09 – 1.23) 0.6 (0.1 – 1.6)

HB Line 0.91. 2.84, 5.05 1.29 (0.05 – 3.10)

It is important to note that while an average was reported, the distribution of ratios more closely 

resembled a log normal distribution than a normal distribution. A distribution fit was conducted using 

the Kolmogrov-Smirnoff test statistic. The log normal distribution was demonstrated the best fit of the 

four distributions chosen (normal log normal, triangular, and uniform). For this reason, the geometric 

mean, geometric standard deviation and the corresponding 95th% is provided in Table 5.3.4.2.1.1-3. 

When incorporating uncertainty into the dose reconstruction, these values should be used in 

accordance with guidance on combining uncertainty provided in the External Dose Reconstruction 

Implementation Guideline (NIOSH 2002). 

Additional neutron to photon doses was found in the Monthly Works Technical Report (DuPont, 1965). 

A survey ratio of 0.66:1 was found for the HB-Line, and a ratio of 1:1 was observed for the 235-F 

facility during this time frame. Brackenbush et al (1987) reported a neutron to gamma ratio of 

approximately 2:1. This variation in neutron to photon ratio is encompassed by the relatively large 

uncertainty distribution observed with worker data. 

5.3.4.2.2 Neutron Source Production Facilities (773-A)

A second area of neutron exposures at the Savannah River Site (SRS) resulted from the production of 

neutron sources such as Cf-252. Following irradiation in the reactors, Cf-252 was separated. Unlike 

the plutonium facilities in which alpha particle interactions with light materials such as fluorine produce 

the neutrons (α,n reactions), Cf-252 produces neutrons by spontaneous fission. As a result, the 

neutron energy spectra is considerably higher than the plutonium facilities. 

5.3.4.2.2.1 Neutron Energy Spectra

Measurements made by Brackenbush et al (1987) at the 773-A facility clearly demonstrated the shift 

in the neutron energy spectra. Using the same methodology described in the plutonium facility 

section, the neutron dose fractions were developed and are provided in Table 5.3.4.2.2.1-1. 
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Table 5.3.4.2.2.1-1. Dose fractions for various 

operations in 773-A area.

Neutron 

energy group Glovebox Shipping cask Default

< 10 keV 0.018 0.015

10 – 100 keV 0.010 0.023

0.1 – 2 MeV 0.419 0.349

2 – 20 MeV 0.553 0.613

Claimant favorable dose fractions

0.1 – 2 MeV 0.45 0.39 0.5

2 – 20 MeV 0.55 0.61 0.5

As with the plutonium facility dose fractions, a claimant favorable default value is provided when there 

is insufficient information to determine which dose fractions should be used. 

5.3.4.2.2.2 Neutron to Photon Ratio

As indicated with the plutonium workers, a review was conducted as part of this Technical Basis 

Document to evaluate the assumption that the Hanford plutonium worker’s neutron to photon ratio 

was similar to that experienced by Savannah River neutron source production workers. Neutron to 

photon ratios were evaluated for 6 neutron source production workers from the 773-A area totaling 30 

monitoring years of data post 1971 (Table 5.3.4.2.2.2-1). While the average neutron to photon ratio 

was slightly higher than the ratio observed by Fix et al (1996) for Hanford plutonium workers, there 

was a factor of 2 increase in the upper range. 

Table 5.3.4.2.2.2-1. Post-1971 SRS ratio of 773-A recorded dose 

components.

Area

Neutron/deep ratio

(GM, GSD, 95th%)

Ratio (average, range)

SRS neutron/deep Hanford neutron/deep

773-A 0.62, 2.29, 2.41 0.85 (0.10 – 3.83) 0.6 (0.1 – 1.6)

5.3.4.2.3 Reactor Facilities (100 Area)

The neutron dose to workers in SRS reactor facilities is very low in comparison to the measured 

photon and beta dose received from contaminated equipment and plant areas, spent reactor fuel, 

activated reactor components, and other areas containing fission and activation produces 

encountered during plant maintenance. However, there are a few areas around the reactors where a 

worker can receive neutron exposure, albeit a small fraction of the photon dose, during reactor 

operations. 

5.3.4.2.3.1 Neutron Energy

Brackenbush et al. (1987) recorded measurement data at the K-reactor door during operations is 

shown in Figure 5.3.4.2-3. The data indicated the lowest average neutron energy (170 keV) of all 

facilities measured at SRS. 

Table 5.3.4.2.3.1-1 provides the dose fraction per neutron energy group. Unlike the previous two 

neutron exposure areas, the dose fraction from the intermediate energy neutrons was approximately 

12% while the lower energy neutrons remained less than 5%. Since the Radiation Effectiveness 

Factor (REF) for intermediate energy neutrons is greater than for lower energy neutrons, a claimant 

favorable simplification is to combine the dose fractions for lower and intermediate energy neutrons. 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 106 of 248

K-Reactor Behind Rector Shielding Wall

Neutron Energy 

10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101 102

Dose Equivalent (mrem)

10-5

10-4

10-3

10-2

10-1

100

101

0.07 mrem

0.26 mrem

1.95 mrem

0.0001 mrem

Figure 5.3.4.2-3. Neutron energy spectra recorded behind the 

K-Reactor shield door with neutron energy groups overlay.

Table 5.3.4.2.3.1-1. Dose fractions for 

heavy water reactors at SRS.

Neutron 

energy group

Behind 

shield door Default

< 10 keV 0.030

10 – 100 keV 0.115

0.1 – 2 MeV 0.855

2 – 20 MeV 0.0001

Claimant favorable dose fractions

10 – 100 keV 0.15 0.15

0.1 – 2 MeV 0.85 0.85

5.3.4.2.3.2 Neutron-to-Photon Ratio

Neutron exposures are expected to be very low compared to beta and photon exposures around the 

reactors because there is no neutron radiation when the reactors are shutdown for maintenance or 

refueling when most beta and photon dose to workers is received. Brackenbush et al (1987) reported 

both the neutron and photon dose rate at the F-Area and the K-Reactor. The neutron to photon dose 

rate ratio for the F-area was 2:1, while the neutron to photon dose rate ratio for the K reactor was 1:1 

or about half of the observed ratio from the F-Area. Using the simple dose-rate ratio reduction factor 

of 2, the neutron to photon ratio for reactor area workers can be developed based on the personal 

dose ratio observed for the F-Area workers. Table 5.3.4.2.3.2-1 provides the F-Area neutron to 

photon ratio and calculated reactor area neutron to photon ratio. 

Fix et al (1996) recalculated neutron doses and evaluated the neutron to photon ratio between the 

recalculated neutron dose and the photon dose using several methods. Based on this data, the most 

claimant favorable neutron to photon ratio for 100 area workers at Hanford between 1950 and 1961 

ranged from 0.13 to 0.73 with a weighted average of 0.26. The SRS neutron to photon ratio is 
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Table 5.3.4.2.3.2-1. Post-1971 SRS ratio of plutonium and reactor recorded dose 

components.

Area

Neutron/photon ratio

(GM, GSD, 95th%)

Ratio (average, range)

SRS neutron/photon Hanford neutron/photon

Pu facility 0.36, 2.52, 1.65 0.52 (0.09 – 1.23) 0.6 (0.1 – 1.6)

Reactor 0.18, 2.52, 0.82 0.26 (0.05 – 0.62) 0.26 (0.13 – 0.73)

expected to be considerably lower than that observed from the Hanford reactors because there was 

minimal neutron radiation in routine work areas during reactor operation due to design (below grade) 

and shielding. Unlike the graphite reactors, in which workers could conduct work on the face of the 

reactor during operation, access to the lower plenum where the reactor core and cooling system were 

located was restricted. In addition, since most of the photon dose was received during refueling

operations, increasing the denominator of the ratio, thus further lowering the ratio compared to 

plutonium facility workers. As a result, the neutron to photon ratio for SRS reactors is considered to 

be an upper bound of the actual neutron to photon ratio. 

Before applying the ratio listed in Table 5.3.4.2.3.1-1, careful consideration should be given to the 

energy employee’s work history. There are occupations around the reactor areas in which no neutron 

dose would be expected (office or administrative staff). Generally only workers conducting work near 

the entrances to the lower plenum (Shield Door), the crane wash area, or the disassembly area have 

the potential for neutron exposure. There was also some neutron exposure in the assembly area due 

specific campaigns in which Pu-Al alloy targets were assembled. In the absence of claim – specific 

data to arrive at a neutron to photon ratio from Table 5.3.4.2.3.2-1 , a neutron to photon ratio of 0.104

5.3.4.2.4 Other Facilities

should be applied to all non-administrative job categories. 

5.3.4.2.4.1 Calibration Facilities (736-A Area)

The neutron exposure spectra for workers at the SRS Calibration Facility cover the entire energy 

spectra. Although Brackenbush et al (1987) measured multiple spectra at this facility, only two that 

represent the extremes are discussed in this analysis. The first is the routine calibration of dosimeters 

using a PuF4 source in free air. The second is the spectra recorded using a PuBe source at a tank 

containing 150 mm of D2O. Table 5.3.4.2.4.1-1 demonstrates the wide variation in the resultant dose 

fractions. Since the relative use of these different sources is not known and the Radiation 

Effectiveness Factor (REF) is the largest (claimant favorable) for fission spectrum neutrons, the 

default was chosen based on the PuF4 source. 

Table 5.3.4.2.4.1-1. Dose fractions at the calibration 

facility (736-A).

Neutron 

energy group

PuF4 source 

w/o drum 

PuBe source at 

150 mm D2O Default

< 10 keV 0.008 0.127

10 – 100 keV 0.007 0.046

0.1 – 2 MeV 0.819 0.374

2 – 20 MeV 0.166 0.453

Claimant favorable dose fractions

0.1 – 2 MeV 0.83

2 – 20 MeV 0.17

 4. Claimant favorable factor determined during June 19, 2003 teleconference among NIOSH, NIOSH contractor and SRS 

technical staff for default use based on minimal neutron dose in SRS reactors, absolutely, and certainly in comparison 

with the associated photon dose. 
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The neutron to photon ratio for the calibration facility is not currently known. However, since 

intermittent exposure is considered the most probable scenario, the neutron to photon ratio is not 

expected to be greater than that observed at the 773-A facility. A reasonable and claimant favorable 

assumption would be to use the neutron to photon ratio for the 773-A facility. 

5.3.4.2.4.2 Fuel Fabrication Area (321-M) 

For certain campaigns, Pu-Al alloy targets were fabricated in the 321-M area. Figure 5.3.4.2-4 

reproduced from Brackenbush et al (1987), compares the neutron spectra of Pu-Al to various other 

neutron sources. As can be seen from the graph, Pu-Al alloy has a similar neutron energy spectra to 

PuF4. Since the Pu-Al alloy is similar to a bare PuF4 source, the dose fraction for the PuF4 source 

w/o Drum should be used (Table 5.3.4.2.3.1-1). Neutron exposures in the Fuel Fabrication Area are 

expected to be intermittent and were not continuous over the operation of the SRS, thus a neutron to 

photon ratio should not be used. Based on dosimetry records from SRS, it appears that during the 

Pu-Al campaigns, workers were monitored for neutron exposure. Since the Pu-Al alloy spectra is of 

similar energy to PuF4, approximately 90% of the spectra is greater than the 500 keV threshold for 

NTA film. As a result, the NTA film measurements are expected to be reasonably accurate to within 

parameters discussed in the External Dose Reconstruction Implementation Guideline (NIOSH 2002). 

Figure 5.3.4.2-4. Comparison of various neutron spectra from 

different plutonium compounds and alloys (Brackenbush et al, 

1987).

5.3.5

Uncertainty in the recorded dose is an important consideration in the worker favorable analyses. The 

overall uncertainty is dependent upon: 1) administrative practices, 2) dosimetry technology, 3) 

calibration and 4) workplace radiation fields. The potential effect of each of these parameters on the 

recorded dose was described in the preceding sections. For dose reconstruction under EEOICPA, 

based on the combination of the foregoing parameters, the judged estimates in this TBD of 

uncertainty in recorded dose for SRS workers according to primary SRS facilities based on the

dosimeter type and period of use are presented in Table 5.3.5-1.

SRS Workplace Recorded Dose Uncertainty
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Table 5.3.5-1. Estimates of uncertainty.

Dosimeter SRS

Laboratory 

uncertaintya

Workplace uncertaintyb

Reactor Plutonium Reprocessing

Beta/gamma dosimeters

Two-element film Used 1951–1959 ±25% ±50% ±75% ±50%

Multi-element film Used 1960–69 ±20% ±40% ±60% ±40%

TLD Used 1970–present ±10% ±20% ±30% ±20%

Neutron dosimeters

NTA Used 1951–1970 ±50% ±100% (need to use another method)

TLND Used 1971–present ±25% ±50% ±75% ±75%

a. In relation to Hp(10) response of dosimeter.

b. 95% confidence interval.

As reported in the PNNL report, measurements with the TEPC, multisphere system and 3HE 

spectrometer were in general agreement. The TLND agreed within about 30% for most measurement 

locations along the plutonium production lines and storage areas. The TLND was within a factor of 3 

(i.e., 0.3 to 3) for the extremes in neutron energy spectra encountered at the K-reactor door (i.e., 

highly thermalized field) and for a californium shipping cast (i.e., where most lower energy neutrons 

had been removed). Over long time periods, workers would generally be expected to be involved in 

several different exposure profiles that will serve to minimize the extremes identified. These results 

are indicative of the technical difficulties to accurately measure neutron dose in the workplace. Table 

5.3.5-2 presents a summary of common workplace neutron dosimeter performance characteristics. 

Table 5.3.5-2. Common workplace neutron dosimeter Hp(10) performance.

a

Parameter Description Workplace biasa

Workplace neutron 

energy spectra

SRS NTA dosimeter response 

decreases and SRS TLD response 

increases with decreasing neutron 

energy.

Depends upon workplace neutron spectra. NTA 

recorded dose of record likely too low. Article by 

Hoy presented in Vallario et al (1969) identified 

ratios of 2 to 4 under-response. 

Exposure 

geometry

NTA Dosimeter response increases 

with increasing exposure angle and 

TLD response decreases with 

increasing exposure angle. 

NTA recorded dose likely too high since dosimeter 

response is higher angles other than A-P. TLD 

recorded dose of record likely too low since 

dosimeter response is lower at angles other than AP. Effect is highly dependent upon neutron energy. 

Dosimeter 

placement

SRS apparently instructed workers to 

wear NTA dosimeter facing the 

workers body to enhance dosimeter 

response (i.e., to respond to 

backscatter neutron radiation).

Recorded dose of record likely too low. Reflected 

neutrons are expected to be less-than the NTA 

energy threshold identified as 0.5 MeV by SRS. 

Mixed low energy 

photon and 

neutron radiation

SRS NTA dosimeter responds 

significantly to lower-energy 

photons that can result in fogging of 

the NTA film.

NTA recorded dose of record likely too low. SRS NTA 

dosimeters selected to minimize photon response. 

SRS TLDs have capability to remove photon dose 

contribution. 

a. Judgment based on SRS dosimeter response characteristics and workplace radiation fields.

b. Recorded dose compared to Hp(10).

Taylor et al (1995) reported that SRS participated in Personal Dosimetry Intercomparison Studies 

starting in 1974. SRS dosimetry demonstrated an average error of approximately 40% in the first 

intercomparison. According to Taylor et al (1995), a marked improvement was noted and the average 

error was 13.9% in 1985. The SRS site has noted a trend of improving neutron dosimetry through the 

years. Measurements of TLND performance at SRS in 1987 (Brackenbush et al., 1987) indicate that 

the SRS measured neutron dose with the TLND (beginning 1 January 1971) is reasonably correct. 

For dose reconstruction under EEOICPA, a claimant favorable standard error estimate of 50% should 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 110 of 248

be made for neutron dosimetry between 1971 and 1985. Starting in 1985, a standard error of 15% 

should be assumed in the neutron recorded dose. 

5.4 ADJUSTMENTS TO RECORDED DOSE

Adjustments to the SRS reported dose are necessary to arrive at a claimant-favorable dose 

considering the uncertainty associated primarily with the complex SRS workplace radiation fields and 

exposure orientations. 

5.4.1

Taylor et al (1995) identified calibration corrections to the reported SRS TLD-Deep photon dose for 

the periods prior to 1986 and 1987, respectively, to estimate the Hp(10) dose. Photon dose 

corrections summarized in Table 5.4.1-1 are necessary to calculate an adjusted dose that is claimant 

favorable considering historical technological limitations of SRS dosimeters. For the years 1987 to 

the present, no correction is needed, the reported dose is Hp(10) equivalent.

Photon Dose Adjustments

Table 5.4.1-1. Adjustments to reported SRS deep photon dose.

Time period Dosimeter Facility Step Adjustment to reported dose

Prior to 

1/1/1986

All beta/photon 

dosimeters

All facilities A Multiply reported TLD-Deep photon dose by a 

factor of 1.119 to estimate Hp(10).

For the year 

1986

TLD beta/photon 

dosimeter

All facilities B Multiply reported TLD-Deep photon dose by a 

factor of 1.039 to estimate Hp(10).

5.4.2

The Savannah River Site (SRS) incorporated the energy variation of the dose equivalent into their 

calibration methodology. As a result, the recorded dose equivalent (DER) is a combination of all 

neutron energies. In order to calculate the probability of causation, the recorded neutron dose must 

be separated into neutron energy groups as discussed in section 5.3.4.2 and subsequently converted 

to ICRP 60 (1990) methodology. 

Neutron Dose Adjustments

5.4.2.1 Neutron Weighting Factor

Adjustment to the neutron dose is necessary to account for the change in neutron quality factors 

between historical and current scientific guidance as described in NIOSH (2002). At SRS, a single 

quality factor was not used, but rather a continuum of quality factors based on neutron energy that 

varies up to 11 at 1 MeV (Singh 2002). The quality factor was incorporated into the calibration 

methodology which used flux to dose-rate conversion factors for varying neutron energies 

(Brackenbush 1987, Taylor 1995). According to site personnel (Crase 2003), the SRS flux to doserate conversion factors were based on NCRP 38 (1971). NCRP 38 (1971) lists both flux to dose-rate 

conversion factors and associated quality factors that vary from 2 at energies less than 1 keV to 11 at 

1 MeV. In order to convert from NCRP 38 quality factors to ICRP 60 (1990) radiation weighting 

factors, a curve was fit describing the neutron quality factors as a function of neutron energy. The 

average quality factor for each neutron energy group was developed by integrating the area under 

curve and dividing by the neutron energy range as shown in equation 5.2. 

(2.0 0.1)

( )

( )

2.0

0.1

,0.1 2.0 − =

∫ − Range

Q E dE

Q E

f

n MeV (5–2)
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Table 5.4.2-1 provides a summary of historical changes in the quality factors and the average NCRP 

38 quality factor for the neutron energy groups used in dose reconstruction. 

5.4.2.2 Neutron Correction Factor

Table 5.4.2-1 provides the average quality factor for the four neutron energy groups which encompass 

SRS neutron exposures. The neutron dose equivalent correction factor can be calculated by dividing 

the dose fractions from section 5.3.4.2 for each neutron energy group (Df(En)) by the corresponding 

energy specific average NCRP 38 quality factor (Q(En)) and then multiplying by the ICRP 60 radiation 

weighting factor (wR) as shown in equation 5.3. 

Table 5.4.2-1. Historical neutron quality or weighting factors.

Neutron 

energy

(MeV)

Historical 

dosimetry 

guidelinea

NCRP 38

quality 

factorsb

Average 

quality factor 

used at SRS

ICRP 60 neutron 

weighting factor, 

wr 

c

2.5 × 10-8 3 2

2.35 5

1 × 10-7

10

2

1 × 10-6 2

1 × 10-5 2

1 × 10-4 2

1 × 10-3 2

1 × 10-2 2.5 5.38 10

1 × 10-1 7.5

5 × 10 10.49 20 -1 11

1 11

2 10

7.56 10

2.5 9

5 8

7 7

10 6.5

14 7.5

20 8

40 7 Not applicable 5

60 5.5

a. Trilateral meeting in 1949 radiation protection guidelines (Fix et al 1994). 

b. Recommendations of NCRP 38.

c. ICRP 60 (1990) 

R

n

f n

f n w

Q E

D E C E = ×

( )

( ) ( ) (5–3)

Table 5.4.2.2-1 summarizes the default neutron dose distributions by energy for each neutron 

exposure area. By multiplying the recorded neutron dose by the area specific correction factors the 

neutron dose equivalent is calculated. For example consider a 1000 mrem recorded neutron dose by 

an energy employee working at the HB-Line, the corrected neutron dose is 1144 mrem from neutrons 

between 0.1-2.0 MeV and 529 mrem from neutrons with energy between 2 and 20 MeV. Thus the 

corrected neutron dose is a total of 1673 mrem. These corrections should be applied to measured 

dose, missed dose and dose determined based on a neutron to photon ratio. Table 5.4.2.2-1 

summarizes the default dose fractions by energy and the associated ICRP 60 correction factors for 

each neutron exposure area. 
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Table 5.4.2.2-1. Summary of dose fractions and associated ICRP 60 correction factors.

Process Description / buildings

Operations Neutron 

energy

Default dose 

fraction (%)

ICRP 60 correction 

Begin End factor

Fuel fabrication Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and plutonium – aluminum 

(Pu-Al) alloy targets.

(321-M) 0.1-2 MeV

2-20 MeV

83%

17%

1.58

0.23

Reactors During Reactor Operation: Low level neutron exposure through shielding. In addition, some intermittent neutron 

exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987

10-100 keV

0.1 – 2 MeV

15%

85%

0.28

1.62

(105K) 10/1954 1/1988

1992 7/1992

(105L) 7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium 

production

Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into production pucks. Work 

is primarily conducted in glove boxes with predominant close anterior exposure to workers. Radiation levels at the 

beginning of the process are fairly constant while levels at end of process are closely related to production levels. 

Plutonium Finishing Process

(221-H, B-Line)

11/1954 1988 0.1-2 MeV

2-20 MeV

60%

40%

1.14

0.53

Plutonium Finishing Process

(221-F, B-Line)

10/1955 1988 0.1-2 MeV 100% 1.91

Plutonium Storage 0.1-2 MeV

2-20 MeV

86%

14%

1.64

0.19

Production Control Laboratory

(772-F)

1955 0.1-2 MeV 100% 1.91

Plutonium Fuel Facility (235-F) 1955 0.1-2 MeV 100% 1.91

Radionuclide 

production and 

calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS has produced several 

radionuclides including 238Pu, 239Pu, americium, curium, californium, etc. At the calibration facility radiological 

instruments are calibrated using a variety of sources with different neutron energy spectra.

Californium Production Facility 

(773-A)

0.1-2 MeV

2-20 MeV

50%

50%

0.95

0.66

Calibration Facility (736-A) 0.1-2 MeV

2-20 MeV

83%

17%

1.58

0.23

5.5 MISSED DOSE

There is undoubtedly missed recorded dose for SRS workers. Analysis of the missed dose has been 

separated according to photon and neutron missed dose. 

5.5.1

Missed photon dose for SRS workers would occur where: 1) there is no recorded dose because 

workers were not monitored or the dose is unavailable and 2) a zero dose is recorded for the 

respective SRS dosimeter systems for any dosimeter response less-than the minimum detection level 

(MDL). Methods to be considered when there is no recorded dose for a period within a working 

career have been examined by Watson et al (1994). Generally, estimates of the missed dose can be 

made using dose results for co-workers or using the recorded dose prior to and after the period of 

missed dose. However, these situations do require careful examination. The missed dose for 

dosimeter results less-than the MDL is particularly important for earlier years when MDLs were higher 

and dosimeter exchange was more frequent. NIOSH (2002) describes options to calculate the 

missed dose. One option is estimate a claimant favorable maximum potential missed dose where the 

Minimum Detection Level (MDL) is multiplied by the number of zero dose results providing an 

estimate of the maximum potential missed dose. The following sections describe the potential missed 

photon dose corrections according to facility/location, dosimeter type, year and energy range.

Photon Missed Dose
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5.5.1.1 Facility/Location

The potential missed photon dose for the laboratory determined MDL and exchange frequency using 

NIOSH (2002) is presented in Table 5.5.1-1. 

Table 5.5.1-1. Missed photon dose according to SRS facility.

SRS facility type Dosimeter Period of use

Exchange 

frequency

MDL 

(mSv)a

Max. annual 

missed dose

(mSv)b

Fuel fabrication, SRS 

reactors, processing 

plants, waste handling

ORNL two-element 

film dosimeter

Prior to 3/52 Weekly 0.4 20.8 

SRS two-element film 

dosimeter

After 3/52 through 

9/30/1957

Weekly 0.4 20.8 

10/1/1957 through 

11/8/1959

Biweekly 0.4 10.4 

Above plus plutonium 

and nuclide production 

SRS multi-element 

film dosimeter

11/9/1959 through 

12/31/1964

Biweekly 0.4 10.4

1/1/1965 through 

12/31/1965

4-week 0.4 5.2

1/1/1966 through 

3/31/1970

Monthly 0.4 4.8

All SRS SRS TLD 4/1/1970 through 

6/30/1983

Monthly 0.15 1.8

Panasonic TLD 7/1/1983 to 2003 

(ongoing)

Monthly 0.05 0.6

Panasonic TLND 1/1/1995 to 2003 

(ongoing)

Monthly (n=12) 0.15 0.9

a. Estimated film dosimeter detection levels based on NIOSH (1993), NRC (1989), and Wilson et al (1990). TLD detection 

levels from Taylor et al (1995) and SRS (1993). 

b. Maximum annual missed dose calculated from OCAS-IG-001 (NIOSH 2000). 

5.5.1.2 Dosimeter Type

The MDLs for the respect SRS beta and photon dosimeters normally cited are based on laboratory 

irradiations. The actual MDLs are greater than these values because of additional uncertainty in 

actual field use. Reasonable MDLs are shown in Table 5.5.1-2 for most applications based on NIOSH 

(1993), NRC (1989), and Wilson et al (1990). The TLD detection levels from Taylor et al (1995) and 

SRS (1993). 

Table 5.5.1-2. Missed photon dose according to dosimeter type.

Dosimeter type Period of use

MDL

(mSv)

Exchange 

frequency

Max. annual 

missed dose

(mSv)

Two-element film Prior to 3/52 through 9/30/1957 0.4 Weekly (n = 52) 20.8

10/1/1957 through 11/8/1959 0.4 Biweekly (n=26) 10.4

Multi-element film 11/9/1959 through 12/31/1964 0.4 Biweekly (n=26) 10.4

1/1/1965 through 12/31/1965 0.4 4-week (n=13) 5.2

1/1/1966 through 3/31/1970 0.4 Monthly (n=12) 4.8

TLD 4/1/1970 through 6/30/1983 0.15 Monthly (n=12) 1.8

7/1/1983 to 2003 (ongoing) 0.05 Monthly (n=12) 0.6

0.05 Quarterly (n = 4) 0.2
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5.5.1.3 Year

Analysis of the missed photon dose according to year (actually by period according to dosimeter type 

and exchange) may be needed to evaluate some claim information particularly if only annual dose 

data are available. Table 5.5.1-3 summarizes the potential missed dose. 

Table 5.5.1-3. Missed photon dose adjustments to recorded deep dose.

Period of use

MDL 

(mSv) Exchange frequency

Max. annual 

missed dose

(mSv)

Prior to 3/52 through 9/30/1957 0.4 Weekly (n = 52) 20.8

10/1/1957 through 11/8/1959 0.4 Biweekly (n=26) 10.4

11/9/1959 through 12/31/1964 0.4 Biweekly (n=26) 10.4

1/1/1965 through 12/31/1965 0.4 4-week (n=13) 5.2

1/1/1966 through 3/31/1970 0.4 Monthly (n=12) 4.8

4/1/1970 through 6/30/1983 0.15 Monthly (n=12) 1.8

7/1/1983 to 2003 (ongoing) 0.05 Monthly (n=12) 0.6

5.5.1.4 Energy Range

An estimate of the missed photon dose by energy range is possible based on the type of facility and 

predominant radionuclides such as intermediate (>100 keV) energies for all facilities handling 

activation and fission product nuclides, primarily lower (<100 keV) energy photons for plutonium 

facilities and lower energy photons for uranium fuel fabrication facilities. The recorded dose from the 

dosimeter response does not typically provide sufficient information to estimate discrete energy 

ranges. It is possible to examine the energy response characteristics of the respective multi-element 

dosimeters but this analysis does not recognize the substantial uncertainties present in the workplace 

associated with shielding, radiation scattering and mixed radiation fields.

5.5.2

There is potential for significant missed dose among workers at the plutonium finishing facilities and 

the neutron source production area. There is also some potential for missed dose from work around 

the reactors and at the fuel fabrication facility. The approach used to calculate the neutron missed 

dose can be divided into two time periods. The first is before 1971 when neutron track emulsion type 

A (NTA) film was used and the second is when thermoluminescent dosimeters were used. Table 

5.5.2-1 provides a summary of the reported limits of detection by the Savannah River Site (SRS).

Neutron Missed Dose

Table 5.5.2-1. Summary of SRS neutron dosimeter limits of detection (LOD).

Dosimeter type

Dates of use

Exchange 

frequency

Reported 

LOD

(mrem)

Estimated 

LOD

(mrem)

Max. annual 

missed dose

Begin End (mrem)

ORNL NTA 1951 8/3/1953 Weekly (n = 52) ~50a ~50 2,600

SRS NTA 8/3/1953 7/13/1960 Weekly (n = 52) 30 ~40b 2,080

SRS NTA 7/14/1960 12/31/1970 Biweekly (n=26) 30 ~40b 1,040

SRS Hoy TLND 1/1/1971 12/31/1994 Monthly (n=12) 10c 20 240

Panasonic TLND 1/1/1995 Present Monthly (n=12) 15 15 180

a. Based on Wilson et al (1990).

b. Correction for estimated 25% under response of NTA film for some facilities.

c. Minimum Recorded Dose not necessarily LOD.
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5.5.2.1 Prior to 1971

Due to the degraded energy spectra at most SRS facilities, the NTA film generally under responded 

due to the energy threshold of approximately 500 keV. The magnitude of the NTA film under 

response ranges from about 10% at the 773-A facility to over 90% around the reactors. According to 

Taylor et al. (1995), the SRS calibration method for NTA film might have accounted for some of this 

under response, however, until this can be further evaluated, the claimant favorable approach is to 

assume that it was not incorporated into the calibration methodology. As a result of the inability of 

NTA film to measure neutrons below 500 keV, the “true” limit of detection varies by energy spectra or 

facility. In the case of reactor operations, the limit of detection is virtually meaningless, while for the 

Californium Production facility (773-A), the reported limit of detection is probably reasonable however 

might be under reported by 10%. 

Due to the uncertainty in whether an energy employee’s NTA badge would respond to the workplace 

neutron spectra, using a ratio to the measured photon dose is recommended as a claimant-favorable 

option to reconstruct an individual worker neutron dose. This is based on the fact that for routine 

neutron exposure, the neutron exposure is essentially always accompanied with measurable photon 

exposure. Table 5.5.2-2 provides a summary of the recommended neutron to photon ratios for the 

different process areas. As can be determined from the table, the recommended method to apply the 

ratio is as a log normal distribution using the geometric mean and geometric standard deviation. 

Caution, however, should be employed when using Table 5.5.2-2 and consideration should be given 

to an individual energy employee’s neutron to photon ratio if there is sufficient individual data. For 

example if an individual has recorded neutron dose either prior to or after the estimation period that 

has a higher neutron to photon ratio than what is provided in the table, then the higher neutron to 

photon ratio should be used provided that the work assignments were similar for the two time periods. 

The values in Table 5.5.2-2 are intended to provide guidance when there is known technical 

limitations in the neutron dosimetry technology or no neutron monitoring data. The data were 

developed by evaluating multiple worker years, thus for a given time period, the “true” ratio might be 

more or less than the geometric mean. Due to potential for photon missed dose effecting the neutron 

to photon ratio, when applying the ratio methodology above, the photon dose should be the 

combination of both the measured photon dose (Dosimeter Dose) and the photon missed dose. 

When overestimating dose reconstructions, the upper 95% ratio can be used as a claimant-favorable 

worst case ratio and the associated neutron uncertainty can be omitted. In underestimating dose 

reconstructions, the geometric mean can be used as a best case ratio. 

5.5.2.2 Post-1971

Since the introduction of thermoluminescent neutron dosimeters at the site, neutron exposures have 

been fairly well monitored and the recorded data is reasonably accurate. As a result the guidance on

missed dose provided on the External dose Reconstruction Implementation Guideline (NIOSH 2002) 

should be followed in accordance with the LOD values presented in Table 5.5.2-1. 

5.6 ORGAN DOSE 

Once the Hp(10) adjusted doses have been calculated for each year, these values are used 

according to NIOSH (2002) to calculate the organ dose distribution. 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 116 of 248

Table 5.5.2-2. Summary of neutron to photon ratios by process.

Process Description/buildings

Operations Neutron to photon ratio

Begin End

Geometric 

Mean (GM)

Geometric Standard 

Deviation (GSD)

Upper 

95th% 

Fuel fabrication Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and plutonium –

aluminum alloy targets.

(321-M) (a) (a) (a)

Reactors During Reactor Operation: Low level neutron exposure through shielding. In addition, some intermittent 

neutron exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987

0.18 2.52 0.82

(105K) 10/1954 1/1988

1992 7/1992

(105L) 7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium production Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into production pucks. 

Work is primarily conducted in glove boxes with predominant close anterior exposure to workers. Radiation 

levels at the beginning of the process are fairly constant while levels at end of process are closely related to 

production levels. 

Plutonium Finishing Process

(221-H, B-Line)

11/1954 1988 0.91 2.84 5.05

Plutonium Finishing Process

(221-F, B-Line)

10/1955 1988 0.36 2.52 1.65

Plutonium Storage

Production Control Laboratory 

(772-F)

1955

Plutonium Fuel Facility (235-F) 1955

Radionuclide

production and 

calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS has produced 

several radionuclides including 238Pu, 239Pu, americium, curium, californium, etc. At the calibration facility 

radiological instruments are calibrated using a variety of sources with different neutron energy spectra. 

Californium Production Facility 

(773-A)

0.62 2.29 2.41

Calibration Facility (736-A) (b) (b) (b)

a. During Pu-Al target production, workers with potential for neutron exposure wore NTA film, thus LOD values listed in table 5.5.2-1 should be used. 

b. Neutron to photon ratio for the calibration facility is not expected to be greater than the ratio observed in the californium production facility. As a claimant 

favorable approximation the Californium Production facility ratio can be used. 

5.6.1

NIOSH (2002) describes the methodology used to calculate the organ dose distribution for the 

respective radiation types using identified exposure geometries. The selection of the worker 

orientation is important to the calculation of the organ dose. Examples of common exposure 

orientations are provided in the NIOSH Implementation Guide (2002), Table 4.2. Unfortunately, there 

is no definitive process to determine the exposure geometry for each worker. Table 5.6.1-1 lists 

proposed default options based on judgment of claimant favorable exposure geometries for long-term 

SRS workers. 

Organ Dose Conversion Factors

Table 5.6.1-1. Default exposure geometries to calculate organ dose.

Claim status Job categoryb Exposure geometry Percentagec

Likely non-compensable All AP 100%

Compensable–workers All AP 50%

ROT 50%

Compensable–supervisors All AP 50%

ISO 50%

Attachment E presents a discussion of the use of the various parameters presented in this section 

that will aid the dose reconstructionist in preparing dose reconstructions.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 117 of 248

5.7 DOSE RECONSTRUCTOR INSTRUCTIONS

The information given below summarizes instructions provided to the dose reconstructors in 

evaluating organ doses from external radiation exposure.

5.7.1

Because of the numerous factors involved in the calculation of external doses from monitoring data, 

the entire approach is not described here. Instead, the intent is to provide sufficient details in the 

dose reconstruction report such that, when combined with the spreadsheets used by the dose 

reconstructors and the relevant information in Section 5, it can be verified that claimant-favorable 

external doses have been assigned. However, several key issues are discussed below.

Assignment of External Doses from Monitoring Data: Maximizing Approach for 

Dose Reconstructions

In general, the approach is to maximize the doses based on information presented in Section 5, 

including the use of maximizing energy ranges (e.g., typically 30-250 keV for photons and 0.1-2 MeV 

for neutrons) and exposure geometries. For simplicity and to ensure a claimant-favorable dose 

reconstruction, organ dose conversion factors of 1 are applied if the maximum dose conversion factor 

is less than 1. The dose reconstructor has the option of using non-maximizing energy distributions 

based on the information in Section 5 if more realistic distributions can be established based on work 

location and the potential for exposure. Maximizing dosimeter correction factors are applied based on 

the information provided in Section 5. 

Missed doses are determined based on the actual zero dosimeter readings recorded, or theoretical 

zero readings based on the relevant badge exchange frequencies. As a maximizing assumption and 

for simplicity, missed doses are added to measured doses and the annual summations are treated as 

point estimates for the purpose of calculating probability of causation. Alternatively, the dose 

reconstructor can separate missed dose and use LOD/2 with a lognormal distribution (applying a 

value of 1.52 in Parameter 2 of the IREP input) in accordance with the NIOSH External Dose 

Reconstruction Implementation Guideline.

In the absence of monitoring data, if it cannot be determined that applying only the environmental (onsite ambient) doses is claimant favorable, missed doses are applied based on the relevant dosimeter 

exchange frequency.

5.7.2

Measured doses are treated as a normal distribution with a s.d. of 30% (or other appropriate value 

that may be provided in Section 5). Missed doses are treated as a lognormal distribution. Based on 

the established exposure conditions, the dose reconstructor selects the most appropriate exposure 

geometries and energy ranges to determine the most appropriate organ dose conversion factors. The 

most appropriate dosimeter correction factors are applied based on the information in Section 5. The 

associated organ dose uncertainties are evaluated using a Monte Carlo approach (Crystal Ball 

program).

Assignment of External Doses from Monitoring Data: Best Estimate Approach for 

Dose Reconstructions

6.0

Reserved

TRADESWORKERS
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Ashley, C. and C.C. Zeigler. 1978a. Environmental Monitoring at the Savannah River Plant, Annual 

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Ashley, C. and C.C. Zeigler. 1983. Environmental Monitoring at the Savannah River Plant, Annual 

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Cummins, Carol L., Martin, Donna K., and Todd, James L. Savannah River Site Environmental 

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Volume I, Text, Volume II, Figures and Data Tables. WSRC-RP-89-59-1. Westinghouse 

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Du Pont. 1959c. Health Physics Regional Monitoring Semiannual Report, January through June 1959.

DPSPU-59-11-30. 

Du Pont. 1960a. Health Physics Regional Monitoring Semiannual Report, July through December 

1959. DPSP-60-11-9. 

Du Pont. 1960b. Health Physics Regional Monitoring Semiannual Report, January through June 1960. 

DPSP-60-25-26. 

Du Pont. 1961a. Health Physics Regional Monitoring Semiannual Report, July through December 

1960. DPSP-61-25-4. 
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DPSP-62-25-2. 

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1962. DPSP-63-25-3. 

Du Pont. 1963e. Health Physics Regional Monitoring Semiannual Report, July through December 

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1971. DPSPU-72-30-1. 

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1972. DPSPU-73-30-1. 

Du Pont. 1974. Environmental Monitoring in the Vicinity of the Savannah River Plant. Annual Report, 

1973. DPSPU-74-30-1. 

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1974. DPSPU-75-30-1. 

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1976. DPSPU-77-30-1. 

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1977. DPSPU-78-30-1. 

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1978. DPSPU-79-30-1. 

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1979. DPSPU-80-30-1. 

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1980. DPSPU-81-30-1. 

Du Pont. 1982. Environmental Monitoring in the Vicinity of the Savannah River Plant. Annual Report, 

1981. DPSPU-82-30-1. 

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1982. DPSPU-83-30-1. 

Du Pont. 1984. Environmental Monitoring in the Vicinity of the Savannah River Plant. Annual Report, 

1983. DPSPU-84-30-1. 
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Semiannual Report, July through December 1958. DPSPU-59-11-23. 

Horton, J.H. 1954. Radioactivity in the Environs of the Savannah River Plant, January to July 1954. 

DP-92. 

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Horton, J.H. and H.G. Mealing, Jr. 1956. Health Physics Regional Monitoring Semiannual Report, July 

through December 1955. DPSP-56-25-54. 

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June 1956. DPSP-57-25-4.

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December 1956. DPSP-57-25-15. 

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January through June 1957. DPSP-57-25-43. 

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Semiannual Report, July through December 1957. DPSP-58-25-17. 

Mikol, S.C., L.T. Burckhalter, J.L. Todd, and D.K. Martin. 1988. Savannah River Plant Environmental 

Report, Annual Report for 1987. Volume I, Text. DPSPU-88-30-1. Health Protection 

Department, Savannah River Plant. 

Zeigler, C.C., I.B. Lawrimore, and W.E. O’Rear. 1985. Environmental Monitoring at the Savannah 

River Plant, Annual Report 1984. DPSPU-85-302. Health Protection Department, Savannah 

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Report, Annual Report for 1985. Volume I, Text. DPSPU-86-30-1. Health Protection 

Department, Savannah River Plant. SRS Phase II Database EIW1993082022.

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1985. Volume II, Figures and Data Tables. DPSPU-86-30-1. Savannah River Plant. 

Zeigler, C.C., E.M Heath, L.B. Taus, J.L. Todd, and J.E. Till. 1987. Savannah River Plant 

Environmental Report, Annual Report for 1986. Volume I, Text, Volume II, Figures and Data 

Tables. DPSPU-87-30-1. Health Protection Department, Savannah River Plant. 
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GLOSSARY

“A,” “B,” “C,” and “D”

Codes used in whole body count results; the letters represent energy regions.

“A” means 810-910keV, “B” means 920-1020 keV, “C” means 1180-1290 keV, and “D” means 

1760-1990 keV.

Absorption type

A term introduced in ICRP’s revised respiratory tract model (ICRP-66) to replace the clearance 

classes in ICRP-30. The absorption types are Type F for fast absorption, formerly called Class 

D; Type M for moderate absorption, formerly called Class W; and Type S for slow absorption, 

formerly called Class Y. Also referred to as Type. 

albedo dosimeter

A dosimeter that measures the slow neutrons which are generated by higher energy neutrons 

incident on the body and which reflect back into the dosimeter. 

annual dose equivalent

The dose equivalent received in a year. The annual dose equivalent is expressed in units of 

rem (sievert).

Atomic Energy Commission

Original agency established for nuclear weapons and power production; a predecessor to the 

U.S. Department of Energy.

beta (β) dose

A designation (i.e., beta) on some SRS external dose records referring to the dose from lessenergetic beta, x-ray and/or gamma radiation (see ALSO open window, or shallow dose). 

beta radiation

Radiation consisting of charged particles of very small mass (i.e., the electron) emitted 

spontaneously from the nuclei of certain radioactive elements. Most (if not all) of the direct 

fission products emit beta radiation. Physically, the beta particle is identical to an electron 

moving at high velocity.

class

The respiratory tract inhalation classification scheme developed in ICRP-30 for inhaled 

material according to its rate of clearance from the pulmonary region of the respiratory tract. 

Materials are classified as D (days), W (weeks), or Y (years), according to how fast they clear 

the lungs; Class D in less than 10 days; Class W in 10 to 100 days; Class Y in more than 100 

days. Recent recommendations in ICRP Report 66 replaced classes D, W, and Y with lung 

absorption Types F (fast), M (moderate), and S (slow).

collective dose equivalent

The sum of the dose equivalents of all individuals in an exposed population. Collective dose is 

expressed in units of person-rem (person-sievert).

collective effective dose equivalent

The sum of the effective dose equivalents of all individuals in an exposed population. 

Collective effective dose is expressed in units of person-rem (person-sievert).
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curie

A special unit of activity. One curie exactly equals 3.7 × 1010 nuclear transitions per second.

deep absorbed dose (Dd)

The absorbed dose at the depth of 1.0 cm in a material of specified geometry and 

composition.

deep dose equivalent (Hd)

The dose equivalent at the respective depth of 1.0 cm in tissue.

detection limit (lower)

The minimum quantifiable exposure or neutron flux that can be detected.

delayed neutron analysis (DNA)

The method was adopted for both enriched and depleted uranium urinalysis analyses in about 

1982.

depleted uranium (DU) analysis

Indicates that the chemical properties of uranium were used for analysis, not necessarily that 

the material in question was depleted uranium

dose equivalent (H)

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The special unit is the rem. When D is expressed in Gy, H is in Sieverts (Sv). 

(1 Sv = 100 rem.)

dose of record

The dose files provided by DOE to NIOSH as part of the individual worker files. 

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. (see albedo dosimeter, film 

dosimeter, neutron film dosimeter, thermoluminescent dosimeter.)

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external and/or internal sources of radiation. 

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, 

and/or extremities. This includes the fabrication, assignment, and processing of the 

dosimeters as well as interpretation and documentation of the results.

DU

As used in this document depleted uranium, used as targets for 239Pu production; isotopic 

activity ratios (NOTE: this is not the mass ratio) listed by SRS as:
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Isotope

234U

Activity fraction

0.0840

235U 0.0145

238U 0.9015

enriched uranium (EU) analysis

As used in this document indicates the analysis was by alpha counting, not necessarily that 

the material in question was enriched uranium

exchange period (frequency)

Time period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

As used in the technical sense, exposure refers to a measure expressed in roentgens (R) of 

the ionization produced by photon radiation (i.e., gamma and x-rays) in air. 

extremity

That portion of the arm extending from and including the elbow through the fingertips, and that 

portion of the leg extending from and including the knee and patella through the tips of the 

toes.

FB-Line

The plutonium finishing line located in the 221-F facility.

Field (f)

Refers to SRS practice to quantify Neutron Track Emulsion Type A (NTA) neutron tracks using 

a projection image covering a 15-inch projection screen. 

field calibration

Dosimeter calibration based on radiation types, intensity and energies present in the work 

environment.

film

Generally means a "film packet" that contains one or more pieces of film in a light-tight 

wrapping. The film when developed has an image caused by radiation that can be measured 

using an optical densitometer. (see Dupont 552, Dupont 558, Eastman Kodak, Nuclear 

Emulsions.)

film density

See optical density.

film dosimeter

A small packet of film within a holder that attaches to a wearer.

flux density (n/cm2

-sec)

A measure of the intensity of neutron radiation in neutrons/cm2

-sec. It is the number of 

neutrons passing through 1 square centimeter of a given target in 1 second. Expressed as nv, 

where n = the number of neutrons per cubic centimeter and v = their velocity in centimeters 

per second. 
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gamma rays (G or γ)

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Physically, gamma 

rays are identical to x-rays of high energy, the only essential difference being that x-rays do 

not originate in the nucleus. 

Gray (Gy)

The special name for the SI unit of absorbed dose. (1 Gy = 1 Jkg-1

)

HB-Line

The plutonium finishing line located in the 221-H facility.

HEU

Highly enriched uranium, used in Naval Reactor Fuel; isotopic ratios listed by SRS as:

Isotope

234U

Activity fraction

0.9806

235U 0.0194

238U 0.0000

Health Protection Radiation Exposure Database (HPRED)

This database replaced the mainframe tape master file system in April 1989. This system 

interfaced with the TLD badge reader system. In 1990, visitor records were included in this 

database.

Health Protection Annual Radiation Exposure History Database (HPAREH)

Starting in 1980, this database was used to generate yearly radiation exposure reports to 

employees.

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

liquid scintillation counter (LSC)

Used for tritium urinalysis starting in 1958.

maximum permissible body burden (MPBB)

Limit for body content of a radionuclide used from startup until 1975

maximum permissible lung burden (MPLB)

During the 1970's the occupational limit for Pu was expressed in terms of a quantity of 

plutonium that could be present in the chest at any given time, equal to 16 nCi (0.25 μg) 239Pu.

maximum permissible organ burden (MPOB)

Limit for organ content of a radionuclide used from startup until 1975.

neutron

A basic particle that is electrically neutral, having nearly the same mass as the hydrogen atom.

neutron, fast

Neutrons with energy equal or greater than 10 keV.
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neutron, intermediate

Neutrons with energy between 0.5 eV and 10 keV.

neutron, thermal

Strictly, neutrons in thermal equilibrium with surroundings. Generally, neutrons with energy 

less than about 0.5 eV.

neutron film dosimeter

A film dosimeter that contains an Neutron Track Emulsion, type A, film packet.

nuclear emulsion

Often referred to as “NTA” film and used to measure personnel dose from neutron radiation.

Nuclear Track Emulsion, Type A (NTA)

A film that is sensitive to fast neutrons. The developed image has tracks caused by neutrons 

that can be seen by using an appropriate imaging capability such as oil immersion and a 

1000X power microscope or a projection capability.

open window (OW)

Designation on SRS film dosimeter reports that implies the use of little (i.e., only security 

credential) shielding. It commonly is used to label the film response corresponding to the open 

window area. 

operating area

Designation of SRS major operational work areas among the respective fuel fabrication (300), 

reactor operations (R, P, L, K and C), chemical separations (F and H), uranium separation (ALine), plutonium finishing (H-Canyon, B-Line), research and development, and transportation, 

communication and general site support.

optical density

The quantitative measurement of photographic blackening; the optical density is defined as D 

= Log10 (Io/I).

Parameter 1

The column in the IREP template where the dose reconstructor will enter the calculated dose. 

Multiple entries based on year of employment, type of radiation and appropriate energy 

ranges, internal and external exposure are possible.

Parameter 2

The column in the IREP template where the dose reconstructor will enter the standard 

deviation of the mean of the organ dose.

PuF4 source

A neutron source whose activating material is plutonium fluoride. The source was used to 

duplicate the neutron energies in SRS plutonium handling facilities.

pencil dosimeters

A type of ionization chamber used by personnel to measure radiation dose. These results 

may be labeled as “Pen” dose. Other names

Pencil, pocket dosimeter, pocket pencil, pocket ionization chamber (PIC).
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personal dose equivalent Hp(d)

Represents the dose equivalent in soft tissue below a specified point on the body at an 

appropriate depth d. The depths selected for personnel dosimetry are 0.07 mm and 10 mm, 

respectively, for the skin and body. These are noted as Hp(0.07) and Hp(10), respectively. 

photon

A unit or "particle" of electromagnetic radiation consisting of x- and/or gamma rays. 

photon - x-ray

Electromagnetic radiation of energies between 10 keV and 100 keV whose source can be an 

x-ray machine or radionuclide.

quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose.

Rad

The traditional unit of absorbed dose (one rad = 100 ergs per gram of material absorbing the 

radiation energy).

Radiation

Alpha, beta, neutron, and photon radiation.

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

Radiological Qualification Badge (RQB)

System starting in 1992 to control each worker’s dosimetry requirements including bioassay 

requirements. Each radiation worker carries his/her personal card. From 1992 to 1999 the 

card listed requirements that had been met by the worker prospectively prior to entering 

radiation areas. That is, the card requirements and the Radiation Work Permit requirements 

had to match prior to entry. From 1999 to present the requirements on the card are listed 

retrospectively based on entries made during the designated period. Prior to the RQB system, 

an electronic database was used to track worker dosimetry requirements. 

relative biological effectiveness (RBE)

A ratio of the absorbed dose of a reference radiation to the absorbed dose of a test radiation 

producing the same biological effects, other conditions being equal.

rem

The rem is the traditional unit of dose equivalent, which is equal to the product of the absorbed 

dose in rad and the quality factor of the radiation.

rep

A historical unit, roentgen-equivalent-physical (mrep = millirep) used when reporting beta 

exposures, usually recorded in mrep; equivalent to 93 ergs of energy per gram. 

Roentgen (R or r)

A unit of exposure to gamma (or x-ray) radiation. It is defined precisely as the quantity of 

gamma (or x) rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 kg of dry air 
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STP. An exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft 

tissue for higher (~>100 keV) energy photons.

RU

Recycled uranium, is SRS reactor fuel; isotopic activity ratios listed by SRS as:

Isotope

234U

Activity fraction

0.8489

235U 0.0120

236U 0.1388

238U 0.0003

shallow absorbed dose (Ds)

The absorbed dose at a depth of 0.07 mm in a material of specified geometry and 

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 mm in tissue.

shielding

Any material or obstruction that absorbs (or attenuates) radiation and thus tends to protect 

personnel or materials from radiation.

shield

Designation on SRS dose records that implies a radiation dose, typically to the whole body, 

from higher energy photon radiation.

sievert (Sv)

The SI unit for dose equivalent. (1 Sv = 100 rem.)

silver shield(s)

The 1-mm thick shields covering the film packet in the early SRS personnel film dosimeters. 

skin dose

Absorbed dose at a tissue depth of 7 mg/cm2 ~ 0.07 mm in tissue.

Snoopy

Portable neutron monitoring instrument with a moderated BF3 detector.

SP

This historical designation stands for “Special Product” and referred to 237Np.

thermoluminescence

Property of a material that causes it to emit light as a result of being excited by heat.

thermoluminescent dosimeter (TLD)

A device used to measure radiation dose. It is comprised of a holder containing solid chips of 

material that when heated will release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose. 
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3% plutonium, 6% plutonium, and 12% plutonium

Designations for typical mixtures of plutonium encountered at SRS, where the percentage 

represents the percent by mass of 240Pu in the mixture.

whole body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1000 mg/cm2

); however, 

this term is also used to refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin or a radiograph
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ATTACHMENT A

DESCRIPTION OF FACILITIES AND PROCESSES

TABLE OF CONTENTS

Section

A.1 Reactors ....................................................................................................................................................

Page

140

A.2 F-Area A-Line Facility ................................................................................................................................ 140

A.3 221-F B-Line Facility.................................................................................................................................. 140

A.4 221-F Canyon Facility................................................................................................................................ 140

A.5 New Special Recovery Facility .................................................................................................................. 141

A.6 Area Outside Facilities............................................................................................................................... 141

A.7 238PuO2 Fuel Form Facility (PuFF) and the 238PuO2 Experimental Facility (PEF)..................................... 146

A.8 235-F Vaults .............................................................................................................................................. 146

A.9 A-9 772-F and 772-1F Production Control Laboratories ........................................................................... 146

A.10 E-Area Solid Waste Disposal Facility (SWDF).......................................................................................... 147

A.11 Plutonium Storage Facility (PSF) .............................................................................................................. 147

A.12 F/H Effluent Treatment Facility (ETF)........................................................................................................ 147

A.13 F & H Cooling Water and Retention Basins .............................................................................................. 148

A.14 F & H Area Tank Farms............................................................................................................................. 148

A.15 Waste Certification Facility (WCF)............................................................................................................. 148

A.16 221-H B-Line Facility ................................................................................................................................. 150

A.17 H-Canyon Facility ...................................................................................................................................... 150

A.18 211-H Area Outside Facilities................................................................................................................... 153

A.19 Receiving Basin for Off-Site Fuel (RBOF) and Resin Regeneration Facility (RRF).................................. 153

A.20 Tritium Facilities......................................................................................................................................... 153

A.21 M-Area ....................................................................................................................................................... 153

A.22 320-M, 322-M, and 341-M ......................................................................................................................... 155

A.23 Fuel Fabrication Facility, 321-M ................................................................................................................ 155

A.24 S-Area- Defense Waste Processing Facility (DWPF) ............................................................................... 155

A.25 Z Area ........................................................................................................................................................ 156

A.26 A Area Building 773-A ............................................................................................................................... 156

A.27 Buildings 735-A and 735-11A.................................................................................................................... 158

A.28 Building 776-A Liquid Waste Handling Facility.......................................................................................... 158

A.29 D-Area Heavy Water Production and Reprocessing................................................................................. 158

A.30 CMX and TNX (TNX Semi-works)............................................................................................................. 158
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LIST OF TABLES

A-1.1 Reactors years of operation...................................................................................... 140

A-1.2 Radionuclides of concern for all reactors .................................................................. 140

A-2 F Area A-Line Facility ............................................................................................... 141

A-3 221-F B-Line Facility................................................................................................. 141

A-4.1 221-F Canyon Facility............................................................................................... 142

A-4.2 221-F Canyon Facility............................................................................................... 142

A-4.3 221-F Canyon Facility............................................................................................... 143

A-5 New Special Recovery (NSR) Facility....................................................................... 143

A-6.1 F-Area Outside Facilities .......................................................................................... 144

A-6.2 F Area Outside Facilities........................................................................................... 144

A-6.3 F Area Outside Facilities........................................................................................... 144

A-6.4 F Area Outside Facilities........................................................................................... 145

A-6.5 F Area Outside Facilities........................................................................................... 145

A-6.6 F Area Outside Facilities........................................................................................... 146

A-7 238PuO2 Fuel Form Facility (PuFF) and 238PuO2 Experimental Facility (PEF) ............ 146

A-8 235-F Vaults............................................................................................................. 147

A-9 772-F and 772-1F production control laboratories .................................................... 147

A-10 E-Area Solid Waste Disposal Facility (SWDF).......................................................... 147

A-11 Plutonium Storage Facility ........................................................................................ 148

A-12 F/H Effluent Treatment Facility ................................................................................. 148

A-13 F & H Area—cooling water and retention basins ...................................................... 149

A-14 F & H Area Tank Farms............................................................................................ 149

A-15 Waste Certification Facility ....................................................................................... 149

A-16.1 221-H B-Line Facility ................................................................................................ 150

A-16.2 221-H B-Line Facility ................................................................................................ 150

A-16.3 221-H B-Line Facility, Np02 and Pu02 Facility ........................................................... 150

A-17.1 H-Canyon Facility ..................................................................................................... 151

A-17.2 H-Canyon Facility ..................................................................................................... 151

A-17.3 H-Canyon Facility ..................................................................................................... 152

A-17.4 H-Canyon Facility ..................................................................................................... 152

A-18.1 H-Area Outside Facilities .......................................................................................... 153

A-18.2 H-Area Outside Facilities .......................................................................................... 153

A-19.1 Receiving Basin for Off-site Fuel (RBOF) ................................................................. 154

A-19.2 Resin Regeneration Facility...................................................................................... 154

A-20 Tritium Facilities........................................................................................................ 154

A-21 Uranium Target Fabrication Facility, 313 M .............................................................. 154

A-22 Uranium Processing Facilities 320-M, 322-M, and 341-M......................................... 155

A-23 Fuel Fabrication Facility, 321-M................................................................................ 155

A-24.1 S Area Defense Waste Processing Facility............................................................... 156

A-24.2 S Area Defense Waste Processing Facility............................................................... 156

A-25.1 Z Area ...................................................................................................................... 157

A-25.2 Z Area ...................................................................................................................... 157
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The primary operational areas and production facilities are discussed below.

A.1 Reactors

Five heavy-water reactors designated C, K, L, P, and R were constructed at SRS in the early 1950s, 

located in C Area, K Area, L Area, P Area, and R Area, respectively. The years of operation and 

radionuclides are shown in Table A-1.1 and A-1.2 below.

Table A-1.1. Reactors years of operation.

Reactor C K L P R

Years of operation 1955 to1985 1954 to1988 1954 to1968

1985 to1988

1954 to1988 12/28/1953 to 

06/1964

Table A-1.2. Radionuclides of 

concern for all reactors.

a

241Am 60Co 131I 103Ru

140Ba 51Cr 133I 106Ru

141Ce 134Cs 140La 90Sr / 90Y

143Ce 137Cs 54Mn 91Sr

144Ce 152Eu 95Nb 92Sr

242Cm 154Eu 239Np 234U

244Cm 59Fe 238Pu 65Zn

58Co 3

H b 239Pu 95Zr

a. The composition and relative 

abundance are not known.

b. Tritium is assumed to produce most of 

the personnel exposure because of the 

large quantities present.

A.2 F-Area A-Line Facility

The A-Line facility accepts dilute, depleted uranyl nitrate solution from the 221-F Facility and converts 

it into uranium trioxide powder. Dilute uranyl nitrate solution is sent to A-Line from the solvent 

extraction process in the 221-F Facility. The A-Line facility also has a dissolver to prepare a uranyl 

nitrate solution from the uranium trioxide product and nitric acid when it is required as a process feed 

stream back into the solvent extraction process in the 221-F Facility. 

A.3 221-F B-Line Facility

The F B-Line Facility, located in the 200-F Separations Area, is used to convert plutonium nitrate into 

plutonium metal or plutonium oxide and can recover plutonium from on-site and off-site scrap. 

A.4 221-F Canyon Facility

The 221-F Canyon Facility consists of two parallel canyons that constitute the process area. These 

canyons contain the high-activity and low-activity materials and are referred to as the hot and warm 

canyons respectively. 
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Table A-2. F Area A-Line Facility.

Facility/process Conversion of dilute depleted uranyl nitrate 

solution (A-line feed) to uranium trioxide powdera,b

Years of operation 1954-1987; 1993-1997

Radionuclides

of concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

239Pu M 5.559E-03 X

235U Fc 8.022E-04 X

236U Fc 8.179E-04 X

238U Fc 7.440E-01 X

141Ce M 1.305E-02 X

144Ce M 6.079E-02 X

103Ru Sd 5.969E-02 X

106Ru Sd 6.111E-02 X

95Nb Sd 5.418E-02 X

a. Dosimetrically, the feed material for A-Line is 238U and 239Pu.

b. For purpose of calculating internal dose, specific activity of the uranium is 

approximately 0.4 nCi/mg.

c. Uranyl nitrate is type F; a solubility study on UO3 powder at Hanford showed it to 

be mostly type F with a small component of type M; assume 80% F, 20% M.

d. In the nitrate solution the ruthenium would be type F; after the oxidation process 

or after years of aging as contamination, the ruthenium would be type S. Type S 

will maximize the lung dose; type F maximizes the dose to all other tissues. 

Same statements apply to niobium except it is type M in the nitrate form.

Table A-3. 221-F B-Line Facility.

Facility/process Conversion of plutonium nitrate into plutonium metal or plutonium oxide

Years of operation 1954-1989

6% mixturea,b 12% mixturea,c

Radionuclides

of concern

Absorption 

types

Activity 

fraction

Activity 

fraction

Significant to 

external exposure

238Pu M-Sd 4.800E-02 2.660E-01 X

239Pu M-Sd 7.760E-01 4.660E-01 X

240Pu M-Sd 1.760E-01 2.680E-01 X

a. The percent refers to the mass percent of 240Pu.

b. The ratio of 241Pu to alpha Pu is 8 to 1.

c. The ratio of 241Pu to alpha Pu is 23 to 1.

d. Pu nitrate is type M; Pu oxides are type S.

A.5 New Special Recovery Facility

Scrap plutonium from offsite sources and from FB-Line facility is processed through the New Special 

Recovery (NSR) Facility. The processed plutonium is conditioned so that it can be blended back into 

the plutonium nitrate feed stream to the FB-Line process.

A.6 Area Outside Facilities

The F-Area Outside Facilities provide general support, principally to the processing of irradiated fuels 

and targets in Building 221-F. The term “Outside Facilities” is used to describe a wide variety of 

processes, utilities, and services that are ancillary to the primary 200-F Area operations. 
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Table A-4.1. 221-F Canyon Facility.

Facility/process Head End Stream (Stream 1) Recovery of 239Pu, 

237Np and 238Ua 

Years of operation 11/1954-1989; 1993-2002

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to

external exposureb

239Pu M 1.191E-03 X

144Ce M 4.765E-01 X

137Cs F 2.621E-02 X

106Ru F 1.588E-01 X

95Zr M 3.177E-01 X

90Sr F 1.959E-02 X

a. The function of the 221-F Separation Facility is to isolate and purify plutonium, 

uranium, and neptunium from irradiated uranium. Fission products will also be 

present and, depending on the age of the material, the radionuclides of concern 

and the fraction of the activity of the materials will change. The majority of the 

internal dose from the Head End stream comes from plutonium, cerium, and 

ruthenium.

b. The majority of the external dose from the Head End stream comes from cerium, 

zirconium, and ruthenium.

Table A-4.2. 221-F Canyon Facility.

Facility/process Second Uranium Cycle (Stream 2)a

Years of operation 11/1954-1989; 1993-2002

Radionuclides of 

concern

Absorption 

types

Activity 

fraction

Significant to

external exposureb 

239Pu M 4.226E-03

235U M 7.433E-04

238U M 6.893E-01 X

144Ce M 5.633E-02 X

137Cs F 3.133E-03 X

106Ru F 5.662E-02 X

95Zr M 1.873E-01 X

90Sr F 2.332E-03 X

a. The function of the 221-F Separation Facility is to isolate and purify plutonium, 

uranium, and neptunium from irradiated uranium. Fission products will also be 

present and, depending on the age of the material, the radionuclides of concern 

and the fraction of the activity of the materials will change. The majority of the 

internal dose from the second uranium cycle (Stream 2) comes from 238U and 

239Pu as indicated above. It can be assumed this stream consists of depleted

uranium with a specific activity of 0.4 nCi/mg and 239Pu (La Bone, 1996).

b. The majority of the external shallow dose from the second uranium cycle (Stream 

2) comes from 238U which feeds 234mPa, and zirconium which emits both 

photons and betas.
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Table A-4.3. 221-F Canyon Facility.

Facility/process Second Cycle Plutonium Stream (Stream 3)a

Years of operation 11/1954-1989; 1993-2002

Radionuclides of 

concern

Absorption 

types

Activity 

fraction

Significant to 

external exposureb

238Pu M 2.045E-04

239Pu M 6.075E-02

241Pu M 9.194E-01 X

144Ce M 2.424E-03 X

106Ru F 9.204E-04 X

95Zr M 1.616E-02 X

a. The function of the 221-F Separation Facility is to isolate and purify plutonium, 

uranium, and neptunium from irradiated uranium. Fission products will also be 

present, and, depending on the age of the material, the radionuclides of concern 

and the fraction of the activity of the materials will change. 

b. The majority of the external dose comes from 241Pu and 95Zr.

Table A-5. New Special Recovery (NSR) Facility.

Facility/process New Special Recovery (Pu Recovery)

Years of operation

Radionuclides of 

concern

Absorption 

types

Activity 

fraction

Significant to 

external exposureb

241Am M

239Pu M To be determined

240Pu M

235U F

236U F

Examples of feed material for NSR Facility

Percent of isotopic content

Material 239Pu 240Pu 241Am Other

Fe, Ni, Cu, U

Impure Plutonium Oxide 87 to 94 6 to 12 0 to 1 -

Plutonium Metal 88 to 94 6 to 12 - -

Unirradiated Reactor Cores - 4 to 42 0 to 10 60 to 80 U

Scrap MgO Crucibles 88 to 94 6 to 12 0 to 1 < 90 MgO

Mixed Oxides and Metals c c - 3 Mo

ZPPR Pu and U Oxide Plates 2 to 88 6 to 18 - 78 U, 2 Mo

a. The majority of the internal dose is delivered by 239Pu, 238Pu, 242Am, 235U, and 236U (assuming 

enriched uranium).

b. The radionuclide that delivers the majority of the external dose is 241Am. Some neutron dose will 

also be present due to spontaneous fission and an (alpha, n) reaction with fluoride, if present.

c. Trace plutonium in uranium oxides.
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Table A-6.1. F-Area Outside Facilities.

Facility/process Acid Recovery Unit (ARU)

Years of operation 1954-1989

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

234U M 7.040E-11

235U M 1.012E-09

236U M 1.008E-09

238U M 1.008E-07

103Ru F 6.950E-02 X

106Ru F 1.023E-01 X

95Zr M 6.176E-01 X

95Nb M 2.107E-01 X

Table A-6.2. F Area Outside Facilities.

Facility/process Water Handling System

Years of operation 1954-1989

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

103Ru S 1.417E-01 X

106Ru S 4.815E-01 X

95Zr S 1.298E-01 X

95Nb S 2.470E-01 X

Table A-6.3. F Area Outside Facilities.

Facility/process General Purpose Evaporators

Years of operation 1954-1989

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposureb

238Pu S 9.626E-05

239Pu S 2.903E-02

241Pu S 3.913E-01 X

234U S 2.911E-08

235U S 4.813E-07

236U S 4.343E-07

238U S 9.343E-05 X

103Ru S 4.852E-02 X

106Ru S 3.369E-01 X

95Zr S 9.743E-02 X

95Nb S 9.665E-02 X

a. The radionuclide of concern for internal exposure for the GP Evaporators, 

Segregated Plutonium Solvent, and Laboratory Waste is plutonium. Type M (type 

F for ruthenium) should be used for nitrate form or from solvent extraction. 

b. The radionuclides of external dosimetric concern for the GP Evaporators, 

Segregated Plutonium Solvent, and Laboratory Waste are as indicated above. 

238U feeds 234Pa with a high energy beta which is also of concern.
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Table A-6.4. F Area Outside Facilities.

Facility/process Segregated Plutonium Solvent

Years of operation 1954-1989

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposureb

238Pu M 2.068E-04

239Pu M 5.941E-02

241Pu M 8.701E-01 X

103Ru F 2.418E-03 X

106Ru F 5.941E-02 X

95Zr M 5.941E-03 X

95Nb M 2.532E-03 X

a. The radionuclide of concern for internal exposure at the GP Evaporators, 

Segregated Plutonium Solvent, and Laboratory Waste is plutonium.

b. The radionuclides of external dosimetric concern for the GP Evaporators, 

Segregated Plutonium Solvent, and Laboratory Waste are as indicated above. 

238U feeds 234Pa with a high energy beta which is also of concern.

Table A-6.5. F Area Outside Facilities.

Facility/process Laboratory Waste

Years of operation 1954-1989

Radionuclides of 

concern 

Absorption 

typesa

Activity 

fraction

Significant to 

external exposureb

238Pu S 6.150E-01

239Pu S 1.921E-02

241Pu S 3.456E-01 X

234U S 4.803E-08

235U S 7.497E-07

236U S 7.209E-07

238U S 7.497E-05 X

141Ce M 1.369E-04 X

144Ce M 9.892E-03 X

134Cs F 1.822E-04 X

137Cs F 1.590E-03 X

103Ru S 1.369E-04 X

106Ru S 8.645E-04 X

95Zr M 6.150E-03 X

95Nb S 1.237E-03 X

a. The radionuclide of concern for internal exposure at the GP Evaporators, 

Segregated Plutonium Solvent, and Laboratory Waste is plutonium. Various 

solubility classes are possible; those listed can be used as defaults. Nitrate forms 

would have been more soluble.

b. The radionuclides of external dosimetric concern for the GP Evaporators, 

Segregated Plutonium Solvent, and Laboratory Waste are as indicated above. 

238U feeds 234Pa with a high energy beta which is also of concern.
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Table A-6.6. F Area Outside Facilities.

Facility/process Segregated Uranium Solvent

Years of operation 1954-1989

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

234U M 3.371E-07

235U M 5.941E-06

236U M 5.775E-06

238U M 5.775E-04 X

103Ru F 1.491E-01 X

106Ru F 6.417E-01 X

95Zr M 1.491E-01 X

95Nb S 5.941E-02 X

A.7 238PuO2 Fuel Form Facility (PuFF) and the 238PuO2 Experimental Facility (PEF)

The primary function of the PuFF facility is to produce encapsulated plutonium oxide fuel forms. The 

conversion of plutonium oxide powder into dense PuO2 fuel forms is conducted in the PEF.

Table A-7. 238PuO2 Fuel Form Facility (PuFF) and 238PuO2 Experimental 

Facility (PEF).

Facility/process Production of encapsulated plutonium oxide fuel forms

Years of operation 1978 -??

Radionuclides 

of concerna

Absorption 

types

Activity 

fraction

Specific 

activity 

(ci/g)

Significant to 

external exposure

241Am Sb 3.067E-04 3.21 X

238Pu S 8.535E-01 1.66E+1 X

239Pu S 1.411E-01 6.13E-2

240Pu S 2.27E-1

241Pu S 4.089E-03 1.12E+2

242Pu S 3.85E-3

237Np M 5.111E-04 6.87E-4 X

232Th M 5.111E-04 1.11E-7 X

Particle Size 

Mass Median Diameter

Ball-milled Powder 2 μm

Feed Powder 5 μm

a. The majority of the dose comes from plutonium.

b. Type M if mostly pure; type S if a contaminant in a plutonium matrix.

A.8 235-F Vaults

The 235-F facility receives plutonium and plutonium dioxide and fabricates it into a variety of items 

used in the nuclear program. There are three storage vaults located in the 235-F facility, the 

Plutonium Oxide Vault, Finished Product Vault, and Scrap Vault.

A.9 A-9772-F and 772-1F Production Control Laboratories

The production control laboratories provide analytical support for the 200-F and 200-H separations 

processes. 
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Table A-8. 235-F Vaults.

Facility/process Plutonium Oxide Vault; Finished Product Vault, 

Scrap Vaulta

Years of operation

Radionuclides

of concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

235U S

238Pu S

239Pu S

237Np M

a. Detailed information on the composition of the radionuclides stored and found in 

these vaults is not available; therefore, they are considered to be pure.

Table A-9. 772-F and 772-1F production control laboratories

Facility/process 772-F & 772-1F Production Control Laboratoriesa

Years of operation

Radionuclides of 

concern

Absorption 

types

Activity 

fraction

Significant to 

external exposure

Pu M

U M

Fission products

a. Specific fission products or Pu and U information is not specified.

A.10 E-Area Solid Waste Disposal Facility (SWDF)

Solid and liquid low-level waste generated at the SRS and the Savannah River Laboratory (SRL) and 

some waste from off-site are stored at the SWDF.

Table A-10. E-Area Solid Waste Disposal Facility (SWDF).

Facility/process Liquid and Solid Waste Disposala

Years of operation

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

a. Due to the diverse nature of the radioactive waste, the radionuclide content of 

each type of waste and absorption types has not been defined. The radionuclides 

of highest concentration (Ci/vol) are tritium, cobalt, plutonium, and curium.

A.11 Plutonium Storage Facility (PSF)

The Plutonium Storage Facility (PSF) is located in the 200-F Separations Area and is used to receive, 

store, monitor, retrieve, and ship packaged plutonium, as needed. 

A.12 F/H Effluent Treatment Facility (ETF)

The F/H Effluent Treatment Facility (ETF) consists of large impermeable storage basins with lift 

stations located in both F and H Areas and a treatment facility, Building 241-84H located in H Area. 

ETF is designed to remove hazardous chemical and radioactive contaminants from the 200-Area 

liquid effluent waste streams and concentrate them for disposal. 
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Table A-11. Plutonium Storage Facility (PSF).

Facility/process Receipt, store, monitor, retrieve and ship plutonium 

Years of operation

Radionuclides of 

concern 

Absorption 

types

Activity fractiona b d

@ 6% Pu button 

Activity fractiona c d

@12% Pu button

Significant to 

external exposure

238Pu S 4.800E-02 2.660E-01

239Pu S 7.760E-01 4.660E-01

240Pu S 1.760E-01 2.680E-01

a. The percent refers to the mass percent of 240Pu.

b. The ratio of 241Pu to alpha Pu is 8.0 to 1.

c. The ratio of 241Pu to alpha Pu is 23 to 1.

d. The activity fractions of alpha-emitting plutonium (alpha-Pu) isotopes are those found in typical 6% or 12% plutonium 

buttons from the 221-F B Line.

Table A-12. F/H Effluent Treatment Facility (ETF).

Facility/process Removal and concentration of waste from liquid 

waste streams 

Years of operation 1988- ??

Radionuclides of 

concern

Absorption 

types

Activity 

fraction

Significant to 

external exposure

238Pu S (Y)a 2.855E-04

239Pu S (Y)a 2.515E-06

241Pu S (Y)a 8.401E-04

144Ce M (W)a 4.690E-01 X

134Cs F 1.340E-02 X

137Cs F 2.723E-02 X

106Ru S 7.763E-02 X

95Zr M 2.382E-01 X

89Sr F 1.601E-01 X

90Sr F 1.335E-02 X

a. Lung Solubility Classes were chosen to maximize the missed committed dose.

A.13 F & H Cooling Water and Retention Basins

The F and H Areas consist of chemical separation plants and appropriate support facilities for the 

production of purified radionuclides, principally uranium, plutonium, and tritium. The cooling water and 

retention basins are support facilities for the separations and waste management areas. Low-level 

radioactive liquid wastes from the separations areas are routed to the basins when activity is 

detected.

A.14 F & H Area Tank Farms

High-level liquid radioactive wastes from SRS are received and managed in large underground tanks 

in the waste tank farms in 241-F and 241-H. The two farms contain 51 large subsurface tanks and 

related facilities required for safe handling, processing, and temporary retention of liquid wastes. 

A.15 Waste Certification Facility (WCF)

The Waste Certification Facility (WCF), Building 724-8E, located in the 643-E Solid Waste Disposal 

Facility, is used to certify and package drums of transuranic (TRU) waste. The functions of the WCF 

are to assay and certify the contents of TRU waste drums and to package, load, and ship the drums.
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Table A-13. F & H Area—cooling water and retention basins.

Facility/process F- Area 

Retention Basinsa H - Area Retention Basinsa

Years of operation 1955-1989

Radionuclides 

of concern 

Absorption 

types Activity fraction Activity fraction

Significant to 

external exposure b 

238Pu S 7.007E-03 1.403E-02

239Pu S 2.803E-02 1.303E-02

238U S 6.206E-02 9.018E-03 X

144Ce S 1.001E-03 1.403E-02 X

137Cs F 7.047E-01 7.335E-01 X

106Ru S 7.508E-02 2.405E-02 X

90Sr F 1.221E-01 1.924E-01 X

a. The majority of the annual internal dose from the F-Area and H-Area Basins comes from plutonium and 

uranium.

b. External exposure is for both F and H Area Retention Basins.

Table A-14. F & H Area Tank Farms

Facility/process F-Area combined 

tank waste

H-Area combined

tank waste

Years of operation 1955-1989

Radionuclides 

of concern 

Absorption 

types Activity fraction Activity fraction

Significant to 

external exposureb

241Am M 1.319E-04 7.346E-05 X

244Cm M 9.341E-04 6.121E-06

238Pu M 2.198E-05 5.632E-03 X c

239Pu M 1.099E-04 - 241Pu M 1.022E-02 5.693E-03 X

144Ce M 1.538E-01 2.632E-01 X

137Cs F 4.286E-01 3.489E-01 X

106Ru S 1.055E-02 2.142E-02 X

90Sr F 3.956E-01 3.550E-01 X

a. When unknown or variable absorption types were chosen to maximize the missed dose.

b. External exposure is for both F and H Area Combined Tank Waste.

c. For the F-Area Combined Tank Waste, 238Pu is not a radionuclide of concern for external exposure.

Table A-15. Waste Certification Facility (WCF).

Facility/process Building 724-8E Certification and Packaging of 

TRU waste

Years of operation

Radionuclides 

of concern a

Absorption 

types

Activity 

fraction

Significant to 

external exposure

a. Radionuclides are considered to be the same as for the Solid Waste Disposal 

Facility (643-E). The radionuclides in the solid waste disposal facility have not 

been identified. It is assumed the radionuclides are tritium, curium, plutonium, 

and activation products.
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A.16 221-H B-Line Facility

The HB Line consists of three separate facilities; the

• Scrap Recovery Facility

• Neptunium Oxide Facility, and 

• Plutonium Oxide Facility.

Table A-16.1. 221-H B-Line Facility

Facility/process Scrap Recovery 238U, 239U

Years of operation

Isotopic composition 

of 238Pu scrap

Isotopic composition 

of 239Pu scrap

Radionuclides 

of concern

Absorption 

types

Activity 

fraction

Activity 

fraction

Significant to 

external exposurea

238Pu S 9.601E-01 - X

239Pu S 1.200E-03 1.003E-01 X

241Pu S 3.870E-02 8.997E-01 X

a. External exposure is for both 238Pu and 239Pu.

Table A-16.2. 221-H B-Line Facility

Facility/process Scrap Recovery 235U

Years of operation

Isotopic composition of 235U scrap

Radionuclides

of concern

Absorption 

types

Activity 

fraction

Significant to 

external exposure

233U S 9.556E-01 X

235U S 4.020E-02 X

236

U S 3.800E-03 X

238U S 4.000E-04 X

Table A-16.3. 221-H B-Line Facility, Np02 and Pu02 Facility.

Facility/process Neptunium oxide facility Plutonium oxide facility

Years of operation

Isotopic composition of 

NpO2 materiala

Isotopic composition of 

PuO2 material

Radionuclides

of concern 

Absorption 

types NpO2 activity fraction PuO2 activity fraction

Significant to 

external exposureb

237Np M 1.000E+02 - X

238Pu S - 9.601E-01 X

239Pu S - 8.992E-04 X

241Pu S - 2.997E-02 X

a. For purposes of dose calculations the NpO2 is assumed to be 100% 237Np.

b. External exposure is for both facilities.

A.17 H-Canyon Facility

The H-Canyon building was designed for the separation and recovery of 239Pu and 238U from irradiated 

natural uranium by the Plutonium Uranium Extraction (PUREX) (Facility/Process) process. Over the 

years, other processes have been added. These operations include the processing of irradiated 

enriched uranium to recover 235U, the processing of irradiated neptunium targets to separate and 

recover 238Pu and 237Np, and the processing of thorium to recover 233U.
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Table A-17.1. H-Canyon Facility

Facility/process First Cycle Stream- Uranium Cycle a

Years of operation 1955-1991

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

238Pu M 5.719E-04 X

239Pu M 4.981E-06

241Pu M 1.679E-03 X

237Np M 1.587E-07

234U M 2.398E-06

235U M 3.321E-08

236U M 3.136E-07

238U M 8.671E-10

144Ce M 5.535E-01 X

137Cs F 3.321E-02 X

106Ru F 2.583E-02 X

95Zr M 2.214E-01 X

89Sr F 1.531E-01 X

90Sr F 1.070E-02 X

a. The majority of the annual dose comes for the First Cycle Stream comes from Pu 

and Ce.

Table A-17.2. H-Canyon Facility.

Facility/process Second Uranium Cycle Stream a

Years of operation 1955-1991

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

238Pu M 9.016E-05

239Pu M 7.889E-07

241Pu M 2.630E-04 X

237Np M 5.071E-04 X

234U M 1.878E-01 X

235U M 2.630E-03 X

236U M 2.630E-02 X

238U M 6.950E-05

144Ce M 4.320E-01 X

137Cs F 2.442E-02 X

106Ru F 2.066E-02 X

95Zr M 1.766E-01 X

89Sr F 1.202E-01 X

90Sr F 8.452E-03 X

a. The majority of the annual dose from the Second Uranium Cycle Stream comes 

from uranium.
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Table A-17.3. H-Canyon Facility.

Facility/process Neptunium Second Cycle a

Years of operation 1955-1991

Radionuclides of 

concern 

Absorption 

typesb Activity fraction

Significant to 

external exposure

238Pu S 3.596E-03 X

239Pu S 3.083E-05

241Pu S 1.028E-02 X

237Np M 4.624E-04 X

234U S 1.284E-06

235U S 1.284E-06

236U S 1.284E-06

144Ce S 1.670E-01 X

137Cs F 1.002E-02 X

106Ru S 1.284E-02 X

95Zr M 6.936E-01 X

89Sr F 9.890E-02 X

90Sr F 3.339E-03 X

a. The majority of the annual dose for the Second Neptunium Cycle Stream comes 

from plutonium and cerium.

b. The plutonium would have been type M in the nitrate phase of the process and 

type S in the oxalate and oxide phases. Uranyl nitrate would have been type F; 

uranium trioxide: 80% F, 20% M.

Table A-17.4. H-Canyon Facility.

Facility/process Work Stream a

Years of operation 1955-1991

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

238Pu S 5.719E-04 X

239Pu S 4.981E-06

241Pu S 1.679E-03 X

237Np M 1.476E-09

234U S 2.214E-10

235U S 3.690E-12

236U S 2.952E-11

238U S 7.380E-14

144Ce S 5.535E-01 X

137Cs F 3.321E-02 X

106Ru S 2.583E-02 X

95Zr M 2.214E-01 X

89Sr F 1.531E-01 X

90Sr F 1.070E-02 X

a. The majority of the annual dose for the Work Stream comes from plutonium and 

cerium.
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A.18 211-H Area Outside Facilities

H-Area Outside Facilities include a number of processes that support the separations function of the 

200-H Area. 

Table A-18.1. H-Area Outside Facilities.

Facility/process A Line a

Years of operation 1955-1991

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposureb

234U F 8.055E-01 X

235U F 8.055E-03 X

236U F 1.859E-01 X

238U F 5.164E-04 X

a. The radionuclide of concern for A-Line is the enriched uranium as indicated.

b. The major contributor to external dose will be photons and x-rays from 234U, 

235U, and 236U.

Table A-18.2. H-Area Outside Facilities.

Facility/process Water Handling System 

Years of operation 1955-1991

Radionuclides of 

concern 

Absorption 

typesb Activity fraction

Significant to 

external exposurea

137Cs F 9.990E-04 X

106Ru S 7.992E-02 X

95Zr M 4.096E-01 X

95Nb S 5.095E-01 X

a. The radionuclides for the other facilities are 95Nb, 95Zr, 103Ru, 106Ru, and 

137Cs. However, the mixture of radionuclides for the Water Handling System is 

considered typical of the other Outside Facilities: Acid Recovery Unit, General 

Purpose Evaporators, Segregated Solvent Facilities, Transfer Tanks, Sump 

Collection Tanks, and Recycle Sumps.

b. Default type unless material is known to have been in the nitrate form, e.g., in the 

ARU, in which case Ru is type F, Zr is type F, and Nb is type M.

A.19 Receiving Basin for Off-Site Fuel (RBOF) and Resin Regeneration Facility (RRF)

The RBOF facility handles casks and fuels of various shapes, sizes, and content. Casks of spent fuel 

are received and washed in the cask basin to remove any dirt accumulated during transit. The 

solubility classes in Table 19.1 are SR-1 for 131I and SR-0 for gases or vapors that are inert and 

constitute a possible external exposure from submersion in cloud of gas and to internal exposure 

from gas within the respiratory tract. Class SR-1 refers to soluble or reactive gases or vapors with 

consideration given to retention in respiratory tract tissues, and to uptake to the systemic circulation, 

which may be less than the 100 % of the inhaled activity (ICRP 68, page9).

A.20 Tritium Facilities

The Tritium Facility includes Building 232-H, 233-H, 234-H, and 238-H. Irradiated lithium targets are 

processed in the tritium facilities to recover, purify, and package tritium.

A.21 M-Area

Uranium Target Fabrication Facility, 313-M - The Uranium Target Fabrication Facility was designed 

and built to manufacture aluminum clad targets for irradiation in the SRS reactors.
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Table A-19.1. Receiving Basin for Off-site Fuel (RBOF).

Facility/process Irradiated Fuel Receipt and Storage a

Years of operation

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

131I SR-1 1.020E-01 X

85Kr-85 SR-0 8.900E-01 X

133Xe SR-0 5.600E-03 X

a. The irradiated fuel contains uranium, fission products and activation products. 

Since the work is performed under water, the inert gases that escape the fuel 

bundles are a primary concern. The distribution above is for a Mark 22 Fuel 

Assembly cooled for 90 days.

Table A-19.2. Resin Regeneration Facility (RRF).

Facility/process Spent Ion Exchange Resin Regeneration a

Years of operation

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

134Cs F 1.500E-01 X

137Cs F 1.500E-01 X

60Co S 7.000E-01 X

a. The resins contain fission and activation products with the approximate 

distribution above. Irradiated target elements that are cut and cleaned at RRF 

contain tritium.

Table A-20. Tritium Facilities.

a

Facility/process Buildings 232-H, 233-H, 234-H, and 238-H

Years of operation 1952-Present

Radionuclides of 

concern 

Absorption 

types

Activity 

fraction

Significant to 

external exposure

Elemental Tritium

Tritiated water

65Zn S

a. Elemental tritium and tritiated water are assumed to be the primary 

radionuclides of concern at the tritium facilities. In addition, the activation 

product 65Zn is associated with the processing of lithium targets in building 

232-H.

Table A-21. Uranium Target Fabrication Facility, 313 M.

Facility/process Uranium Target Fabrication a

Years of operation 1953-1989

Radionuclides of 

concern

Absorption 

typesb Activity fraction

Significant to 

external exposure

234U S 8.399E-02 X

235U S 1.450E-02 X

238U S 9.015E-01 X

a. No information is available on the radionuclide distribution of the 313-M Facility. 

Therefore the distribution has been assumed. The majority of the annual dose 

from this mixture is delivered by 238U.

b. Based on solubility studies at similar facilities at Hanford, there may have been a 

type F component at about 10-20%. 100% type S should be assumed unless the 

urinalysis data show otherwise. 
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A.22 320-M, 322-M, and 341-M

Target extrusion was conducted in Building 320-M, and Building 322-M was a Chemical and 

Metallurgical Laboratory. The M-Area facilities processed uranium, lithium, and aluminum into fuel and 

target components for the nuclear reactors. 

Table A-22. Uranium Processing Facilities 320-M, 322-M, and 341-M.

a

Facility/process Depleted Uranium b Enriched Uranium c

Significant to 

external exposure e 

Years of operation 1953-1989

Radionuclides of 

concern 

Absorption 

typesd Activity fraction Activity fraction

234U S 8.399E-02 8.500E-01 X

235U S 1.450E-02 1.100E-02 X

236U S - 1.380E-01 X

238U S 9.015E-01 1.000E-03 X

a. No information is available on the processes or radionuclide distribution of the 320-M, 322-M, or 341-M Facilities, 

therefore the above distribution has been assumed.

b. The majority of the annual internal dose from this mixture is delivered by 238U.

c. The majority of the annual internal dose from this mixture is delivered by 234U.

d. Based on solubility studies at similar facilities at Hanford, there may have been a type F component at about 10-20%. 

100% type S should be assumed unless the urinalysis data show otherwise.

e. The majority of the external shallow dose from this mixture is delivered by 238U which decays to 234Pa which emits a high 

energy beta. External exposure is for both types of uranium.

A.23 Fuel Fabrication Facility, 321-M

The Fuel Fabrication Facility was designed and built to manufacture aluminum clad fuel elements for 

irradiation in the SRS reactors.

Table A-23. Fuel Fabrication Facility, 321-M.

a 

Facility/process Fuel Fabrication

Years of operation 1953-1989

Radionuclide content of 

material processed in 1996

Mass 

fraction

Absorption 

typesb Activity fraction C Significant to external exposure

234U 0.015 S 8.500E-01 X

235U 0.580 S 1.100E-02 X

236U 0.245 S 1.380E-01 X

238U 0.160 S 1.000E-03 X

237Np 1.0 - - X

a. Plutonium dioxide was processed in 321-M in 1980 and Pu-Al Alloy (DPSP-65-01-08) work was 

conducted in the 1960s.

b. Based on solubility studies at similar facilities at Hanford, there may have been a type F component at 

about 10-20%. 100% type S should be assumed unless the urinalysis data show otherwise. 

c. The majority of the dose from this mixture is delivered by 234U.

A.24 S-Area- Defense Waste Processing Facility (DWPF)

Radioactive wastes are stored in existing H-Area waste tank facilities in the form of settled sludge, 

supernatant liquid, and salt cake. The insoluble solids (sludge) undergo aluminum dissolving and 

washing operations in preparation for the feed to DWPF. 
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Table A-24.1. S Area Defense Waste Processing Facility (DWPF).

Facility/process Sludge Slurry Feed Stream (Stream 1) 

Years of operation 1996-???

Radionuclides of 

concern 

Absorption 

types

Activity 

fractiona,b Solubilityc

Significant to 

external exposure 244Cm M 1.682E-03 Unknown X 238Pu S 2.326E-02 PuO2, PuO2(NaTi2O5)2, Na2PuO2(OH)4 (sol) X 241Pu S 2.684E-02 Same as 238Pu X 144Ce S 1.557E-01 Unknown X 137Cs F 2.326E-02 Cs2O, CsOH (sol.) X 106Ru S 3.578E-02 Na2RuO4,(sol.) X 90Sr F 7.335E-01 Sr(NaTi2O5)2, Sr(OH)2(sol.), SrCO3, SrO X

a. Specific activity 56 Ci/gal.

b. The majority of the dose from this mixture is delivered by 90Sr and 238Pu.

c. Chemical Form is insoluble unless otherwise indicated. Pu oxides are type S, but there is some indications that pure 

238Pu oxides are removed from the lung more rapidly than 239Pu oxides.

Table A-24.2. S Area Defense Waste Processing Facility (DWPF).

Facility/process Precipitate-Slurry Feed Stream (Stream 201)

Years of operation 1996-???

Radionuclides of 

concern 

Absorption 

types

Activity 

fractiona,b Solubility c 

Significant to 

external exposure d

244Cm M 2.537E-05 Unknown X

238Pu S 3.171E-04 Same as sludge slurry X

241Pu S 2.537E-04 Same as sludge slurry X

144Ce S 1.839E-01 X

134Cs F 3.594E-03 CsO, CsOH (sol.), CsB(CH) X

137Cs F 7.611E-01 Same as 134Cs X

106Ru S 4.228E-02 X

90Sr F 8.457E-03 Sr(NaTi O)2, Sr(OH)(sol.), SrCO X

a. Specific activity is 37 Ci/gal.

b. The majority of the internal dose from this mixture is delivered by 137Cs and 144Ce.

c. Chemical Form is insoluble unless otherwise indicated.

d. The majority of the external dose comes from 144Ce, 137Cs, 106Ru, 90Sr, 238Pu, and 234Pu.

A.25 Z Area

Waste tank supernate (low-level waste) is transferred from H Area to Z Area via pipeline. At Z Area, 

the salt solution (supernate) is mixed with cement, flyash, and blast furnace slag, and the resulting 

grout is pumped to vaults where it sets into a saltstone monolith.

A.26 A Area Building 773-A

The radionuclide distribution found in some of the various sections is not available. However, general 

guidelines are given for each division based on the type of work performed.

The radionuclides of concern for the many processes in Building 773-A include the following:

Actinide Technology

The material used by the Actinide Technology Section should be similar to that found in 

Building 221-F, FB Line, Building 221-H, and HB Line.

The radionuclides of concern would include plutonium, neptunium, americium, and uranium. 
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Table A-25.1. Z Area.

Facility/process Waste Tank Supernate - Decontaminated Salt Solution

Years of operation

Radionuclides of 

concern

15-year aged a

(nCi/g)

Absorption 

types

Activity 

fraction

Significant to 

external exposureb

3

H 15.7

90Sr 0.8 (Including Y) F 1.374E-2 X

Tc-99 53.0

106Ru 60 (including Rh) M 5.888E-01 X

Sb-125 8.0

137Cs 9.4 F 3.925E-01 X

238Pu 0.6 M 4.907E-03 X

239Pu 0.007 M

All TRU elements 0.9 M

a. The compositions are based on the tetraphenylborate and the sodium titanate (NaTi2O5) precipitation 

processes for removing 137Cs and 90Sr, respectively from the salt solution. Composition varies according to 

tanks.

b. The majority of the external dose from by the salt solution is delivered by 238Pu, 137Cs, 90Sr, and 106Ru.

Table A-25.2. Z Area.

Facility/process Saltstone

Years of operation

Radionuclides of 

concern 

15-Year ageda

(nCi/g)

Absorption 

types

Activity 

fraction

Significant to 

external exposureb

3

H 7.0

90Sr 0.36 (double for Y) S 1.401E-02 X

99Tc 25.0

106Ru 26 (Double Rh) M 5.602E-01 X

125Sb 4.0

137Cs 4.4 F 4.202E-01 X

238Pu 0.3 M 5.602E-03 X

239Pu 0.003

All TRU elements 0.4

a. The compositions are based on the tetra-phenylborate and the sodium titanate (NaTi2O5) precipitation 

processes for removing 137Cs and 90Sr, respectively from the salt solution. Composition varies according to 

tanks.

b. In the saltstone, the majority of the external dose is delivered by 238Pu, 137Cs, 90Sr, and 106Ru.

Analytical Development

The material used by the Analytic Development Section may include fission products, uranium, 

neptunium, curium, and/or californium.

Defense Waste Processing Technology

The material used by the Defense Waste Processing Technology Section would be similar to 

the material processed in the S and Z area and should contain principally fission/activation 

products and plutonium.

Hydrogen Technology

The type of radioactive material used by the Hydrogen Technology Section is unknown at this 

time.
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Interim Waste Technology

The material used by the Interim Waste Processing Section is assumed to be similar to the 

material processed in the S Area and Z Area and should contain principally fission/activation 

products and plutonium.

Environmental Technology

The material used by the Environmental technology Section may contain 129I and 14C.

Laboratory Services

The material used by the Laboratory Services Section in the High Level cells and High Bay 

Experimental areas may contain any of the radionuclides listed above. 

The F-Wing Experimental Area has research and pilot scale production facilities for work with 

alpha and neutron emitters such as 239Pu, 244Cm, 243Am, and 252Cf.

Materials Technology

The material used by the Materials technology Section may contain fission/activation products, 

tritium, and depleted and/or enriched uranium.

A.27 Buildings 735-A and 735-11A

Buildings 735-A and 735-11A house the Radiological and Environmental Science facilities. Their work 

involves the radioanalysis of low-level environmental and bioassay samples and external dosimetry. 

The radionuclides of concern would include fission/activation products.

A.28 Building 776-A Liquid Waste Handling Facility

Located in Building 776-A, this facility collects the aqueous waste from Building 773-A. Facilities in 

Building 776-A are provided to strain solids from the waste streams, collect waste in batch receiving 

tanks, sample and adjust the pH of the waste tank contents, and transfer the waste to tank trailers for 

disposal. 

The radionuclides of concern would include a composite of the materials used in the building 773-A 

operations.

A.29 D-Area Heavy Water Production and Reprocessing

A heavy water production plant, in D-Area, began operation early in SRS history to concentrate heavy 

water from Savannah River to moderate and cool the site’s reactors.

A.30 CMX and TNX (TNX Semi-works) 

CMX operated from 1952 to 1984 and was used to test all new types of fuel and target assemblies 

before they were approved for use in the reactors. The type and distribution of any possible 

radionuclides in this facility is not known at this time.

TNX operated from 1952 and is currently being decommissioned. The type and distribution of 

radionuclides is not known, however, based on the processes and equipment it would appear the 

radionuclides of concern would be Am, Cm, U, and Pu. Residual contamination was determined to be

depleted uranium. 
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ATTACHMENT B

OCCUPATIONAL MEDICAL DOSE

LIST OF TABLES

Table

B-1 Average absorbed dose from PA & LAT chest procedures (mGy) for selected 

organs for 1 Gy entrance kerma (air kerma without backscatter) for a beam 

quality of 2.5 mm Al HVL and 3.5 mm Aluminum (HVL)............................................

Page

160

B-2 Organ doses for Savannah River Site for PA & LAT chest radiography .................... 161

B-3 Average absorbed dose from chest photofluorography (mGy) for selected 

organs for 1 Gy entrance kerma (air kerma without backscatter) for a beam 

quality of 2.5 mm aluminum (HVL)............................................................................ 162

B-4 Organ dose estimates for SRS Photofluorography, 1951-1957................................. 162

B.0 SRS Medical X-Ray

SRS conducted pre-employment and annual physical examinations as part of their occupational 

health program. These examinations typically included a chest x-ray. For some workers, and 

occupations, chest x-rays could be more frequent. 

B.1 Organ Doses from Medical X-Rays

X-ray organ doses for occupational x-rays at SRS are estimated for all years from 1950 to present. 

The schedule for these exams for all SRS employees over this time period is shown in Table 2.1

above. The organ dose information for SRS determined from ICRP 34 Tables A2 through A8 are 

shown in Table B-2 and B-4. X-ray organ dose estimates were made for Type I equipment (used prior

to 1970), Type II equipment (used from 1971 to 7/1985), Type III equipment (used from 8/1985 to 

5/1999), and Type IV equipment (used from 6/1999 to present). For Type 1 equipment used prior to 

1970, to insure claimant favorability dose calculations are based on minimal collimation-i.e., no 

additional external collimation or cones- and a beam quality of 2.5 mm Al, in accordance with 

instructions from T. D. Taulbee NIOSH per E. D. Scalsky (Scalsky 2004). Dose reconstruction 

instructions suggested in ORAU -OTIB-0006 Rev. 2.0 (ORAU 2003) were used to determine organ 

doses. If there is evidence that that collimation or cones were used, then organ doses should be 

revised. 

For organs outside the chest cavity, other ICRP 34 organ dose conversion factors besides PA and 

lateral chest were used to ensure that the organ doses reflected their presence in the primary beam 

based on the “Substitute Dose Conversion Factor” Table below. For example, the dose to the ovaries 

for Type I equipment assumes they were in the primary beam, so the “Substitute Dose Conversion 

Factor” in Table 2.6 were used to determine the dose conversion factor for ovaries in the abdomen.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 160 of 248

Table B-1. Average absorbed dose from PA & LAT chest procedures (mGy) for selected organs for 1 

Gy entrance kerma (air kerma without backscatter) for a beam quality of 2.5 mm Al HVL and 3.5 mm 

Aluminum (HVL). a

Organ View

Source to 

image 

distance (cm)

Image 

receptor 

size (cm)

Dose conversion 

factor (mGy per 

Gy air kerma)a

HVL 2.5 mm Al

Minimal 

collimation

Dose conversion 

factor 

(mGy per Gy air 

kerma) 

(beam quality 3.5 

mm aluminum 

HVL) collimated

Thyroid PA 183 35.6 x 43.2 174 c 62

LAT 183 35.6 x 43.2 137 151

Eye/Brain PA 183 35.6 x 43.2 32 e 62

LAT 183 35.6 x 43.2 137 151

Ovaries PA 183 35.6 x 43.2 168 d 3.2

LAT 183 35.6 x 43.2 57 d 1.6

Testes PA 183 35.6 x 43.2 9.1 d 0.01

LAT 183 35.6 x 43.2 3.3 d 0.1

Lungs (male) PA 183 35.6 x 43.2 419 565

LAT 183 35.6 x 43.2 193 276

Lungs (female) PA 183 35.6 x 43.2 451 610

LAT 183 35.6 x 43.2 220 310

Breast PA 183 35.6 x 43.2 49 91

LAT 183 35.6 x 43.2 255 316

Uterus/(embryo) PA 183 35.6 x 43.2 149 d 3

LAT 183 35.6 x 43.2 43 d 1.4

Bone marrow (male) PA 183 35.6 x 43.2 92 146

LAT 183 35.6 x 43.2 37 61

Bone marrow (female) PA 183 35.6 x 43.2 86 141

LAT 183 35.6 x 43.2 29 48

Skin b PA 1.35 X Entrance 

air kerma

1.4X Entrance air 

kerma

LAT 1.35 X Entrance 

air kerma

1.4X Entrance air 

kerma

a. Dose conversion factors from Tables A.2 through A.8, ICRP Publication 34 (1982).

b. Skin dose is entrance air kerma, multiplied by a backscatter factor of 1.35 and 1.4 for 2.5 mm Al HVL and 3.5 mm Al 

HVL respectively, from NCRP Report No. 102 (1989), Table B-8.

c. Dose Conversion Factor for AP c-spine, corrected for depth by 0.2.

d. Substitute Dose Conversion Factors Table 5.1-1 ORAUT-OTIB –0006 Rev. No. 02 (ORAU 2003) and (Table 2.6 above, 

if measurement data is available it should be used.

e. Used PA skull and LAT Chest as both were larger DCFs. 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 161 of 248

Table B-2. Organ doses for Savannah River Site for PA & LAT chest radiography. 

Organ View

Organ dose 

(rem) 1950-1970 

Type I

Organ dose 

(rem) 1971–

7/1985 Type I

Organ dose 

(rem) 8/1985-

5/1999 Type III

Organ dose (rem) 

6/1999-2003 Type IV

Entrance Air Kerma PA 1.08 mGy 0.44 mGy 0.33 mGy 0.33 mGy

Entrance Air Kerma LAT 2.70 mGy 1.10 mGy 0.83 mGy 0.83 mGy

Thyroid PA 1.88E-02 2.73E-03 2.05E-03 2.05E-03

LAT 3.70E-02 1.66E-02 1.25E-02 1.25E-02

Eye/Brain PA 3.46E-03 2.73E-03 2.05E-03 2.05E-03

LAT 3.70E-02 1.66E-02 1.25E-02 1.25E-02

Ovaries PA 1.81E-02 1.41E-04 1.06E-04 1.06E-04

LAT 1.54E-02 1.76E-04 1.32E-04 1.32E-04

Urinary/Bladder PA 1.81E-02 1.41E-04 1.06E-04 1.06E-04

LAT 1.54E-02 1.76E-04 1.32E-04 1.32E-04

Colon/Rectum PA 1.81E-02 1.41E-04 1.06E-04 1.06E-04

LAT 1.54E-02 1.76E-04 1.32E-04 1.32E-04

Testes PA 9.83E-04 4.40E-07 3.30E-07 3.30E-07

LAT 8.91E-04 1.10E-05 8.25E-06 8.25E-06

Lungs (male) PA 4.53E-02 2.49E-02 1.86E-02 1.86E-02

LAT 5.21E-02 3.04E-02 2.28E-02 2.28E-02

Lungs (female) PA 4.87E-02 2.68E-02 2.01E-02 2.01E-02

LAT 5.94E-02 3.41E-02 2.56E-02 2.56E-02

Thymus PA 4.87E-02 2.68E-02 2.01E-02 2.01E-02

LAT 5.94E-02 3.41E-02 2.56E-02 2.56E-02

Esophagus PA 4.87E-02 2.68E-02 2.01E-02 2.01E-02

LAT 5.94E-02 3.41E-02 2.56E-02 2.56E-02

Stomach PA 4.87E-02 2.68E-02 2.01E-02 2.01E-02

LAT 5.94E-02 3.41E-02 2.56E-02 2.56E-02

Bone Surface PA 4.87E-02 2.68E-02 2.01E-02 2.01E-02

LAT 5.94E-02 3.41E-02 2.56E-02 2.56E-02

Liver/Gall Bladder 

/Spleen

PA 4.87E-02 2.68E-02 2.01E-02 2.01E-02

LAT 5.94E-02 3.41E-02 2.56E-02 2.56E-02

Remainder PA 4.87E-02 2.68E-02 2.01E-02 2.01E-02

Organs LAT 5.94E-02 3.41E-02 2.56E-02 2.56E-02

Breast PA 5.29E-03 4.00E-03 3.00E-03 3.00E-03

LAT 6.89E-02 3.48E-02 2.61E-02 2.61E-02

Uterus/(Embryo) PA 1.61E-02 1.32E-04 9.90E-05 9.90E-05

LAT 1.16E-02 1.54E-04 1.16E-04 1.16E-04

Bone marrow 

(male)

PA 9.94E-03 6.42E-03 4.82E-03 4.82E-03

LAT 9.99E-03 6.71E-03 5.03E-03 5.03E-03

Bone marrow 

(female)

PA 9.29E-03 6.20E-03 4.65E-03 4.65E-03

LAT 7.83E-03 5.28E-03 3.96E-03 3.96E-03

Skin PA 1.46E-01 6.16E-02 4.62E-02 4.62E-02

LAT 3.65E-01 1.54E-01 1.16E-01 1.16E-01
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 162 of 248

Table B-3. Average absorbed dose from chest photofluorography (mGy) for selected 

organs for 1 Gy entrance kerma (air kerma without backscatter) for a beam quality of 

2.5 mm aluminum (HVL).

a

Organ View

Source-image 

distance(cm)

Image 

receptor 

size (cm)

Dose conversion factor 

(mGy per Gy air kerma) 

(beam quality 2.5 mm 

aluminum HVL)

Thyroid PFG 102 10.2 × 12.7 32

Eye/Brain PFG 102 10.2 × 12.7 32

Ovaries PFG 102 10.2 × 12.7 1

Testes PFG 102 10.2 × 12.7 0.01

Lungs (male) PFG 102 10.2 × 12.7 419

Lungs (female) PFG 102 10.2 × 12.7 451

Breast PFG 102 10.2 × 12.7 49

Uterus (Embryo) PFG 102 10.2 × 12.7 1.3

Bone marrow (male) PFG 102 10.2 × 12.7 92

Bone marrow (female PFG 102 10.2 × 12.7 86

Skin b PFG 1.35 X ESE

a. Dose conversion factors from Tables A.2 through A.8, ICRP Publication 34 (1982).

b. Skin dose is entrance air kerma, multiplied by a backscatter factor of 1.35 for 2.5 mm Al HVL, from NCRP 

Report No.102 (1989), Table B-8.

Table B-4. Organ dose estimates for SRS Photofluorography, 1951 to 

1957.

Organ View Organ dose (mGy) Organ dose (rem)

Thyroid PFG 4.80E-01 4.80E-02

Eye/Brain PFG 4.80E-01 4.80E-02

Ovaries PFG 1.50E-02 1.50E-03

Urinary/Bladder PFG 1.50E-02 1.50E-03

Colon/Rectum PFG 1.50E-02 1.50E-03

Testes PFG 1.50E-04 1.50E-05

Lungs (male) PFG 6.30E+00 6.30E-01

Lungs (female) PFG 6.77E+00 6.77E-01

Thymus PFG 6.77E+00 6.77E-01

Esophagus PFG 6.77E+00 6.77E-01

Stomach PFG 6.77E+00 6.77E-01

Bone Surface PFG 6.77E+00 6.77E-01

Liver/Gall Bladder/Spleen PFG 6.77E+00 6.77E-01

Remainder organs PFG 6.77E+00 6.77E-01

Breast PFG 7.35E-01 7.35E-02

Uterus (Embryo) PFG 1.95E-02 1.95E-03

Bone marrow (male) PFG 1.38E+00 1.38E-01

Bone marrow (female) PFG 1.29E+00 1.29E-01

Skin PFG 2.03E+01 2.03E+00
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ATTACHMENT C

OCCUPATIONAL ENVIRONMENTAL DOSE

TABLE OF CONTENTS

Section

C.0 Occupational Environmental Dose ..........................................................................................

Page

164

C.1 Internal Dose from On-Site Concentrations of Radionuclides.................................................. 164

C.1.1 On-Site Releases to Air.................................................................................................. 164

C.1.2 Resuspension of Radioactive Materials from Soil........................................................... 164

C.2 External Dose ......................................................................................................................... 164

C.2.1 Ambient Radiation.......................................................................................................... 164

C.2.2 Releases of Noble Gases............................................................................................... 164

LIST OF TABLES

Table

C-1 A Area annual intakes and geometric standard deviations (Bq per year)..................

Page

165

C-2 C Reactor area annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 166

C-3 D Area annual intakes and geometric standard deviations (Bq per year).................. 167

C-4 E Area annual intakes and geometric standard deviations (Bq per year).................. 168

C-5 F Area annual intakes and geometric standard deviations (Bq per year) .................. 169

C-6 H Area annual intakes and geometric standard deviations (Bq per year).................. 170

C-7 S Area annual intakes and geometric standard deviations (Bq per year).................. 171

C-8 Z Area annual intakes and geometric standard deviations (Bq per year) .................. 172

C-9 Forest Service Headquarters calculated air concentration (Bq per m3) .................... 173

C-10 Central Shops annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 174

C-11 L Reactor Area annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 175

C-12 P Reactor Area annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 176

C-13 K Reactor Area annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 177

C-14 R Reactor Area annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 178

C-15 Rail Yard annual intakes and geometric standard deviations (Bq per year) .............. 179

C-16 Average site-wide annual intakes and geometric standard deviations (Bq per 

year)......................................................................................................................... 180

C-17 Maximum site-wide annual intakes and geometric standard deviations (Bq 

per year)................................................................................................................... 181

C-18 Annual intakes for F and H Areas (Bq per year) ....................................................... 182

C-19 Ambient radiation levels for selected SRS locations (mrem/year)............................. 182

C-20 Argon-41 50th-percentile dose to skin and bone surfaces resulting from 

reactor releases (mrem per year) and geometric standard deviation ........................ 184

C-21 Argon-41 50th-percentile dose to whole body and other organs resulting from 

reactor releases (mrem per year) and geometric standard deviation ........................ 186
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C.0 Occupational Environmental Dose

Unmonitored workers can be exposed to occupational doses internally from the resuspension of 

radioactive materials in soil and on-site radiation releases into the air, and externally from ambient 

radiation and releases of radioactive noble gases into air.

C.1 Internal Dose from On-Site Concentrations of Radionuclides

C.1.1

Tables C-1 through C-15 show the derived intakes (assuming a 2,400 cubic meters per year 

ventilation rate) at the 50th-percentile and the associated geometric standard deviations for 3

H, 131I, 

Pu, and U, in each SRS area by year. Table C-16 shows the average of the SRS area intake s by 

year. Table C-17 shows the maximum of the C-1 through C-15 annual intakes by radionuclide and

year. To expedite dose reconstruction the claimant-favorable Table C-17 values will be used to 

estimate the annual intakes. In addition the intakes listed for 1955 are assumed to apply to the years 

1950 to 1954.

On-Site Releases to Air

C.1.2

Soil analyses indicated that elevated levels of certain radionuclides. including plutonium, were 

detected at several locations near the release points in both F and H Areas. It was concluded that the 

measured soil concentrations in F and H Areas were due to deposition during the years of the 

greatest releases. Annual intakes based on ventilation rate of 2,400 cubic meters per year resulting 

from resuspension of radionuclides in soil are presented in Table C-18.

Resuspension of Radioactive Materials from Soil

For purposes of reconstructing potential missed or unmonitored dose, the annual intakes provided in 

Table C-18 should be used only for individuals whose employment records indicate that they spent 

considerable time in either F or H Areas and only when bioassay data are unavailable. The 239Pu 

annual intakes should be assumed to have persisted since 1955 and the 238Pu annual intakes since 

1969. Annual intakes of 90Sr and 137Cs should be conservatively assumed to have persisted 

essentially unchanged since 1955.

C.2 External Dose

C.2.1

The ambient radiation levels in Table C-19 are presented in mrem/year for various locations at the 

SRP. These levels are for a full work year of 2,000 hours which represent a fraction (i.e., 0.228) of 

the full annual doses ( i.e., 8766 hours) that were derived from the SRS Annual Environmental 

Reports. The values in Table C-19 include the contributions from natural background radiation and 

nuclear weapons fallout. Therefore, use of these data to account for potential missed or unmonitored 

dose should in all cases represent an overestimation of the worker’s actual dose.

Ambient Radiation

C.2.2

Argon-41 was released in large quantities during reactor operation. The potential dose resulting from 

submersion in this noble gas is external because the gas is not absorbed by the body. Calculated air 

concentrations were estimated for populated areas at SRS and then converted to organ and whole 

body doses. The annual doses (assuming 2,000 hours per year of exposure) have been tabulated 

based on 50th-percentile source term and presented in Tables C-20 and C-21. The geometric 

Releases of Noble Gases
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 165 of 248

standard deviations provided in these tables were based on the estimated 95th-percentile source 

terms using the relationship provided in Section 3.1.1. 

Table C-1. A Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 2.73E+03 1.46 1.01E+02 1.79 9.54E-01 1.82 9.57E-02 1.82

1956 3.81E+04 1.26 2.26E+03 1.66 3.77E-03 1.82 4.39E-02 1.82

1957 1.25E+05 1.23 4.03E+02 1.89 5.04E-03 1.82 7.24E-04 1.82

1958 1.69E+05 1.21 2.60E+01 2.14 2.48E-03 1.26 4.77E-04 1.26

1959 6.21E+04 1.23 2.93E+02 1.86 2.80E-03 1.26 2.36E-03 1.26

1960 5.48E+04 1.24 1.33E+01 2.14 8.60E-03 1.26 6.60E-03 1.26

1961 5.01E+04 1.23 1.57E+02 1.81 1.34E-03 1.26 3.28E-03 1.26

1962 6.17E+04 1.26 4.87E+00 2.86 1.05E-03 1.26 2.32E-03 1.26

1963 5.94E+04 1.31 2.49E+00 3.19 3.57E-04 1.26 8.08E-03 1.26

1964 8.64E+04 1.30 1.12E+00 1.62 5.00E-04 1.26 1.40E-02 1.26

1965 4.22E+04 1.28 1.71E+00 1.36 1.39E-03 1.26 1.39E-02 1.26

1966 3.67E+04 1.28 2.92E+00 1.20 8.58E-04 1.26 5.47E-03 1.26

1967 3.86E+04 1.27 2.12E+00 1.57 9.09E-04 1.26 5.10E-03 1.26

1968 4.54E+04 1.27 2.33E+00 1.53 5.91E-04 1.26 5.95E-03 1.26

1969 2.77E+04 1.26 3.56E+00 1.50 1.22E-01 1.26 1.62E-02 1.26

1970 2.91E+04 1.25 3.13E+00 1.32 2.44E-03 1.26 4.28E-03 1.26

1971 3.84E+04 1.24 2.16E+00 1.22 2.42E-03 1.26 1.22E-03 1.26

1972 4.62E+04 1.23 2.68E-01 1.51 1.63E-03 1.26 1.90E-03 1.26

1973 3.16E+04 1.24 1.68E-01 1.28 1.86E-03 1.26 1.27E-03 1.26

1974 4.98E+04 1.51 1.65E-01 1.16 6.57E-04 1.26 2.33E-03 1.26

1975 2.77E+04 1.39 1.06E-02 1.17 1.99E-04 1.26 1.27E-03 1.26

1976 1.52E+04 1.26 1.34E-02 1.17 1.33E-03 1.26 1.33E-03 1.26

1977 2.13E+04 1.26 5.43E-03 1.25 4.13E-04 1.26 3.95E-04 1.26

1978 1.94E+04 1.28 5.66E-03 1.18 5.73E-04 1.26 7.88E-04 1.26

1979 1.83E+04 1.26 7.65E-03 1.32 1.58E-04 1.26 6.29E-04 1.26

1980 1.67E+04 1.25 2.51E-03 1.54 3.31E-04 1.26 1.04E-03 1.26

1981 2.07E+04 1.25 4.51E-03 1.44 6.12E-04 1.26 1.51E-03 1.26

1982 2.42E+04 1.24 1.02E-02 1.44 5.60E-04 1.26 2.42E-03 1.26

1983 3.48E+04 1.25 7.77E-03 1.34 2.92E-04 1.26 1.20E-03 1.26

1984 4.42E+04 1.25 2.50E-02 1.25 1.75E-04 1.26 5.62E-04 1.26

1985 3.77E+04 1.25 5.83E-03 1.43 1.03E-04 1.26 6.52E-04 1.26

1986 2.25E+04 1.23 2.45E-03 1.34 1.88E-04 1.26 3.80E-04 1.26

1987 3.20E+04 1.32 1.75E-03 2.04 1.87E-04 1.26 2.28E-03 1.26

1988 2.03E+04 1.23 8.78E-05 2.34 1.15E-04 1.26 3.87E-04 1.26

1989 1.51E+04 1.16 3.13E-05 1.15 1.57E-04 1.26 1.35E-03 1.26

1990 1.42E+04 1.16 1.04E-03 1.99 2.27E-04 1.82 3.59E-04 1.82

1991 1.08E+04 1.36 8.14E-04 1.99 5.09E-04 1.82 1.87E-04 1.82

1992 9.35E+03 1.19 2.39E-06 1.99 5.89E-05 1.82 1.20E-04 1.82

1993 1.03E+04 2.14 3.49E-04 1.99 1.32E-04 1.82 1.46E-04 1.82

1994 8.70E+03 1.71 1.28E-06 1.99 1.34E-04 1.82 1.81E-04 1.82

1995 5.32E+03 1.51 3.39E-04 1.99 1.13E-04 1.82 1.22E-04 1.82

1996 2.89E+03 1.58 8.00E-04 1.99 4.75E-05 1.82 1.33E-04 1.82

1997 3.24E+03 1.53 1.34E-04 1.99 3.94E-06 1.82 6.42E-06 1.82

1998 3.56E+03 1.48 1.68E-06 1.99 1.15E-05 1.82 3.76E-06 1.82

1999 2.27E+02 2.47 1.68E-06 1.99 4.07E-05 1.82 3.81E-06 1.82

2000 2.17E+02 2.27 1.11E-03 1.99 4.10E-05 1.82 8.12E-06 1.82

2001 2.26E+02 1.94 1.38E-03 1.99 8.30E-06 1.82 1.07E-05 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 166 of 248

Table C-2. C Reactor area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 1.23E+04 1.42 2.33E+02 1.83 2.22E+00 1.82 2.16E-01 1.82

1956 1.39E+05 1.27 5.12E+03 1.67 8.81E-03 1.82 1.03E-01 1.82

1957 3.81E+05 1.23 9.29E+02 1.98 1.17E-02 1.82 1.69E-03 1.82

1958 5.93E+05 1.22 9.30E+01 2.74 5.78E-03 1.26 1.10E-03 1.26

1959 5.11E+05 1.23 7.12E+02 1.95 6.53E-03 1.26 5.52E-03 1.26

1960 5.86E+05 1.28 3.80E+01 2.43 2.02E-02 1.26 1.55E-02 1.26

1961 3.77E+05 1.26 3.81E+02 1.88 3.11E-03 1.26 7.64E-03 1.26

1962 4.42E+05 1.27 1.83E+01 3.08 2.40E-03 1.26 5.40E-03 1.26

1963 5.90E+05 1.32 1.60E+01 3.40 8.29E-04 1.26 1.88E-02 1.26

1964 8.46E+05 1.28 3.43E+00 2.04 1.13E-03 1.26 3.24E-02 1.26

1965 7.73E+05 1.32 4.70E+00 1.70 3.09E-03 1.26 3.26E-02 1.26

1966 6.04E+05 1.28 9.26E+00 1.80 1.91E-03 1.26 1.28E-02 1.26

1967 6.39E+05 1.28 7.38E+00 2.15 2.03E-03 1.26 1.19E-02 1.26

1968 8.05E+05 1.29 7.21E+00 2.05 1.33E-03 1.26 1.34E-02 1.26

1969 4.86E+05 1.30 9.99E+00 1.84 2.66E-01 1.26 3.69E-02 1.26

1970 4.53E+05 1.29 1.16E+01 2.14 5.39E-03 1.26 9.95E-03 1.26

1971 5.28E+05 1.27 7.00E+00 1.90 5.35E-03 1.26 2.78E-03 1.26

1972 5.62E+05 1.26 1.80E+00 2.72 3.59E-03 1.26 4.38E-03 1.26

1973 4.36E+05 1.27 4.88E-01 1.74 4.08E-03 1.26 2.92E-03 1.26

1974 5.08E+05 1.35 4.07E-01 1.31 1.47E-03 1.26 5.47E-03 1.26

1975 2.79E+05 1.31 2.50E-02 1.19 4.37E-04 1.26 2.98E-03 1.26

1976 2.58E+05 1.26 3.47E-02 1.41 2.94E-03 1.26 3.13E-03 1.26

1977 4.96E+05 1.29 2.10E-02 2.09 9.08E-04 1.26 9.17E-04 1.26

1978 3.95E+05 1.29 1.39E-02 1.30 1.26E-03 1.26 1.83E-03 1.26

1979 3.55E+05 1.33 2.05E-02 1.65 3.57E-04 1.26 1.46E-03 1.26

1980 3.80E+05 1.28 8.03E-03 2.05 7.33E-04 1.26 2.42E-03 1.26

1981 2.51E+05 1.27 2.11E-02 2.41 1.36E-03 1.26 3.48E-03 1.26

1982 3.88E+05 1.31 3.39E-02 2.04 1.29E-03 1.26 5.65E-03 1.26

1983 3.88E+05 1.29 2.64E-02 2.01 6.57E-04 1.26 2.80E-03 1.26

1984 7.16E+05 1.30 5.95E-02 1.29 4.01E-04 1.26 1.31E-03 1.26

1985 9.53E+05 1.28 1.38E-02 1.44 2.39E-04 1.26 1.52E-03 1.26

1986 2.38E+05 1.27 6.03E-03 1.45 4.16E-04 1.26 8.85E-04 1.26

1987 8.64E+04 1.30 4.22E-03 2.07 4.15E-04 1.26 5.35E-03 1.26

1988 6.31E+04 1.24 2.50E-04 2.50 2.66E-04 1.26 9.05E-04 1.26

1989 4.67E+04 1.13 7.34E-05 1.15 3.64E-04 1.26 3.17E-03 1.26

1990 4.22E+04 1.12 2.43E-03 1.99 5.16E-04 1.82 8.18E-04 1.82

1991 3.22E+04 1.26 1.91E-03 1.99 1.16E-03 1.82 4.25E-04 1.82

1992 2.69E+04 1.15 5.23E-06 1.99 1.34E-04 1.82 2.81E-04 1.82

1993 2.75E+04 2.73 7.95E-04 1.99 2.99E-04 1.82 3.42E-04 1.82

1994 2.10E+04 2.03 2.79E-06 1.99 2.99E-04 1.82 4.24E-04 1.82

1995 1.77E+04 1.58 7.80E-04 1.99 2.61E-04 1.82 2.87E-04 1.82

1996 9.34E+03 1.69 1.86E-03 1.99 1.08E-04 1.82 3.11E-04 1.82

1997 1.10E+04 1.60 2.92E-04 1.99 8.87E-06 1.82 1.44E-05 1.82

1998 9.77E+03 1.55 3.95E-06 1.99 2.59E-05 1.82 8.80E-06 1.82

1999 2.25E+03 2.00 3.95E-06 1.99 9.04E-05 1.82 8.88E-06 1.82

2000 4.99E+03 1.62 2.46E-03 1.99 9.14E-05 1.82 1.90E-05 1.82

2001 5.19E+03 1.52 3.06E-03 1.99 1.83E-05 1.82 2.51E-05 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 167 of 248

Table C-3. D Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 2.12E+05 1.18 7.48E+01 1.80 6.96E-01 1.82 7.21E-02 1.82

1956 2.36E+05 1.08 1.70E+03 1.67 2.74E-03 1.82 3.18E-02 1.82

1957 3.09E+05 1.09 3.16E+02 1.91 3.66E-03 1.82 5.27E-04 1.82

1958 3.35E+05 1.11 2.29E+01 2.33 1.80E-03 1.26 3.53E-04 1.26

1959 5.40E+05 1.04 2.16E+02 1.88 2.04E-03 1.26 1.71E-03 1.26

1960 3.57E+05 1.13 1.07E+01 2.30 6.22E-03 1.26 4.77E-03 1.26

1961 2.92E+05 1.13 1.15E+02 1.84 9.84E-04 1.26 2.38E-03 1.26

1962 2.56E+05 1.17 6.02E+00 3.10 7.81E-04 1.26 1.69E-03 1.26

1963 2.60E+05 1.22 2.74E+00 3.30 2.61E-04 1.26 5.90E-03 1.26

1964 2.92E+05 1.22 1.06E+00 2.05 3.76E-04 1.26 1.03E-02 1.26

1965 2.58E+05 1.21 1.40E+00 1.64 1.06E-03 1.26 1.01E-02 1.26

1966 4.68E+05 1.13 2.22E+00 1.32 6.54E-04 1.26 3.97E-03 1.26

1967 4.70E+05 1.11 2.16E+00 2.09 6.92E-04 1.26 3.70E-03 1.26

1968 2.46E+05 1.17 2.26E+00 2.08 4.46E-04 1.26 4.49E-03 1.26

1969 2.98E+05 1.13 3.48E+00 2.02 9.50E-02 1.26 1.20E-02 1.26

1970 2.36E+05 1.12 2.54E+00 1.58 1.88E-03 1.26 3.13E-03 1.26

1971 2.31E+05 1.18 1.61E+00 1.31 1.87E-03 1.26 9.07E-04 1.26

1972 2.25E+05 1.18 2.29E-01 1.84 1.27E-03 1.26 1.40E-03 1.26

1973 1.97E+05 1.19 1.33E-01 1.51 1.45E-03 1.26 9.33E-04 1.26

1974 1.64E+05 1.31 1.20E-01 1.18 5.01E-04 1.26 1.69E-03 1.26

1975 1.72E+05 1.21 7.84E-03 1.23 1.54E-04 1.26 9.20E-04 1.26

1976 1.02E+05 1.20 9.76E-03 1.19 1.03E-03 1.26 9.65E-04 1.26

1977 1.06E+05 1.21 4.12E-03 1.37 3.19E-04 1.26 2.89E-04 1.26

1978 1.05E+05 1.20 4.19E-03 1.24 4.44E-04 1.26 5.75E-04 1.26

1979 8.35E+04 1.26 5.84E-03 1.47 1.19E-04 1.26 4.59E-04 1.26

1980 7.89E+04 1.21 2.11E-03 1.82 2.54E-04 1.26 7.58E-04 1.26

1981 8.01E+04 1.21 3.65E-03 1.68 4.70E-04 1.26 1.11E-03 1.26

1982 8.46E+04 1.23 8.26E-03 1.68 4.13E-04 1.26 1.76E-03 1.26

1983 9.64E+04 1.22 6.03E-03 1.50 2.21E-04 1.26 8.67E-04 1.26

1984 1.04E+05 1.23 1.87E-02 1.34 1.29E-04 1.26 4.10E-04 1.26

1985 1.01E+05 1.22 4.70E-03 1.64 7.59E-05 1.26 4.76E-04 1.26

1986 5.70E+04 1.21 2.08E-03 1.66 1.44E-04 1.26 2.77E-04 1.26

1987 3.76E+04 1.29 1.79E-03 2.38 1.43E-04 1.26 1.65E-03 1.26

1988 3.11E+04 1.24 1.20E-04 2.76 8.45E-05 1.26 2.81E-04 1.26

1989 2.59E+04 1.12 2.26E-05 1.15 1.14E-04 1.26 9.78E-04 1.26

1990 2.55E+04 1.11 7.50E-04 1.99 1.69E-04 1.82 2.68E-04 1.82

1991 2.32E+04 1.16 5.88E-04 1.99 3.80E-04 1.82 1.39E-04 1.82

1992 1.08E+04 1.17 1.87E-06 1.99 4.41E-05 1.82 8.71E-05 1.82

1993 2.16E+04 2.37 2.61E-04 1.99 9.84E-05 1.82 1.06E-04 1.82

1994 1.49E+04 1.84 9.97E-07 1.99 1.02E-04 1.82 1.32E-04 1.82

1995 1.21E+04 1.35 2.51E-04 1.99 8.33E-05 1.82 8.84E-05 1.82

1996 1.01E+04 1.34 5.83E-04 1.99 3.56E-05 1.82 9.59E-05 1.82

1997 1.09E+04 1.28 1.04E-04 1.99 2.98E-06 1.82 4.88E-06 1.82

1998 1.23E+04 1.22 1.22E-06 1.99 8.72E-06 1.82 2.72E-06 1.82

1999 5.82E+03 1.26 1.22E-06 1.99 3.12E-05 1.82 2.77E-06 1.82

2000 5.80E+03 1.21 8.55E-04 1.99 3.13E-05 1.82 5.90E-06 1.82

2001 5.78E+03 1.16 1.05E-03 1.99 6.42E-06 1.82 7.72E-06 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 168 of 248

Table C-4. E Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 1.64E+04 1.50 7.40E+02 1.79 6.86E+00 1.82 7.20E-01 1.82

1956 2.61E+05 1.27 1.70E+04 1.66 2.70E-02 1.82 3.13E-01 1.82

1957 9.25E+05 1.23 3.14E+03 1.88 3.61E-02 1.82 5.19E-03 1.82

1958 1.25E+06 1.21 1.93E+02 2.03 1.77E-02 1.26 3.50E-03 1.26

1959 4.36E+05 1.23 2.10E+03 1.84 2.01E-02 1.26 1.68E-02 1.26

1960 3.54E+05 1.25 8.96E+01 2.05 6.11E-02 1.26 4.69E-02 1.26

1961 3.43E+05 1.23 1.11E+03 1.80 9.72E-03 1.26 2.35E-02 1.26

1962 4.02E+05 1.27 1.85E+01 2.28 7.76E-03 1.26 1.66E-02 1.26

1963 3.97E+05 1.27 8.40E+00 2.80 2.57E-03 1.26 5.83E-02 1.26

1964 5.27E+05 1.32 6.99E+00 1.33 3.75E-03 1.26 1.02E-01 1.26

1965 2.21E+05 1.29 1.16E+01 1.24 1.07E-02 1.26 9.89E-02 1.26

1966 2.07E+05 1.29 2.04E+01 1.15 6.56E-03 1.26 3.90E-02 1.26

1967 2.06E+05 1.27 1.33E+01 1.30 6.94E-03 1.26 3.64E-02 1.26

1968 2.68E+05 1.27 1.51E+01 1.30 4.46E-03 1.26 4.49E-02 1.26

1969 1.71E+05 1.24 2.31E+01 1.27 9.61E-01 1.26 1.20E-01 1.26

1970 1.76E+05 1.22 2.15E+01 1.22 1.89E-02 1.26 3.08E-02 1.26

1971 2.39E+05 1.22 1.51E+01 1.19 1.88E-02 1.26 9.02E-03 1.26

1972 3.08E+05 1.22 1.78E+00 1.34 1.28E-02 1.26 1.39E-02 1.26

1973 2.02E+05 1.22 1.16E+00 1.20 1.46E-02 1.26 9.24E-03 1.26

1974 3.44E+05 1.54 1.17E+00 1.16 5.02E-03 1.26 1.66E-02 1.26

1975 1.81E+05 1.40 7.50E-02 1.16 1.56E-03 1.26 9.04E-03 1.26

1976 9.28E+04 1.23 9.48E-02 1.16 1.04E-02 1.26 9.48E-03 1.26

1977 1.30E+05 1.24 3.78E-02 1.20 3.22E-03 1.26 2.86E-03 1.26

1978 1.21E+05 1.27 4.00E-02 1.16 4.49E-03 1.26 5.67E-03 1.26

1979 1.11E+05 1.23 5.24E-02 1.21 1.19E-03 1.26 4.53E-03 1.26

1980 9.95E+04 1.22 1.62E-02 1.30 2.56E-03 1.26 7.47E-03 1.26

1981 1.41E+05 1.23 3.00E-02 1.28 4.73E-03 1.26 1.10E-02 1.26

1982 1.58E+05 1.22 6.73E-02 1.27 4.09E-03 1.26 1.73E-02 1.26

1983 2.37E+05 1.23 5.29E-02 1.23 2.22E-03 1.26 8.52E-03 1.26

1984 2.99E+05 1.23 1.74E-01 1.19 1.28E-03 1.26 4.05E-03 1.26

1985 2.37E+05 1.23 3.85E-02 1.25 7.51E-04 1.26 4.69E-03 1.26

1986 1.59E+05 1.21 1.65E-02 1.22 1.45E-03 1.26 2.74E-03 1.26

1987 2.40E+05 1.31 9.16E-03 1.57 1.44E-03 1.26 1.62E-02 1.26

1988 1.40E+05 1.22 4.07E-04 1.81 8.36E-04 1.26 2.77E-03 1.26

1989 9.42E+04 1.20 2.22E-04 1.15 1.13E-03 1.26 9.62E-03 1.26

1990 9.17E+04 1.20 7.37E-03 1.99 1.68E-03 1.82 2.67E-03 1.82

1991 6.96E+04 1.44 5.78E-03 1.99 3.78E-03 1.82 1.39E-03 1.82

1992 6.38E+04 1.21 1.89E-05 1.99 4.40E-04 1.82 8.56E-04 1.82

1993 7.31E+04 1.66 2.59E-03 1.99 9.81E-04 1.82 1.04E-03 1.82

1994 6.42E+04 1.51 1.01E-05 1.99 1.02E-03 1.82 1.30E-03 1.82

1995 3.96E+04 1.44 2.49E-03 1.99 8.25E-04 1.82 8.68E-04 1.82

1996 2.08E+04 1.47 5.75E-03 1.99 3.54E-04 1.82 9.42E-04 1.82

1997 2.44E+04 1.44 1.05E-03 1.99 2.98E-05 1.82 4.89E-05 1.82

1998 2.72E+04 1.43 1.20E-05 1.99 8.73E-05 1.82 2.68E-05 1.82

1999 1.27E+03 1.92 1.20E-05 1.99 3.14E-04 1.82 2.73E-05 1.82

2000 1.06E+03 1.89 8.61E-03 1.99 3.14E-04 1.82 5.80E-05 1.82

2001 1.22E+03 1.68 1.06E-02 1.99 6.48E-05 1.82 7.59E-05 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 169 of 248

Table C-5. F Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 4.50E+04 1.61 2.06E+03 1.79 2.46E+01 1.82 1.46E+00 1.82

1956 1.02E+06 1.26 3.52E+04 1.66 1.02E-01 1.82 1.22E+00 1.82

1957 1.80E+06 1.21 2.11E+03 1.88 1.35E-01 1.82 1.94E-02 1.82

1958 2.37E+06 1.20 9.54E+01 2.09 6.74E-02 1.26 1.01E-02 1.26

1959 2.23E+05 1.24 7.58E+03 1.84 7.48E-02 1.26 6.73E-02 1.26

1960 1.88E+05 1.27 3.57E+02 2.00 2.47E-01 1.26 1.89E-01 1.26

1961 1.75E+05 1.25 4.50E+03 1.80 3.26E-02 1.26 8.75E-02 1.26

1962 2.06E+05 1.29 5.07E+01 1.56 2.06E-02 1.26 6.20E-02 1.26

1963 2.05E+05 1.32 1.73E+01 1.95 9.06E-03 1.26 2.05E-01 1.26

1964 2.78E+05 1.35 2.69E+01 1.18 7.87E-03 1.26 3.39E-01 1.26

1965 1.34E+05 1.33 4.62E+01 1.18 1.41E-02 1.26 3.99E-01 1.26

1966 1.19E+05 1.32 8.21E+01 1.12 9.71E-03 1.26 1.53E-01 1.26

1967 1.23E+05 1.31 5.16E+01 1.16 1.08E-02 1.26 1.43E-01 1.26

1968 1.55E+05 1.30 5.91E+01 1.19 8.64E-03 1.26 8.70E-02 1.26

1969 9.95E+04 1.26 9.07E+01 1.18 4.65E-01 1.26 3.07E-01 1.26

1970 1.02E+05 1.24 8.60E+01 1.17 1.98E-02 1.26 1.10E-01 1.26

1971 1.36E+05 1.24 6.09E+01 1.16 1.84E-02 1.26 2.31E-02 1.26

1972 1.67E+05 1.23 6.88E+00 1.21 9.56E-03 1.26 4.24E-02 1.26

1973 1.13E+05 1.23 4.63E+00 1.17 8.88E-03 1.26 2.91E-02 1.26

1974 1.82E+05 1.52 4.72E+00 1.15 7.66E-03 1.26 6.73E-02 1.26

1975 9.80E+04 1.39 2.99E-01 1.16 1.10E-03 1.26 3.67E-02 1.26

1976 5.37E+04 1.25 3.85E-01 1.16 9.30E-03 1.26 3.81E-02 1.26

1977 7.89E+04 1.25 1.50E-01 1.17 2.60E-03 1.26 9.85E-03 1.26

1978 7.12E+04 1.28 1.62E-01 1.16 3.22E-03 1.26 2.02E-02 1.26

1979 6.58E+04 1.25 2.08E-01 1.17 2.41E-03 1.26 1.63E-02 1.26

1980 6.06E+04 1.24 6.24E-02 1.19 2.91E-03 1.26 2.71E-02 1.26

1981 7.55E+04 1.24 1.18E-01 1.19 5.60E-03 1.26 3.49E-02 1.26

1982 8.86E+04 1.24 2.65E-01 1.18 1.27E-02 1.26 6.68E-02 1.26

1983 1.27E+05 1.24 2.11E-01 1.17 3.98E-03 1.26 3.35E-02 1.26

1984 1.65E+05 1.24 7.01E-01 1.16 3.63E-03 1.26 1.43E-02 1.26

1985 1.43E+05 1.25 1.52E-01 1.17 2.43E-03 1.26 1.69E-02 1.26

1986 8.26E+04 1.23 6.60E-02 1.17 1.65E-03 1.26 9.88E-03 1.26

1987 1.16E+05 1.32 3.26E-02 1.26 1.80E-03 1.26 6.55E-02 1.26

1988 7.04E+04 1.23 1.33E-03 1.36 2.72E-03 1.26 1.07E-02 1.26

1989 4.88E+04 1.20 9.05E-04 1.15 3.96E-03 1.26 3.84E-02 1.26

1990 4.70E+04 1.19 3.00E-02 1.99 4.15E-03 1.82 6.58E-03 1.82

1991 3.58E+04 1.41 2.36E-02 1.99 9.31E-03 1.82 3.42E-03 1.82

1992 3.23E+04 1.21 9.02E-06 1.99 1.01E-03 1.82 3.37E-03 1.82

1993 3.66E+04 1.90 6.40E-03 1.99 2.31E-03 1.82 4.09E-03 1.82

1994 3.15E+04 1.61 4.82E-06 1.99 1.62E-03 1.82 5.00E-03 1.82

1995 1.94E+04 1.47 7.31E-03 1.99 2.58E-03 1.82 3.52E-03 1.82

1996 1.03E+04 1.53 2.08E-02 1.99 8.42E-04 1.82 3.84E-03 1.82

1997 1.19E+04 1.49 5.23E-04 1.99 5.58E-05 1.82 7.91E-05 1.82

1998 1.32E+04 1.46 4.88E-05 1.99 1.55E-04 1.82 1.06E-04 1.82

1999 7.18E+02 2.25 4.88E-05 1.99 3.77E-04 1.82 1.02E-04 1.82

2000 6.84E+02 2.12 9.26E-03 1.99 4.46E-04 1.82 2.23E-04 1.82

2001 7.49E+02 1.84 1.41E-02 1.99 5.22E-05 1.82 3.08E-04 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 170 of 248

Table C-6. H Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 3.04E+04 1.28 1.51E+03 1.79 7.08E+00 1.82 2.17E+00 1.82

1956 1.46E+05 1.25 5.03E+04 1.66 2.05E-02 1.82 1.92E-01 1.82

1957 2.88E+06 1.24 1.50E+04 1.87 2.94E-02 1.82 4.23E-03 1.82

1958 3.93E+06 1.20 9.41E+02 1.98 1.30E-02 1.26 6.67E-03 1.26

1959 2.12E+06 1.22 2.05E+03 1.84 1.70E-02 1.26 7.84E-03 1.26

1960 1.68E+06 1.22 3.74E+01 2.13 2.60E-02 1.26 2.09E-02 1.26

1961 1.72E+06 1.20 4.46E+02 1.81 1.29E-02 1.26 1.97E-02 1.26

1962 2.01E+06 1.24 1.29E+01 2.81 1.73E-02 1.26 1.38E-02 1.26

1963 1.97E+06 1.22 7.78E+00 3.22 2.83E-03 1.26 6.41E-02 1.26

1964 2.54E+06 1.28 3.05E+00 1.58 1.10E-02 1.26 1.36E-01 1.26

1965 1.00E+06 1.22 4.76E+00 1.33 4.19E-02 1.26 4.35E-02 1.26

1966 1.01E+06 1.23 8.37E+00 1.19 2.45E-02 1.26 2.34E-02 1.26

1967 9.22E+05 1.22 5.71E+00 1.50 2.52E-02 1.26 2.04E-02 1.26

1968 1.27E+06 1.22 6.43E+00 1.46 1.40E-02 1.26 1.41E-01 1.26

1969 8.03E+05 1.21 9.74E+00 1.43 4.82E+00 1.26 2.80E-01 1.26

1970 8.58E+05 1.19 8.72E+00 1.29 8.10E-02 1.26 3.25E-02 1.26

1971 1.10E+06 1.21 6.23E+00 1.21 8.23E-02 1.26 2.12E-02 1.26

1972 1.47E+06 1.20 7.70E-01 1.46 5.97E-02 1.26 2.37E-02 1.26

1973 1.00E+06 1.19 4.93E-01 1.25 7.10E-02 1.26 1.46E-02 1.26

1974 1.68E+06 1.53 5.02E-01 1.16 1.84E-02 1.26 6.87E-03 1.26

1975 8.69E+05 1.39 3.69E-02 1.17 7.37E-03 1.26 3.73E-03 1.26

1976 4.69E+05 1.20 3.81E-02 1.17 4.66E-02 1.26 4.40E-03 1.26

1977 5.77E+05 1.21 1.75E-02 1.23 1.48E-02 1.26 3.41E-03 1.26

1978 5.58E+05 1.24 1.67E-02 1.17 2.11E-02 1.26 5.96E-03 1.26

1979 5.06E+05 1.20 2.40E-02 1.29 3.61E-03 1.26 4.54E-03 1.26

1980 4.50E+05 1.19 7.95E-03 1.47 1.07E-02 1.26 7.20E-03 1.26

1981 7.13E+05 1.21 1.27E-02 1.41 1.94E-02 1.26 1.70E-02 1.26

1982 7.41E+05 1.20 2.88E-02 1.41 6.65E-03 1.26 1.14E-02 1.26

1983 1.16E+06 1.22 2.22E-02 1.32 7.37E-03 1.26 4.88E-03 1.26

1984 1.41E+06 1.21 7.20E-02 1.24 2.57E-03 1.26 4.42E-03 1.26

1985 1.05E+06 1.21 1.68E-02 1.39 1.11E-03 1.26 4.70E-03 1.26

1986 8.77E+05 1.19 6.98E-03 1.30 6.06E-03 1.26 2.70E-03 1.26

1987 1.28E+06 1.30 4.76E-03 1.98 5.80E-03 1.26 7.14E-03 1.26

1988 7.17E+05 1.21 2.28E-04 2.27 1.20E-03 1.26 1.86E-03 1.26

1989 4.38E+05 1.21 8.65E-05 1.15 1.27E-03 1.26 4.80E-03 1.26

1990 4.33E+05 1.20 2.86E-03 1.99 4.15E-03 1.82 6.58E-03 1.82

1991 3.28E+05 1.46 2.24E-03 1.99 9.31E-03 1.82 3.42E-03 1.82

1992 3.06E+05 1.21 9.47E-05 1.99 1.18E-03 1.82 4.92E-04 1.82

1993 3.54E+05 1.45 6.38E-03 1.99 2.56E-03 1.82 6.06E-04 1.82

1994 3.16E+05 1.42 5.06E-05 1.99 3.66E-03 1.82 8.59E-04 1.82

1995 1.95E+05 1.41 4.61E-03 1.99 1.33E-03 1.82 3.70E-04 1.82

1996 1.01E+05 1.42 5.67E-03 1.99 9.11E-04 1.82 3.71E-04 1.82

1997 1.21E+05 1.41 5.27E-03 1.99 9.60E-05 1.82 1.74E-04 1.82

1998 1.35E+05 1.41 4.64E-06 1.99 2.92E-04 1.82 1.39E-05 1.82

1999 5.55E+03 1.52 4.64E-06 1.99 1.29E-03 1.82 2.29E-05 1.82

2000 4.08E+03 1.53 3.66E-02 1.99 1.20E-03 1.82 4.05E-05 1.82

2001 4.98E+03 1.47 4.14E-02 1.99 2.98E-04 1.82 3.19E-05 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 171 of 248

Table C-7. S Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 1.31E+04 1.35 6.23E+02 1.79 3.68E+00 1.82 8.18E-01 1.82

1956 1.03E+05 1.27 1.90E+04 1.66 1.22E-02 1.82 1.28E-01 1.82

1957 1.08E+06 1.24 5.24E+03 1.87 1.70E-02 1.82 2.44E-03 1.82

1958 1.47E+06 1.21 3.30E+02 2.00 7.90E-03 1.26 2.80E-03 1.26

1959 7.35E+05 1.23 1.09E+03 1.85 9.65E-03 1.26 6.11E-03 1.26

1960 5.83E+05 1.23 3.23E+01 2.13 2.15E-02 1.26 1.68E-02 1.26

1961 5.82E+05 1.21 3.83E+02 1.81 6.06E-03 1.26 1.12E-02 1.26

1962 6.79E+05 1.26 1.12E+01 2.81 6.96E-03 1.26 7.88E-03 1.26

1963 6.68E+05 1.24 6.87E+00 3.23 1.43E-03 1.26 3.24E-02 1.26

1964 8.72E+05 1.30 2.64E+00 1.56 4.17E-03 1.26 6.37E-02 1.26

1965 3.37E+05 1.25 4.10E+00 1.32 1.51E-02 1.26 3.52E-02 1.26

1966 3.28E+05 1.26 7.14E+00 1.18 8.88E-03 1.26 1.58E-02 1.26

1967 3.16E+05 1.24 4.94E+00 1.48 9.19E-03 1.26 1.43E-02 1.26

1968 4.22E+05 1.25 5.51E+00 1.45 5.24E-03 1.26 5.28E-02 1.26

1969 2.67E+05 1.23 8.40E+00 1.42 1.67E+00 1.26 1.12E-01 1.26

1970 2.78E+05 1.20 7.53E+00 1.28 2.87E-02 1.26 1.67E-02 1.26

1971 3.78E+05 1.21 5.28E+00 1.21 2.91E-02 1.26 8.42E-03 1.26

1972 5.01E+05 1.21 6.45E-01 1.45 2.09E-02 1.26 1.03E-02 1.26

1973 3.28E+05 1.20 4.12E-01 1.25 2.48E-02 1.26 6.48E-03 1.26

1974 5.71E+05 1.55 4.13E-01 1.16 6.70E-03 1.26 5.78E-03 1.26

1975 2.97E+05 1.40 2.80E-02 1.17 2.58E-03 1.26 3.14E-03 1.26

1976 1.48E+05 1.22 3.27E-02 1.17 1.64E-02 1.26 3.45E-03 1.26

1977 1.97E+05 1.22 1.38E-02 1.23 5.20E-03 1.26 1.67E-03 1.26

1978 1.88E+05 1.26 1.40E-02 1.17 7.40E-03 1.26 3.07E-03 1.26

1979 1.70E+05 1.21 1.93E-02 1.30 1.36E-03 1.26 2.38E-03 1.26

1980 1.51E+05 1.20 6.32E-03 1.49 3.80E-03 1.26 3.85E-03 1.26

1981 2.33E+05 1.23 1.09E-02 1.40 6.94E-03 1.26 7.60E-03 1.26

1982 2.52E+05 1.21 2.46E-02 1.41 2.93E-03 1.26 7.32E-03 1.26

1983 3.91E+05 1.22 1.89E-02 1.32 2.73E-03 1.26 3.38E-03 1.26

1984 4.82E+05 1.22 6.13E-02 1.24 1.07E-03 1.26 2.24E-03 1.26

1985 3.59E+05 1.22 1.42E-02 1.40 5.04E-04 1.26 2.47E-03 1.26

1986 2.72E+05 1.20 5.93E-03 1.30 2.16E-03 1.26 1.43E-03 1.26

1987 4.20E+05 1.31 4.10E-03 1.98 2.08E-03 1.26 5.78E-03 1.26

1988 2.40E+05 1.22 1.93E-04 2.25 5.51E-04 1.26 1.18E-03 1.26

1989 1.57E+05 1.20 7.57E-05 1.15 6.38E-04 1.26 3.60E-03 1.26

1990 1.54E+05 1.20 2.51E-03 1.99 1.63E-03 1.82 2.59E-03 1.82

1991 1.16E+05 1.45 1.97E-03 1.99 3.66E-03 1.82 1.34E-03 1.82

1992 1.08E+05 1.21 3.29E-05 1.99 4.55E-04 1.82 3.40E-04 1.82

1993 1.24E+05 1.52 2.51E-03 1.99 9.92E-04 1.82 4.16E-04 1.82

1994 1.10E+05 1.45 1.76E-05 1.99 1.33E-03 1.82 5.49E-04 1.82

1995 6.81E+04 1.42 1.95E-03 1.99 5.88E-04 1.82 3.06E-04 1.82

1996 3.56E+04 1.43 3.00E-03 1.99 3.55E-04 1.82 3.22E-04 1.82

1997 4.21E+04 1.42 1.83E-03 1.99 3.57E-05 1.82 6.35E-05 1.82

1998 4.70E+04 1.41 4.07E-06 1.99 1.08E-04 1.82 1.02E-05 1.82

1999 2.00E+03 1.67 4.07E-06 1.99 4.59E-04 1.82 1.30E-05 1.82

2000 1.52E+03 1.68 1.30E-02 1.99 4.32E-04 1.82 2.52E-05 1.82

2001 1.82E+03 1.55 1.49E-02 1.99 1.05E-04 1.82 2.66E-05 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 172 of 248

Table C-8. Z Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 8.30E+03 1.40 3.78E+02 1.79 2.50E+00 1.82 4.69E-01 1.82

1956 7.88E+04 1.27 1.09E+04 1.66 8.79E-03 1.82 9.53E-02 1.82

1957 6.14E+05 1.24 2.84E+03 1.88 1.20E-02 1.82 1.73E-03 1.82

1958 8.36E+05 1.21 1.80E+02 2.02 5.71E-03 1.26 1.72E-03 1.26

1959 4.00E+05 1.24 7.51E+02 1.85 6.80E-03 1.26 4.75E-03 1.26

1960 3.20E+05 1.25 2.62E+01 2.16 1.69E-02 1.26 1.31E-02 1.26

1961 3.15E+05 1.23 3.05E+02 1.82 3.95E-03 1.26 7.92E-03 1.26

1962 3.69E+05 1.27 1.02E+01 2.90 4.17E-03 1.26 5.57E-03 1.26

1963 3.62E+05 1.27 6.86E+00 3.29 9.64E-04 1.26 2.18E-02 1.26

1964 4.77E+05 1.32 2.17E+00 1.62 2.40E-03 1.26 4.15E-02 1.26

1965 1.88E+05 1.27 3.30E+00 1.35 8.35E-03 1.26 2.76E-02 1.26

1966 1.79E+05 1.28 5.68E+00 1.19 4.95E-03 1.26 1.18E-02 1.26

1967 1.77E+05 1.27 4.02E+00 1.53 5.14E-03 1.26 1.08E-02 1.26

1968 2.31E+05 1.26 4.44E+00 1.49 2.98E-03 1.26 3.01E-02 1.26

1969 1.47E+05 1.23 6.79E+00 1.46 9.03E-01 1.26 6.63E-02 1.26

1970 1.52E+05 1.21 6.02E+00 1.29 1.58E-02 1.26 1.13E-02 1.26

1971 2.10E+05 1.22 4.19E+00 1.22 1.59E-02 1.26 5.00E-03 1.26

1972 2.75E+05 1.21 5.16E-01 1.47 1.14E-02 1.26 6.42E-03 1.26

1973 1.79E+05 1.21 3.27E-01 1.27 1.34E-02 1.26 4.10E-03 1.26

1974 3.12E+05 1.55 3.25E-01 1.16 3.74E-03 1.26 4.57E-03 1.26

1975 1.64E+05 1.41 2.16E-02 1.17 1.40E-03 1.26 2.49E-03 1.26

1976 8.05E+04 1.23 2.60E-02 1.17 8.97E-03 1.26 2.68E-03 1.26

1977 1.10E+05 1.23 1.08E-02 1.24 2.83E-03 1.26 1.11E-03 1.26

1978 1.04E+05 1.26 1.11E-02 1.18 4.02E-03 1.26 2.08E-03 1.26

1979 9.49E+04 1.22 1.53E-02 1.33 7.78E-04 1.26 1.63E-03 1.26

1980 8.40E+04 1.21 5.06E-03 1.55 2.09E-03 1.26 2.65E-03 1.26

1981 1.26E+05 1.23 8.75E-03 1.43 3.82E-03 1.26 4.83E-03 1.26

1982 1.38E+05 1.21 1.99E-02 1.45 1.84E-03 1.26 5.38E-03 1.26

1983 2.13E+05 1.23 1.52E-02 1.34 1.54E-03 1.26 2.54E-03 1.26

1984 2.64E+05 1.22 4.89E-02 1.26 6.49E-04 1.26 1.51E-03 1.26

1985 2.00E+05 1.23 1.15E-02 1.44 3.22E-04 1.26 1.69E-03 1.26

1986 1.44E+05 1.21 4.75E-03 1.33 1.19E-03 1.26 9.80E-04 1.26

1987 2.24E+05 1.31 3.51E-03 2.08 1.15E-03 1.26 4.54E-03 1.26

1988 1.30E+05 1.22 1.68E-04 2.33 3.54E-04 1.26 8.66E-04 1.26

1989 8.81E+04 1.20 6.04E-05 1.15 4.28E-04 1.26 2.77E-03 1.26

1990 8.55E+04 1.20 2.00E-03 1.99 9.62E-04 1.82 1.53E-03 1.82

1991 6.44E+04 1.44 1.57E-03 1.99 2.16E-03 1.82 7.92E-04 1.82

1992 5.91E+04 1.21 1.77E-05 1.99 2.65E-04 1.82 2.56E-04 1.82

1993 6.80E+04 1.60 1.48E-03 1.99 5.80E-04 1.82 3.12E-04 1.82

1994 6.01E+04 1.48 9.48E-06 1.99 7.48E-04 1.82 4.04E-04 1.82

1995 3.71E+04 1.43 1.20E-03 1.99 3.70E-04 1.82 2.41E-04 1.82

1996 1.94E+04 1.45 2.06E-03 1.99 2.08E-04 1.82 2.57E-04 1.82

1997 2.28E+04 1.43 9.88E-04 1.99 2.03E-05 1.82 3.56E-05 1.82

1998 2.55E+04 1.42 3.25E-06 1.99 6.09E-05 1.82 7.78E-06 1.82

1999 1.13E+03 1.82 3.25E-06 1.99 2.53E-04 1.82 9.20E-06 1.82

2000 8.77E+02 1.83 7.14E-03 1.99 2.40E-04 1.82 1.84E-05 1.82

2001 1.03E+03 1.64 8.26E-03 1.99 5.70E-05 1.82 2.10E-05 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 173 of 248

Table C-9. Forest Service Headquartersa calculated air concentration (Bq per m3).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 2.36E+03 1.49 9.55E+01 1.80 8.19E-01 1.82 9.94E-02 1.82

1956 3.09E+04 1.28 2.34E+03 1.66 3.15E-03 1.82 3.61E-02 1.82

1957 1.30E+05 1.24 4.88E+02 1.88 4.23E-03 1.82 6.08E-04 1.82

1958 1.75E+05 1.22 3.16E+01 2.11 2.06E-03 1.26 4.47E-04 1.26

1959 7.08E+04 1.26 2.51E+02 1.86 2.36E-03 1.26 1.91E-03 1.26

1960 5.95E+04 1.29 1.10E+01 2.18 6.95E-03 1.26 5.34E-03 1.26

1961 5.63E+04 1.27 1.27E+02 1.82 1.19E-03 1.26 2.76E-03 1.26

1962 6.70E+04 1.33 5.06E+00 2.99 1.01E-03 1.26 1.95E-03 1.26

1963 6.45E+04 1.36 2.99E+00 3.30 3.09E-04 1.26 6.99E-03 1.26

1964 8.84E+04 1.40 9.45E-01 1.71 5.15E-04 1.26 1.24E-02 1.26

1965 4.06E+04 1.36 1.39E+00 1.39 1.56E-03 1.26 1.13E-02 1.26

1966 3.60E+04 1.36 2.36E+00 1.21 9.50E-04 1.26 4.50E-03 1.26

1967 3.76E+04 1.34 1.75E+00 1.62 9.99E-04 1.26 4.18E-03 1.26

1968 4.59E+04 1.32 1.91E+00 1.58 6.21E-04 1.26 6.26E-03 1.26

1969 3.02E+04 1.26 2.92E+00 1.54 1.51E-01 1.26 1.58E-02 1.26

1970 3.16E+04 1.24 2.53E+00 1.33 2.83E-03 1.26 3.69E-03 1.26

1971 4.25E+04 1.24 1.74E+00 1.23 2.84E-03 1.26 1.19E-03 1.26

1972 5.26E+04 1.23 2.18E-01 1.53 1.96E-03 1.26 1.75E-03 1.26

1973 3.54E+04 1.23 1.36E-01 1.30 2.27E-03 1.26 1.15E-03 1.26

1974 5.79E+04 1.53 1.33E-01 1.16 7.24E-04 1.26 1.88E-03 1.26

1975 3.19E+04 1.40 8.62E-03 1.18 2.40E-04 1.26 1.03E-03 1.26

1976 1.67E+04 1.25 1.08E-02 1.17 1.58E-03 1.26 1.08E-03 1.26

1977 2.31E+04 1.25 4.40E-03 1.25 4.93E-04 1.26 3.45E-04 1.26

1978 2.13E+04 1.28 4.57E-03 1.18 6.91E-04 1.26 6.78E-04 1.26

1979 2.00E+04 1.25 6.31E-03 1.38 1.65E-04 1.26 5.39E-04 1.26

1980 1.79E+04 1.24 2.14E-03 1.64 3.81E-04 1.26 8.87E-04 1.26

1981 2.36E+04 1.25 3.71E-03 1.49 7.01E-04 1.26 1.37E-03 1.26

1982 2.71E+04 1.23 8.46E-03 1.51 5.11E-04 1.26 2.00E-03 1.26

1983 3.99E+04 1.24 6.38E-03 1.39 3.12E-04 1.26 9.80E-04 1.26

1984 5.02E+04 1.24 2.04E-02 1.29 1.65E-04 1.26 4.86E-04 1.26

1985 4.08E+04 1.25 4.90E-03 1.51 9.28E-05 1.26 5.59E-04 1.26

1986 2.58E+04 1.23 2.00E-03 1.37 2.16E-04 1.26 3.25E-04 1.26

1987 3.86E+04 1.32 1.63E-03 2.20 2.12E-04 1.26 1.85E-03 1.26

1988 2.40E+04 1.24 8.01E-05 2.45 1.03E-04 1.26 3.21E-04 1.26

1989 1.75E+04 1.19 2.52E-05 1.15 1.36E-04 1.26 1.10E-03 1.26

1990 1.63E+04 1.19 8.35E-04 1.99 2.24E-04 1.82 3.55E-04 1.82

1991 1.21E+04 1.41 6.55E-04 1.99 5.02E-04 1.82 1.84E-04 1.82

1992 1.08E+04 1.21 2.96E-06 1.99 5.93E-05 1.82 9.85E-05 1.82

1993 1.22E+04 1.97 3.44E-04 1.99 1.32E-04 1.82 1.20E-04 1.82

1994 1.05E+04 1.64 1.58E-06 1.99 1.46E-04 1.82 1.50E-04 1.82

1995 6.43E+03 1.49 3.16E-04 1.99 1.03E-04 1.82 9.87E-05 1.82

1996 3.44E+03 1.55 6.84E-04 1.99 4.74E-05 1.82 1.07E-04 1.82

1997 3.91E+03 1.50 1.65E-04 1.99 4.17E-06 1.82 7.00E-06 1.82

1998 4.34E+03 1.46 1.35E-06 1.99 1.23E-05 1.82 3.07E-06 1.82

1999 2.29E+02 2.41 1.35E-06 1.99 4.66E-05 1.82 3.21E-06 1.82

2000 1.97E+02 2.33 1.29E-03 1.99 4.57E-05 1.82 6.75E-06 1.82

2001 2.15E+02 1.98 1.55E-03 1.99 9.91E-06 1.82 8.62E-06 1.82

a. Values are for plant perimeter.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 174 of 248

Table C-10. Central Shops annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 4.45E+04 1.21 1.49E+02 1.81 1.31E+00 1.82 1.51E-01 1.82

1956 9.48E+04 1.19 3.57E+03 1.67 5.06E-03 1.82 5.83E-02 1.82

1957 2.49E+05 1.21 7.28E+02 1.90 6.79E-03 1.82 9.76E-04 1.82

1958 3.27E+05 1.20 5.29E+01 2.31 3.32E-03 1.26 6.97E-04 1.26

1959 2.37E+05 1.17 4.06E+02 1.89 3.79E-03 1.26 3.09E-03 1.26

1960 1.97E+05 1.25 1.92E+01 2.29 1.13E-02 1.26 8.65E-03 1.26

1961 1.60E+05 1.24 2.08E+02 1.84 1.88E-03 1.26 4.42E-03 1.26

1962 1.73E+05 1.29 1.05E+01 3.09 1.57E-03 1.26 3.13E-03 1.26

1963 1.82E+05 1.35 6.03E+00 3.34 4.91E-04 1.26 1.11E-02 1.26

1964 2.42E+05 1.37 1.83E+00 1.99 7.86E-04 1.26 1.96E-02 1.26

1965 1.76E+05 1.34 2.47E+00 1.59 2.34E-03 1.26 1.82E-02 1.26

1966 1.93E+05 1.26 4.21E+00 1.43 1.43E-03 1.26 7.26E-03 1.26

1967 2.00E+05 1.25 3.49E+00 1.95 1.50E-03 1.26 6.75E-03 1.26

1968 1.83E+05 1.29 3.58E+00 1.87 9.44E-04 1.26 9.51E-03 1.26

1969 1.44E+05 1.23 5.35E+00 1.77 2.22E-01 1.26 2.44E-02 1.26

1970 1.32E+05 1.22 4.79E+00 1.68 4.22E-03 1.26 5.88E-03 1.26

1971 1.52E+05 1.24 3.10E+00 1.47 4.21E-03 1.26 1.84E-03 1.26

1972 1.65E+05 1.23 5.26E-01 2.19 2.90E-03 1.26 2.74E-03 1.26

1973 1.26E+05 1.24 2.39E-01 1.49 3.35E-03 1.26 1.81E-03 1.26

1974 1.54E+05 1.42 2.18E-01 1.20 1.09E-03 1.26 3.05E-03 1.26

1975 1.01E+05 1.31 1.40E-02 1.19 3.55E-04 1.26 1.66E-03 1.26

1976 6.63E+04 1.24 1.80E-02 1.24 2.34E-03 1.26 1.75E-03 1.26

1977 9.84E+04 1.27 8.32E-03 1.59 7.29E-04 1.26 5.48E-04 1.26

1978 8.56E+04 1.27 7.53E-03 1.23 1.02E-03 1.26 1.08E-03 1.26

1979 7.71E+04 1.29 1.07E-02 1.50 2.51E-04 1.26 8.60E-04 1.26

1980 7.49E+04 1.26 3.91E-03 1.85 5.68E-04 1.26 1.42E-03 1.26

1981 6.95E+04 1.25 7.63E-03 1.92 1.05E-03 1.26 2.15E-03 1.26

1982 8.84E+04 1.27 1.56E-02 1.75 8.02E-04 1.26 3.22E-03 1.26

1983 1.07E+05 1.26 1.17E-02 1.64 4.73E-04 1.26 1.58E-03 1.26

1984 1.53E+05 1.27 3.36E-02 1.33 2.57E-04 1.26 7.73E-04 1.26

1985 1.64E+05 1.27 8.24E-03 1.59 1.46E-04 1.26 8.91E-04 1.26

1986 6.73E+04 1.25 3.42E-03 1.49 3.23E-04 1.26 5.19E-04 1.26

1987 6.34E+04 1.32 2.97E-03 2.31 3.17E-04 1.26 2.99E-03 1.26

1988 4.28E+04 1.25 1.82E-04 2.67 1.62E-04 1.26 5.16E-04 1.26

1989 3.19E+04 1.18 4.08E-05 1.15 2.16E-04 1.26 1.78E-03 1.26

1990 2.98E+04 1.17 1.35E-03 1.99 3.45E-04 1.82 5.47E-04 1.82

1991 2.31E+04 1.33 1.06E-03 1.99 7.74E-04 1.82 2.84E-04 1.82

1992 1.84E+04 1.20 4.35E-06 1.99 9.10E-05 1.82 1.59E-04 1.82

1993 2.21E+04 2.34 5.31E-04 1.99 2.02E-04 1.82 1.93E-04 1.82

1994 1.76E+04 1.82 2.32E-06 1.99 2.20E-04 1.82 2.42E-04 1.82

1995 1.17E+04 1.51 4.94E-04 1.99 1.62E-04 1.82 1.60E-04 1.82

1996 7.00E+03 1.58 1.09E-03 1.99 7.29E-05 1.82 1.73E-04 1.82

1997 7.81E+03 1.50 2.43E-04 1.99 6.33E-06 1.82 1.05E-05 1.82

1998 8.46E+03 1.44 2.20E-06 1.99 1.86E-05 1.82 4.95E-06 1.82

1999 1.62E+03 1.73 2.20E-06 1.99 6.95E-05 1.82 5.14E-06 1.82

2000 1.81E+03 1.60 1.92E-03 1.99 6.86E-05 1.82 1.09E-05 1.82

2001 1.83E+03 1.46 2.32E-03 1.99 1.47E-05 1.82 1.40E-05 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 175 of 248

Table C-11. L Reactor Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 5.72E+03 2.71 9.81E+01 1.90 8.19E-01 1.82 9.94E-02 1.82

1956 4.07E+04 1.77 2.35E+03 1.68 3.15E-03 1.82 3.61E-02 1.82

1957 1.54E+05 1.64 5.25E+02 2.02 4.23E-03 1.82 6.08E-04 1.82

1958 2.11E+05 1.72 6.89E+01 2.94 2.06E-03 1.26 4.47E-04 1.26

1959 1.18E+05 2.09 2.84E+02 2.10 2.36E-03 1.26 1.91E-03 1.26

1960 1.30E+05 2.29 2.25E+01 2.93 6.95E-03 1.26 5.34E-03 1.26

1961 1.23E+05 2.44 1.51E+02 2.10 1.19E-03 1.26 2.76E-03 1.26

1962 1.66E+05 2.49 2.49E+01 3.35 1.01E-03 1.26 1.95E-03 1.26

1963 1.44E+05 2.80 1.54E+01 3.48 3.09E-04 1.26 6.99E-03 1.26

1964 3.14E+05 3.12 1.43E+01 3.07 5.15E-04 1.26 1.24E-02 1.26

1965 2.00E+05 2.98 7.72E+00 2.90 1.56E-03 1.26 1.13E-02 1.26

1966 1.63E+05 2.96 4.42E+00 2.20 9.50E-04 1.26 4.50E-03 1.26

1967 1.58E+05 3.03 1.28E+01 2.94 9.99E-04 1.26 4.18E-03 1.26

1968 1.33E+05 3.18 3.73E+00 2.52 6.21E-04 1.26 6.26E-03 1.26

1969 6.99E+04 1.29 4.96E+00 2.30 1.51E-01 1.26 1.58E-02 1.26

1970 8.06E+04 1.28 3.00E+00 1.71 2.83E-03 1.26 3.69E-03 1.26

1971 9.45E+04 1.27 1.81E+00 1.34 2.84E-03 1.26 1.19E-03 1.26

1972 9.96E+04 1.25 2.75E-01 1.96 1.96E-03 1.26 1.75E-03 1.26

1973 8.11E+04 1.27 1.64E-01 1.70 2.27E-03 1.26 1.15E-03 1.26

1974 9.95E+04 1.43 1.34E-01 1.18 7.24E-04 1.26 1.88E-03 1.26

1975 6.56E+04 1.35 8.96E-03 1.28 2.40E-04 1.26 1.03E-03 1.26

1976 4.58E+04 1.27 1.09E-02 1.19 1.58E-03 1.26 1.08E-03 1.26

1977 4.88E+04 1.28 4.64E-03 1.38 4.93E-04 1.26 3.45E-04 1.26

1978 4.50E+04 1.29 4.74E-03 1.29 6.91E-04 1.26 6.78E-04 1.26

1979 4.59E+04 1.28 8.06E-03 1.90 1.65E-04 1.26 5.39E-04 1.26

1980 4.30E+04 1.27 3.65E-03 2.35 3.81E-04 1.26 8.87E-04 1.26

1981 4.67E+04 1.27 5.02E-03 2.01 7.01E-04 1.26 1.37E-03 1.26

1982 5.25E+04 1.27 1.26E-02 2.14 5.11E-04 1.26 2.00E-03 1.26

1983 7.29E+04 1.31 8.15E-03 1.86 3.12E-04 1.26 9.80E-04 1.26

1984 8.36E+04 1.27 2.42E-02 1.68 1.65E-04 1.26 4.86E-04 1.26

1985 7.63E+04 1.30 7.79E-03 2.18 9.28E-05 1.26 5.59E-04 1.26

1986 5.07E+04 1.72 2.80E-03 1.95 2.16E-04 1.26 3.25E-04 1.26

1987 6.11E+04 1.64 6.65E-03 2.90 2.12E-04 1.26 1.85E-03 1.26

1988 4.57E+04 1.70 3.42E-03 3.16 1.03E-04 1.26 3.21E-04 1.26

1989 3.72E+04 1.98 2.52E-05 1.15 1.36E-04 1.26 1.10E-03 1.26

1990 3.20E+04 2.00 8.35E-04 1.99 2.24E-04 1.82 3.55E-04 1.82

1991 2.38E+04 2.20 6.55E-04 1.99 5.02E-04 1.82 1.84E-04 1.82

1992 1.90E+04 1.82 2.96E-06 1.99 5.93E-05 1.82 9.85E-05 1.82

1993 1.94E+04 3.54 3.44E-04 1.99 1.32E-04 1.82 1.20E-04 1.82

1994 1.45E+04 2.78 1.58E-06 1.99 1.46E-04 1.82 1.50E-04 1.82

1995 8.48E+03 2.33 3.16E-04 1.99 1.03E-04 1.82 9.87E-05 1.82

1996 5.43E+03 2.88 6.84E-04 1.99 4.74E-05 1.82 1.07E-04 1.82

1997 5.29E+03 2.41 1.65E-04 1.99 4.17E-06 1.82 7.00E-06 1.82

1998 5.48E+03 2.05 1.35E-06 1.99 1.23E-05 1.82 3.07E-06 1.82

1999 1.43E+03 3.27 1.35E-06 1.99 4.66E-05 1.82 3.21E-06 1.82

2000 1.54E+03 3.37 1.29E-03 1.99 4.57E-05 1.82 6.75E-06 1.82

2001 1.61E+03 2.49 1.55E-03 1.99 9.91E-06 1.82 8.62E-06 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 176 of 248

Table C-12. P Reactor Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 1.73E+04 1.43 8.85E+01 1.91 7.04E-01 1.82 9.26E-02 1.82

1956 1.01E+05 1.31 2.19E+03 1.69 2.67E-03 1.82 3.04E-02 1.82

1957 2.23E+05 1.27 5.25E+02 2.05 3.60E-03 1.82 5.18E-04 1.82

1958 2.74E+05 1.26 5.73E+01 2.82 1.75E-03 1.26 3.99E-04 1.26

1959 2.49E+05 1.31 2.47E+02 2.13 2.01E-03 1.26 1.60E-03 1.26

1960 4.08E+05 1.32 2.35E+01 3.06 5.79E-03 1.26 4.45E-03 1.26

1961 4.11E+05 1.32 1.16E+02 1.99 1.03E-03 1.26 2.35E-03 1.26

1962 6.30E+05 1.32 2.35E+02 3.43 9.18E-04 1.26 1.66E-03 1.26

1963 4.36E+05 1.37 1.44E+01 3.48 2.66E-04 1.26 6.03E-03 1.26

1964 6.41E+05 1.37 1.80E+00 2.59 4.79E-04 1.26 1.08E-02 1.26

1965 5.07E+05 1.34 1.69E+00 2.05 1.50E-03 1.26 9.39E-03 1.26

1966 3.70E+05 1.34 2.16E+00 1.44 9.05E-04 1.26 3.79E-03 1.26

1967 3.40E+05 1.35 2.50E+00 2.32 9.49E-04 1.26 3.51E-03 1.26

1968 2.30E+05 1.37 5.55E+00 2.88 5.81E-04 1.26 5.85E-03 1.26

1969 2.55E+05 1.29 9.40E+00 2.82 1.49E-01 1.26 1.44E-02 1.26

1970 3.24E+05 1.30 2.50E+00 1.71 2.74E-03 1.26 3.17E-03 1.26

1971 3.17E+05 1.28 1.80E+00 1.72 2.75E-03 1.26 1.08E-03 1.26

1972 3.39E+05 1.27 4.00E-01 2.56 1.92E-03 1.26 1.55E-03 1.26

1973 4.46E+05 1.26 2.63E-01 2.47 2.23E-03 1.26 1.01E-03 1.26

1974 4.48E+05 1.32 1.11E-01 1.17 6.89E-04 1.26 1.57E-03 1.26

1975 4.32E+05 1.29 7.32E-03 1.22 2.35E-04 1.26 8.54E-04 1.26

1976 3.25E+05 1.26 9.03E-03 1.18 1.54E-03 1.26 9.03E-04 1.26

1977 2.42E+05 1.29 3.79E-03 1.33 4.81E-04 1.26 2.99E-04 1.26

1978 2.20E+05 1.29 3.87E-03 1.23 6.76E-04 1.26 5.83E-04 1.26

1979 2.69E+05 1.29 3.33E-02 3.05 1.53E-04 1.26 4.63E-04 1.26

1980 2.08E+05 1.28 2.51E-02 3.08 3.68E-04 1.26 7.60E-04 1.26

1981 1.81E+05 1.27 2.05E-02 2.96 6.75E-04 1.26 1.20E-03 1.26

1982 1.75E+05 1.27 6.74E-02 3.01 4.51E-04 1.26 1.69E-03 1.26

1983 2.16E+05 1.29 3.18E-02 2.89 2.94E-04 1.26 8.24E-04 1.26

1984 3.02E+05 1.30 7.85E-02 2.90 1.48E-04 1.26 4.19E-04 1.26

1985 2.64E+05 1.29 4.94E-02 2.97 8.15E-05 1.26 4.80E-04 1.26

1986 2.65E+05 1.28 6.56E-03 2.69 2.09E-04 1.26 2.79E-04 1.26

1987 3.29E+05 1.29 4.99E-02 3.08 2.04E-04 1.26 1.54E-03 1.26

1988 3.07E+05 1.29 2.73E-04 3.01 9.02E-05 1.26 2.71E-04 1.26

1989 3.91E+05 1.06 2.09E-05 1.15 1.17E-04 1.26 9.20E-04 1.26

1990 2.27E+05 1.07 6.94E-04 1.99 2.04E-04 1.82 3.24E-04 1.82

1991 5.99E+04 1.15 5.45E-04 1.99 4.59E-04 1.82 1.68E-04 1.82

1992 1.41E+04 1.21 2.92E-06 1.99 5.46E-05 1.82 8.28E-05 1.82

1993 1.89E+04 2.31 3.15E-04 1.99 1.21E-04 1.82 1.01E-04 1.82

1994 4.66E+04 1.56 1.56E-06 1.99 1.39E-04 1.82 1.27E-04 1.82

1995 3.41E+04 1.46 2.82E-04 1.99 9.10E-05 1.82 8.23E-05 1.82

1996 2.67E+04 1.47 5.87E-04 1.99 4.35E-05 1.82 8.88E-05 1.82

1997 1.18E+04 1.49 1.63E-04 1.99 3.91E-06 1.82 6.63E-06 1.82

1998 1.45E+04 1.44 1.13E-06 1.99 1.16E-05 1.82 2.57E-06 1.82

1999 3.24E+02 3.18 1.13E-06 1.99 4.48E-05 1.82 2.73E-06 1.82

2000 2.90E+02 3.08 1.25E-03 1.99 4.37E-05 1.82 5.70E-06 1.82

2001 2.91E+02 2.57 1.48E-03 1.99 9.68E-06 1.82 7.18E-06 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 177 of 248

Table C-13. K Reactor Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 1.24E+04 1.46 1.09E+02 1.88 9.56E-01 1.82 1.06E-01 1.82

1956 6.46E+04 1.32 2.51E+03 1.68 3.73E-03 1.82 4.31E-02 1.82

1957 2.54E+05 1.27 5.26E+02 2.03 5.00E-03 1.82 7.18E-04 1.82

1958 2.67E+05 1.25 6.93E+01 2.95 2.45E-03 1.26 5.00E-04 1.26

1959 3.03E+05 1.31 3.26E+02 2.07 2.79E-03 1.26 2.30E-03 1.26

1960 4.56E+05 1.30 3.11E+01 3.01 8.37E-03 1.26 6.43E-03 1.26

1961 3.11E+05 1.31 1.73E+02 2.03 1.36E-03 1.26 3.25E-03 1.26

1962 5.37E+05 1.32 1.84E+01 3.29 1.12E-03 1.26 2.30E-03 1.26

1963 1.57E+05 3.30 7.31E+00 3.41 3.59E-04 1.26 8.13E-03 1.26

1964 7.72E+05 1.37 2.32E+00 2.51 5.51E-04 1.26 1.43E-02 1.26

1965 8.84E+05 1.32 3.41E+00 2.40 1.61E-03 1.26 1.36E-02 1.26

1966 6.70E+05 1.32 3.55E+00 1.68 9.83E-04 1.26 5.37E-03 1.26

1967 8.13E+05 1.34 1.43E+01 2.93 1.04E-03 1.26 5.00E-03 1.26

1968 7.13E+05 1.32 1.65E+01 3.10 6.58E-04 1.26 6.63E-03 1.26

1969 6.13E+05 1.30 2.67E+01 3.00 1.49E-01 1.26 1.73E-02 1.26

1970 7.64E+05 1.30 8.19E+00 2.61 2.87E-03 1.26 4.30E-03 1.26

1971 8.22E+05 1.29 2.21E+00 1.37 2.86E-03 1.26 1.30E-03 1.26

1972 7.37E+05 1.27 5.26E-01 2.48 1.96E-03 1.26 1.97E-03 1.26

1973 6.52E+05 1.28 4.09E-01 2.52 2.26E-03 1.26 1.31E-03 1.26

1974 5.39E+05 1.31 1.63E-01 1.21 7.51E-04 1.26 2.27E-03 1.26

1975 3.92E+05 1.31 1.51E-02 1.92 2.40E-04 1.26 1.24E-03 1.26

1976 4.12E+05 1.25 1.32E-02 1.21 1.59E-03 1.26 1.30E-03 1.26

1977 2.72E+05 1.28 5.98E-03 1.53 4.94E-04 1.26 4.00E-04 1.26

1978 2.84E+05 1.29 7.81E-03 1.89 6.89E-04 1.26 7.90E-04 1.26

1979 2.87E+05 1.29 8.71E-03 1.69 1.75E-04 1.26 6.30E-04 1.26

1980 3.37E+05 1.27 4.37E-03 2.35 3.88E-04 1.26 1.04E-03 1.26

1981 3.76E+05 1.27 5.41E-03 1.85 7.15E-04 1.26 1.55E-03 1.26

1982 3.72E+05 1.28 1.69E-02 2.25 5.79E-04 1.26 2.38E-03 1.26

1983 5.31E+05 1.31 8.61E-03 1.63 3.28E-04 1.26 1.17E-03 1.26

1984 3.82E+05 1.28 2.73E-02 1.54 1.84E-04 1.26 5.65E-04 1.26

1985 3.60E+05 1.28 9.56E-03 2.20 1.06E-04 1.26 6.53E-04 1.26

1986 2.65E+05 1.29 9.86E-03 2.70 2.20E-04 1.26 3.81E-04 1.26

1987 2.37E+05 1.29 5.24E-03 2.79 2.17E-04 1.26 2.22E-03 1.26

1988 2.45E+05 1.29 3.25E-04 2.97 1.18E-04 1.26 3.81E-04 1.26

1989 1.17E+05 1.09 3.04E-05 1.15 1.58E-04 1.26 1.32E-03 1.26

1990 1.63E+05 1.01 1.01E-03 1.99 2.44E-04 1.82 3.87E-04 1.82

1991 2.08E+05 0.98 7.91E-04 1.99 5.48E-04 1.82 2.01E-04 1.82

1992 1.72E+05 1.07 2.92E-06 1.99 6.41E-05 1.82 1.18E-04 1.82

1993 1.55E+05 5.03 3.76E-04 1.99 1.43E-04 1.82 1.43E-04 1.82

1994 5.83E+04 4.68 1.56E-06 1.99 1.52E-04 1.82 1.79E-04 1.82

1995 1.65E+04 3.88 3.54E-04 1.99 1.17E-04 1.82 1.19E-04 1.82

1996 1.31E+04 4.27 8.00E-04 1.99 5.14E-05 1.82 1.29E-04 1.82

1997 1.17E+04 3.93 1.63E-04 1.99 4.40E-06 1.82 7.29E-06 1.82

1998 1.00E+04 3.58 1.64E-06 1.99 1.29E-05 1.82 3.67E-06 1.82

1999 7.01E+03 4.32 1.64E-06 1.99 4.75E-05 1.82 3.78E-06 1.82

2000 5.02E+03 4.40 1.31E-03 1.99 4.72E-05 1.82 8.01E-06 1.82

2001 3.32E+03 4.19 1.59E-03 1.99 9.93E-06 1.82 1.04E-05 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 178 of 248

Table C-14. R Reactor Area annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 4.12E+03 2.09 1.31E+02 1.86 9.76E-01 1.82 1.46E-01 1.82

1956 4.14E+04 1.75 3.44E+03 1.68 3.61E-03 1.82 4.05E-02 1.82

1957 2.05E+05 1.44 8.55E+02 1.98 4.89E-03 1.82 7.03E-04 1.82

1958 2.89E+05 1.57 7.54E+01 2.58 2.36E-03 1.26 5.92E-04 1.26

1959 1.49E+05 1.94 3.26E+02 2.05 2.75E-03 1.26 2.09E-03 1.26

1960 1.47E+05 2.27 3.55E+01 3.13 7.55E-03 1.26 5.82E-03 1.26

1961 1.20E+05 1.84 1.65E+02 2.12 1.47E-03 1.26 3.20E-03 1.26

1962 1.73E+05 2.40 2.58E+01 3.34 1.39E-03 1.26 2.26E-03 1.26

1963 1.63E+05 2.37 1.01E+02 3.51 3.71E-04 1.26 8.41E-03 1.26

1964 2.16E+05 2.46 1.72E+00 2.37 7.54E-04 1.26 1.54E-02 1.26

1965 7.96E+04 1.40 1.66E+00 1.60 2.47E-03 1.26 1.23E-02 1.26

1966 6.91E+04 1.39 2.63E+00 1.27 1.48E-03 1.26 5.03E-03 1.26

1967 7.09E+04 1.38 2.24E+00 1.90 1.54E-03 1.26 4.65E-03 1.26

1968 8.24E+04 1.35 2.39E+00 1.87 9.25E-04 1.26 9.31E-03 1.26

1969 5.65E+04 1.26 3.69E+00 1.83 2.55E-01 1.26 2.19E-02 1.26

1970 6.08E+04 1.25 2.84E+00 1.42 4.58E-03 1.26 4.41E-03 1.26

1971 7.88E+04 1.24 1.92E+00 1.27 4.61E-03 1.26 1.65E-03 1.26

1972 9.58E+04 1.23 2.57E-01 1.71 3.25E-03 1.26 2.26E-03 1.26

1973 6.93E+04 1.24 1.57E-01 1.45 3.81E-03 1.26 1.47E-03 1.26

1974 1.07E+05 1.51 1.45E-01 1.17 1.12E-03 1.26 2.04E-03 1.26

1975 6.30E+04 1.38 9.52E-03 1.19 4.00E-04 1.26 1.11E-03 1.26

1976 3.50E+04 1.26 1.17E-02 1.18 2.59E-03 1.26 1.18E-03 1.26

1977 4.39E+04 1.26 4.88E-03 1.29 8.12E-04 1.26 4.21E-04 1.26

1978 4.04E+04 1.28 4.99E-03 1.20 1.15E-03 1.26 8.10E-04 1.26

1979 3.95E+04 1.25 8.11E-03 1.75 2.43E-04 1.26 6.40E-04 1.26

1980 3.45E+04 1.25 3.36E-03 2.18 6.12E-04 1.26 1.05E-03 1.26

1981 4.37E+04 1.25 4.86E-03 1.85 1.12E-03 1.26 1.74E-03 1.26

1982 4.92E+04 1.24 1.19E-02 1.96 6.56E-04 1.26 2.26E-03 1.26

1983 7.19E+04 1.25 8.12E-03 1.72 4.72E-04 1.26 1.09E-03 1.26

1984 9.04E+04 1.24 2.48E-02 1.57 2.21E-04 1.26 5.83E-04 1.26

1985 7.36E+04 1.25 7.28E-03 2.01 1.17E-04 1.26 6.63E-04 1.26

1986 4.97E+04 1.24 2.45E-03 1.62 3.47E-04 1.26 3.85E-04 1.26

1987 7.25E+04 1.32 3.90E-03 2.72 3.38E-04 1.26 2.01E-03 1.26

1988 4.76E+04 1.25 1.54E-04 2.78 1.29E-04 1.26 3.63E-04 1.26

1989 3.82E+04 1.17 2.72E-05 1.15 1.64E-04 1.26 1.21E-03 1.26

1990 3.27E+04 1.18 9.02E-04 1.99 3.14E-04 1.82 4.98E-04 1.82

1991 2.19E+04 1.40 7.08E-04 1.99 7.05E-04 1.82 2.58E-04 1.82

1992 1.85E+04 1.21 5.01E-06 1.99 8.49E-05 1.82 1.10E-04 1.82

1993 2.10E+04 2.03 4.83E-04 1.99 1.87E-04 1.82 1.34E-04 1.82

1994 1.86E+04 1.66 2.67E-06 1.99 2.25E-04 1.82 1.70E-04 1.82

1995 1.15E+04 1.50 4.17E-04 1.99 1.32E-04 1.82 1.07E-04 1.82

1996 6.34E+03 1.57 8.10E-04 1.99 6.72E-05 1.82 1.15E-04 1.82

1997 6.79E+03 1.51 2.79E-04 1.99 6.25E-06 1.82 1.08E-05 1.82

1998 7.56E+03 1.47 1.46E-06 1.99 1.86E-05 1.82 3.38E-06 1.82

1999 3.99E+02 2.51 1.46E-06 1.99 7.44E-05 1.82 3.72E-06 1.82

2000 3.39E+02 2.46 2.08E-03 1.99 7.16E-05 1.82 7.66E-06 1.82

2001 3.67E+02 2.07 2.45E-03 1.99 1.63E-05 1.82 9.37E-06 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 179 of 248

Table C-15. Rail Yard annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 2.80E+03 1.53 6.67E+01 1.82 5.47E-01 1.82 7.13E-02 1.82

1956 2.55E+04 1.32 1.68E+03 1.67 2.08E-03 1.82 2.37E-02 1.82

1957 1.01E+05 1.27 3.74E+02 1.92 2.80E-03 1.82 4.02E-04 1.82

1958 1.35E+05 1.26 2.84E+01 2.37 1.36E-03 1.26 3.09E-04 1.26

1959 6.79E+04 1.34 1.72E+02 1.92 1.56E-03 1.26 1.24E-03 1.26

1960 7.16E+04 1.39 9.78E+00 2.60 4.51E-03 1.26 3.47E-03 1.26

1961 6.79E+04 1.36 8.49E+01 1.88 8.02E-04 1.26 1.83E-03 1.26

1962 9.11E+04 1.42 2.74E+01 3.38 7.09E-04 1.26 1.29E-03 1.26

1963 7.50E+04 1.51 6.58E+00 3.45 2.07E-04 1.26 4.68E-03 1.26

1964 1.09E+05 1.55 1.05E+00 2.38 3.69E-04 1.26 8.39E-03 1.26

1965 6.82E+04 1.46 1.12E+00 1.83 1.15E-03 1.26 7.31E-03 1.26

1966 5.45E+04 1.46 1.62E+00 1.35 6.95E-04 1.26 2.94E-03 1.26

1967 5.50E+04 1.45 1.64E+00 2.15 7.29E-04 1.26 2.73E-03 1.26

1968 5.36E+04 1.43 1.83E+00 2.23 4.47E-04 1.26 4.50E-03 1.26

1969 4.23E+04 1.28 2.89E+00 2.19 1.14E-01 1.26 1.11E-02 1.26

1970 4.90E+04 1.28 1.77E+00 1.51 2.10E-03 1.26 2.46E-03 1.26

1971 5.70E+04 1.26 1.18E+00 1.35 2.11E-03 1.26 8.34E-04 1.26

1972 6.41E+04 1.25 1.74E-01 1.93 1.47E-03 1.26 1.20E-03 1.26

1973 5.74E+04 1.26 1.07E-01 1.72 1.71E-03 1.26 7.85E-04 1.26

1974 7.39E+04 1.44 8.65E-02 1.17 5.29E-04 1.26 1.22E-03 1.26

1975 5.33E+04 1.34 5.68E-03 1.21 1.80E-04 1.26 6.65E-04 1.26

1976 3.40E+04 1.27 7.02E-03 1.18 1.18E-03 1.26 7.03E-04 1.26

1977 3.49E+04 1.28 2.93E-03 1.32 3.69E-04 1.26 2.32E-04 1.26

1978 3.17E+04 1.29 3.00E-03 1.22 5.18E-04 1.26 4.53E-04 1.26

1979 3.45E+04 1.28 6.99E-03 2.30 1.18E-04 1.26 3.59E-04 1.26

1980 2.86E+04 1.27 3.81E-03 2.68 2.82E-04 1.26 5.90E-04 1.26

1981 3.03E+04 1.27 4.27E-03 2.32 5.18E-04 1.26 9.31E-04 1.26

1982 3.33E+04 1.26 1.18E-02 2.48 3.49E-04 1.26 1.31E-03 1.26

1983 4.58E+04 1.27 6.91E-03 2.20 2.26E-04 1.26 6.41E-04 1.26

1984 5.88E+04 1.27 1.96E-02 2.05 1.15E-04 1.26 3.25E-04 1.26

1985 5.04E+04 1.27 7.85E-03 2.52 6.31E-05 1.26 3.72E-04 1.26

1986 3.76E+04 1.27 1.89E-03 1.99 1.60E-04 1.26 2.17E-04 1.26

1987 4.99E+04 1.32 6.10E-03 2.96 1.57E-04 1.26 1.20E-03 1.26

1988 3.94E+04 1.28 1.53E-04 2.94 6.99E-05 1.26 2.11E-04 1.26

1989 3.99E+04 1.12 1.63E-05 1.15 9.10E-05 1.26 7.16E-04 1.26

1990 2.80E+04 1.13 5.41E-04 1.99 1.58E-04 1.82 2.50E-04 1.82

1991 1.39E+04 1.32 4.24E-04 1.99 3.54E-04 1.82 1.30E-04 1.82

1992 9.53E+03 1.21 2.24E-06 1.99 4.21E-05 1.82 6.44E-05 1.82

1993 1.07E+04 2.44 2.43E-04 1.99 9.32E-05 1.82 7.83E-05 1.82

1994 1.08E+04 1.80 1.19E-06 1.99 1.07E-04 1.82 9.85E-05 1.82

1995 6.81E+03 1.55 2.18E-04 1.99 7.04E-05 1.82 6.40E-05 1.82

1996 4.23E+03 1.63 4.55E-04 1.99 3.35E-05 1.82 6.91E-05 1.82

1997 3.53E+03 1.60 1.25E-04 1.99 3.01E-06 1.82 5.10E-06 1.82

1998 3.98E+03 1.53 8.77E-07 1.99 8.92E-06 1.82 2.00E-06 1.82

1999 2.27E+02 3.06 8.77E-07 1.99 3.44E-05 1.82 2.12E-06 1.82

2000 2.01E+02 2.96 9.56E-04 1.99 3.35E-05 1.82 4.43E-06 1.82

2001 2.04E+02 2.47 1.14E-03 1.99 7.42E-06 1.82 5.59E-06 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 180 of 248

Table C-16. Average site-widea annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Plutonium Uranium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 2.86E+04 1.33E+00 4.31E+02 1.80E+00 3.64E+00 1.82E+00 4.53E-01 1.82E+00

1956 1.61E+05 1.27E+00 1.06E+04 1.66E+00 1.40E-02 1.82E+00 1.60E-01 1.82E+00

1957 6.29E+05 1.24E+00 2.26E+03 1.89E+00 1.88E-02 1.82E+00 2.70E-03 1.82E+00

1958 8.42E+05 1.23E+00 1.51E+02 2.16E+00 9.15E-03 1.26E+00 2.01E-03 1.26E+00

1959 4.15E+05 1.26E+00 1.12E+03 1.87E+00 1.05E-02 1.26E+00 8.43E-03 1.26E+00

1960 3.73E+05 1.31E+00 5.04E+01 2.24E+00 3.06E-02 1.26E+00 2.35E-02 1.26E+00

1961 3.40E+05 1.28E+00 5.61E+02 1.83E+00 5.30E-03 1.26E+00 1.22E-02 1.26E+00

1962 4.18E+05 1.35E+00 3.20E+01 3.13E+00 4.58E-03 1.26E+00 8.65E-03 1.26E+00

1963 3.82E+05 1.43E+00 1.48E+01 3.32E+00 1.37E-03 1.26E+00 3.11E-02 1.26E+00

1964 5.54E+05 1.44E+00 4.76E+00 1.93E+00 2.35E-03 1.26E+00 5.55E-02 1.26E+00

1965 3.27E+05 1.38E+00 6.48E+00 1.52E+00 7.19E-03 1.26E+00 4.96E-02 1.26E+00

1966 3.01E+05 1.35E+00 1.06E+01 1.25E+00 4.36E-03 1.26E+00 1.99E-02 1.26E+00

1967 3.04E+05 1.35E+00 8.67E+00 1.82E+00 4.58E-03 1.26E+00 1.85E-02 1.26E+00

1968 3.26E+05 1.34E+00 9.19E+00 1.76E+00 2.83E-03 1.26E+00 2.85E-02 1.26E+00

1969 2.34E+05 1.25E+00 1.41E+01 1.72E+00 6.99E-01 1.26E+00 7.14E-02 1.26E+00

1970 2.48E+05 1.24E+00 1.15E+01 1.41E+00 1.31E-02 1.26E+00 1.64E-02 1.26E+00

1971 2.95E+05 1.24E+00 7.75E+00 1.26E+00 1.31E-02 1.26E+00 5.38E-03 1.26E+00

1972 3.40E+05 1.23E+00 1.02E+00 1.65E+00 9.08E-03 1.26E+00 7.84E-03 1.26E+00

1973 2.64E+05 1.23E+00 6.19E-01 1.38E+00 1.05E-02 1.26E+00 5.16E-03 1.26E+00

1974 3.52E+05 1.47E+00 5.87E-01 1.17E+00 3.32E-03 1.26E+00 8.30E-03 1.26E+00

1975 2.15E+05 1.35E+00 3.82E-02 1.19E+00 1.11E-03 1.26E+00 4.52E-03 1.26E+00

1976 1.44E+05 1.24E+00 4.77E-02 1.18E+00 7.29E-03 1.26E+00 4.77E-03 1.26E+00

1977 1.65E+05 1.25E+00 1.97E-02 1.29E+00 2.28E-03 1.26E+00 1.54E-03 1.26E+00

1978 1.53E+05 1.27E+00 2.03E-02 1.20E+00 3.20E-03 1.26E+00 3.02E-03 1.26E+00

1979 1.45E+05 1.26E+00 2.90E-02 1.48E+00 7.50E-04 1.26E+00 2.40E-03 1.26E+00

1980 1.38E+05 1.24E+00 1.05E-02 1.82E+00 1.76E-03 1.26E+00 3.94E-03 1.26E+00

1981 1.61E+05 1.24E+00 1.74E-02 1.62E+00 3.23E-03 1.26E+00 6.12E-03 1.26E+00

1982 1.78E+05 1.24E+00 4.02E-02 1.66E+00 2.29E-03 1.26E+00 8.86E-03 1.26E+00

1983 2.49E+05 1.25E+00 2.94E-02 1.49E+00 1.43E-03 1.26E+00 4.33E-03 1.26E+00

1984 3.07E+05 1.25E+00 9.25E-02 1.36E+00 7.44E-04 1.26E+00 2.16E-03 1.26E+00

1985 2.74E+05 1.25E+00 2.35E-02 1.67E+00 4.15E-04 1.26E+00 2.48E-03 1.26E+00

1986 1.74E+05 1.24E+00 9.31E-03 1.49E+00 9.97E-04 1.26E+00 1.45E-03 1.26E+00

1987 2.19E+05 1.31E+00 9.22E-03 2.42E+00 9.77E-04 1.26E+00 8.14E-03 1.26E+00

1988 1.44E+05 1.25E+00 4.91E-04 2.67E+00 4.61E-04 1.26E+00 1.42E-03 1.26E+00

1989 1.06E+05 1.18E+00 1.11E-04 1.15E+00 6.05E-04 1.26E+00 4.85E-03 1.26E+00

1990 9.48E+04 1.17E+00 3.68E-03 1.99E+00 1.01E-03 1.82E+00 1.61E-03 1.82E+00

1991 6.96E+04 1.35E+00 2.89E-03 1.99E+00 2.27E-03 1.82E+00 8.34E-04 1.82E+00

1992 5.85E+04 1.20E+00 1.37E-05 1.99E+00 2.69E-04 1.82E+00 4.35E-04 1.82E+00

1993 6.50E+04 2.46E+00 1.56E-03 1.99E+00 5.97E-04 1.82E+00 5.30E-04 1.82E+00

1994 5.36E+04 1.87E+00 7.34E-06 1.99E+00 6.70E-04 1.82E+00 6.64E-04 1.82E+00

1995 3.27E+04 1.57E+00 1.42E-03 1.99E+00 4.62E-04 1.82E+00 4.35E-04 1.82E+00

1996 1.84E+04 1.69E+00 3.04E-03 1.99E+00 2.15E-04 1.82E+00 4.70E-04 1.82E+00

1997 1.98E+04 1.60E+00 7.67E-04 1.99E+00 1.90E-05 1.82E+00 3.21E-05 1.82E+00

1998 2.18E+04 1.52E+00 5.97E-06 1.99E+00 5.63E-05 1.82E+00 1.35E-05 1.82E+00

1999 2.01E+03 2.54E+00 5.97E-06 1.99E+00 2.14E-04 1.82E+00 1.42E-05 1.82E+00

2000 1.91E+03 2.36E+00 5.94E-03 1.99E+00 2.10E-04 1.82E+00 2.99E-05 1.82E+00

2001 1.92E+03 1.98E+00 7.12E-03 1.99E+00 4.58E-05 1.82E+00 3.80E-05 1.82E+00

a. Values are for plant perimeter.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 181 of 248

Table C-17. Maximum site-wide annual intakes and geometric standard deviations (Bq per year).

Year

Tritium Iodine-131 Uranium Plutonium

50th-percentile GSD 50th-percentile GSD 50th-percentile GSD 50th-percentile GSD

1955 2.12E+05 1.18 2.06E+03 1.79 2.17E+00 1.82 2.46E+01 1.82

1956 1.02E+06 1.26 5.03E+04 1.66 1.22E+00 1.82 1.02E-01 1.82

1957 2.88E+06 1.24 1.50E+04 1.87 1.94E-02 1.82 1.35E-01 1.82

1958 3.93E+06 1.20 9.41E+02 1.98 1.01E-02 1.26 6.74E-02 1.26

1959 2.12E+06 1.22 7.58E+03 1.84 6.73E-02 1.26 7.48E-02 1.26

1960 1.68E+06 1.22 3.57E+02 2.00 1.89E-01 1.26 2.47E-01 1.26

1961 1.72E+06 1.20 4.50E+03 1.80 8.75E-02 1.26 3.26E-02 1.26

1962 2.01E+06 1.24 2.35E+02 3.43 6.20E-02 1.26 2.06E-02 1.26

1963 1.97E+06 1.22 1.01E+02 3.51 2.05E-01 1.26 9.06E-03 1.26

1964 2.54E+06 1.28 2.69E+01 2.09 3.39E-01 1.26 1.10E-02 1.26

1965 1.00E+06 1.23 4.62E+01 1.18 3.99E-01 1.26 4.19E-02 1.26

1966 1.01E+06 1.23 8.21E+01 1.12 1.53E-01 1.26 2.45E-02 1.26

1967 9.22E+05 1.24 5.16E+01 1.34 1.43E-01 1.26 2.52E-02 1.26

1968 1.27E+06 1.22 5.91E+01 1.43 1.41E-01 1.26 1.40E-02 1.26

1969 8.03E+05 1.21 9.07E+01 1.42 3.07E-01 1.26 4.82E+00 1.26

1970 8.58E+05 1.21 8.60E+01 1.17 1.10E-01 1.26 8.10E-02 1.26

1971 1.10E+06 1.21 6.09E+01 1.16 2.31E-02 1.26 8.23E-02 1.26

1972 1.47E+06 1.20 6.88E+00 1.21 4.24E-02 1.26 5.97E-02 1.26

1973 1.00E+06 1.19 4.63E+00 1.17 2.91E-02 1.26 7.10E-02 1.26

1974 1.68E+06 1.53 4.72E+00 1.15 6.73E-02 1.26 1.84E-02 1.26

1975 8.69E+05 1.39 2.99E-01 1.16 3.67E-02 1.26 7.37E-03 1.26

1976 4.69E+05 1.20 3.85E-01 1.16 3.81E-02 1.26 4.66E-02 1.26

1977 5.77E+05 1.21 1.50E-01 1.17 9.85E-03 1.26 1.48E-02 1.26

1978 5.58E+05 1.24 1.62E-01 1.16 2.02E-02 1.26 2.11E-02 1.26

1979 5.06E+05 1.20 2.08E-01 1.17 1.63E-02 1.26 3.61E-03 1.26

1980 4.50E+05 1.19 6.24E-02 1.77 2.71E-02 1.26 1.07E-02 1.26

1981 7.13E+05 1.21 1.18E-01 1.19 3.49E-02 1.26 1.94E-02 1.26

1982 7.41E+05 1.20 2.65E-01 1.31 6.68E-02 1.26 1.27E-02 1.26

1983 1.16E+06 1.22 2.11E-01 1.17 3.35E-02 1.26 7.37E-03 1.26

1984 1.41E+06 1.21 7.01E-01 1.16 1.43E-02 1.26 3.63E-03 1.26

1985 1.05E+06 1.21 1.52E-01 1.50 1.69E-02 1.26 2.43E-03 1.26

1986 8.77E+05 1.19 6.60E-02 1.17 9.88E-03 1.26 6.06E-03 1.26

1987 1.28E+06 1.30 4.99E-02 3.08 6.55E-02 1.26 5.80E-03 1.26

1988 7.17E+05 1.21 3.42E-03 3.16 1.07E-02 1.26 2.72E-03 1.26

1989 4.38E+05 1.21 9.05E-04 1.15 3.84E-02 1.26 3.96E-03 1.26

1990 4.33E+05 1.20 3.00E-02 1.99 6.58E-03 1.82 4.15E-03 1.82

1991 3.28E+05 1.46 2.36E-02 1.99 3.42E-03 1.82 9.31E-03 1.82

1992 3.06E+05 1.21 9.47E-05 1.99 3.37E-03 1.82 1.18E-03 1.82

1993 3.54E+05 3.05 6.40E-03 1.99 4.09E-03 1.82 2.56E-03 1.82

1994 3.16E+05 1.68 5.06E-05 1.99 5.00E-03 1.82 3.66E-03 1.82

1995 1.95E+05 1.41 7.31E-03 1.99 3.52E-03 1.82 2.58E-03 1.82

1996 1.01E+05 1.42 2.08E-02 1.99 3.84E-03 1.82 9.11E-04 1.82

1997 1.21E+05 1.41 5.27E-03 1.99 1.74E-04 1.82 9.60E-05 1.82

1998 1.35E+05 1.41 4.88E-05 1.99 1.06E-04 1.82 2.92E-04 1.82

1999 7.01E+03 4.32 4.88E-05 1.99 1.02E-04 1.82 1.29E-03 1.82

2000 5.80E+03 4.03 3.66E-02 1.99 2.23E-04 1.82 1.20E-03 1.82

2001 5.78E+03 2.99 4.14E-02 1.99 3.08E-04 1.82 2.98E-04 1.82
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 182 of 248

Table C-18. Annual intakes for F and H Areas (Bq per year).

Location

90Sr 137Cs 238Pu 239Pu

50th GSD 50th GSD 50th GSD 50th GSD

F Area 2.46E-04 1.81 1.19E-02 1.06 4.59E-03 1.04 6.04E-03 1.03

H Area 3.29E-04 1.63 1.50E-02 1.04 2.42E-04 1.09 6.30E-04 1.05

Table C-19. Ambient radiation levels for selected SRS locations (mrem/year).

A Area C Area D-Area E Area F Area H Area S Area Z Area

Year 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD

1954 73 62 58 62

1955 78 69 75 75 74

1956 75 74 69 77 90

1957 96 86 80 90 107

1958 102 103 97 98 120

1959 59 56 54 87 77

1960 61 41 40 182 81

1961 44 41 38 49 57

1962 64 47 50 49 61

1963 68 50 57 60 81

1964 35 47 30 57 75

1965 32 47 28 48 62

1966 24 22 49 105

1967 23 20 38 89

1968 18 18 40 47

1969 20 22 51 67

1970 19 13 56 81

1971 19 14 67 47

1972 22 17 12 76 47

1973 26 1.02 19 1.13 16 1.00 79 1.77 55 1.87

1974 22 1.05 17 1.06 15 1.31 71 1.06 56 1.05

1975 31 1.10 17 1.08 16 1.06 68 1.07 62 1.08

1976 43 1.11 17 1.08 16 1.06 54 1.10 53 1.07

1977 41 1.17 17 1.06 17 1.06 72 1.09 67 1.10

1978 28 1.07 17 1.06 17 1.06 49 1.08 40 1.09

1979 34 1.06 24 1.06 14 1.07 52 1.08 45 1.09

1980 27 1.06 16 1.06 15 1.07 35 1.04 28 1.09

1981 32 1.08 19 1.08 18 1.05 40 1.06 59 1.05

1982 27 1.04 18 1.08 17 1.06 27 1.06 31 1.10 52 1.10

1983 22 1.09 17 1.06 16 1.06 28 1.05 28 1.10 55 1.09

1984 22 1.07 17 1.06 17 1.06 34 1.06 27 1.11 74 1.07 20 1.05

1985 21 1.07 15 1.07 16 1.06 36 1.06 26 1.08 91 1.17 28 1.07 16 1.06

1986 21 1.09 15 1.07 15 1.07 24 1.06 31 1.08 47 1.09 26 1.08 15 1.07

1987 22 1.07 15 1.07 16 1.06 22 1.07 22 1.09 42 1.08 23 1.06 15 1.07

1988 21 1.14 17 1.12 16 1.12 21 1.14 24 1.18 40 1.40 23 1.13 17 1.12

1989 18 1.18 12 1.15 13 1.15 18 1.16 20 1.19 27 1.29 20 1.15 12 1.17

1990 15 1.14 10 1.12 10 1.15 20 1.17 18 1.20 23 1.25 18 1.13 11 1.12

1991 15 1.14 10 1.12 10 1.15 22 1.25 17 1.18 20 1.18 18 1.13 10 1.12

1992 14 1.15 9 1.11 9 1.11 15 1.11 15 1.11 17 1.11 16 1.11 10 1.13

1993 20 1.11 16 1.11 15 1.11 22 1.11 21 1.11 23 1.11 23 1.11 16 1.11

1994 20 1.11 17 1.11 16 1.11 22 1.11 22 1.11 24 1.11 24 1.11 17 1.11

1995 22 1.11 18 1.11 17 1.11 24 1.11 23 1.11 25 1.11 26 1.11 18 1.11

1996 21 1.11 18 1.11 17 1.11 23 1.11 24 1.11 25 1.11 25 1.11 18 1.11

1997 21 1.11 23 1.11 23 1.11 26 1.11 26 1.11 18 1.11

1998 22 1.11 27 1.11 24 1.11 26 1.11 27 1.11 19 1.11

1999 18 1.11 28 1.11

2000 19 1.11 29 1.11

2001 17 1.11 27 1.11
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 183 of 248

Table C-19 (Continued).

Forest 

servicea Central shops L-Area P Area K Area R Area Rail yard Site perimeter

Year 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD

1954 67 77

1955 84 83 80 75

1956 69 77 70 75 72 69

1957 82 95 92 91 93 82

1958 90 110 104 106 103 90

1959 41 61 56 55 64 41

1960 30 48 42 47 47 30

1961 26 38 36 43 32 26

1962 37 50 49 49 49 37

1963 36 59 57 57 58 36

1964 32 52 42 45 44 32

1965 28 28

1966 21 21

1967 19 19

1968 22 22

1969 22 22

1970 15 15

1971 14 14

1972 12 18 32 12

1973 17 1.06 22 1.11 35 1.66 17 1.06

1974 13 1.07 19 1.10 28 1.10 13 1.07

1975 13 1.07 22 1.07 20 1.07 30 1.10 13 1.07

1976 14 1.10 20 1.07 18 1.11 27 1.09 14 1.10

1977 14 1.07 28 1.04 17 1.06 27 1.09 14 1.07

1978 15 1.07 46 1.11 18 1.05 26 1.10 15 1.07

1979 16 1.06 28 1.05 25 1.06 32 1.09 16 1.06

1980 13 1.07 18 1.08 17 1.06 20 1.07 13 1.07

1981 14 1.20 23 1.06 18 1.05 19 1.05 24 1.06 14 1.20

1982 16 1.06 20 1.05 17 1.06 19 1.08 24 1.08 16 1.06

1983 15 1.07 20 1.05 17 1.06 19 1.08 23 1.08 15 1.07

1984 14 1.07 29 1.10 17 1.12 18 1.05 25 1.08 16 1.06 20 1.05 14 1.07

1985 15 1.07 30 1.08 17 1.06 17 1.06 23 1.08 17 1.06 19 1.08 15 1.07

1986 17 1.06 26 1.06 17 1.08 16 1.06 22 1.09 16 1.06 19 1.05 17 1.06

1987 17 1.06 27 1.07 19 1.05 18 1.05 22 1.07 15 1.07 21 1.05 17 1.06

1988 14 1.17 28 1.34 18 1.13 17 1.11 22 1.17 17 1.11 21 1.12 14 1.17

1989 12 1.20 40 1.31 15 1.19 15 1.13 22 1.22 15 1.13 17 1.14 12 1.20

1990 10 1.25 37 1.31 12 1.13 12 1.12 15 1.19 12 1.14 15 1.12 10 1.25

1991 10 1.34 18 1.16 12 1.12 12 1.12 13 1.14 11 1.14 14 1.11 10 1.34

1992 8 1.11 16 1.11 10 1.11 10 1.11 12 1.11 10 1.11 12 1.11 8 1.11

1993 16 1.11 25 1.11 16 1.11 16 1.11 18 1.11 16 1.11 19 1.11 16 1.11

1994 15 1.11 25 1.11 17 1.11 17 1.11 17 1.11 17 1.11 20 1.11 15 1.11

1995 17 1.11 27 1.11 18 1.11 18 1.11 18 1.11 18 1.11 22 1.11 17 1.11

1996 19 1.11 25 1.11 18 1.11 18 1.11 17 1.11 18 1.11 21 1.11 19 1.11

1997 16 1.11 26 1.11 18 1.11 18 1.11 18 1.11 18 1.11 21 1.11 16 1.11

1998 15 1.11 30 1.11 19 1.11 19 1.11 19 1.11 18 1.11 22 1.11 15 1.11

1999 18 1.11 18 1.11

2000 18 1.11 18 1.11

2001 17 1.11 17 1.11

a. Source: Plant perimeter values – SRS Annual Environmental Reports (see reference section).
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 184 of 248

Table C-20. Argon-41 50th-percentile dosea to skinb and bone surfacesc resulting from reactor 

releases (mrem per year) and geometric standard deviation.

Year

A 

Area GSD

C 

Area GSD

D

Area GSD

FArea GSD BG GSDH 

Area GSD

S 

Area GSD

Z 

Area GSD

R 

Area GSD

1955 0.17 1.15 1.81 1.15 0.30 1.15 0.39 1.15 0.42 1.15 0.42 1.15 0.36 1.15 0.35 1.15 2.14 1.15

1956 0.25 1.15 4.08 1.15 0.48 1.15 0.67 1.15 0.68 1.15 0.64 1.15 0.54 1.15 0.51 1.15 2.61 1.15

1957 0.31 1.15 4.22 1.15 0.56 1.15 0.77 1.15 0.79 1.15 0.78 1.15 0.66 1.15 0.63 1.15 4.09 1.15

1958 0.36 1.15 5.04 1.15 0.66 1.15 0.91 1.15 0.94 1.15 0.92 1.15 0.78 1.15 0.75 1.15 4.90 1.15

1959 0.46 1.15 6.67 1.15 0.85 1.15 1.17 1.15 1.21 1.15 1.17 1.15 0.99 1.15 0.94 1.15 5.80 1.15

1960 0.45 1.15 6.61 1.15 0.81 1.15 1.15 1.15 1.19 1.15 1.18 1.15 1.00 1.15 0.96 1.15 7.12 1.15

1961 0.45 1.15 6.61 1.15 0.81 1.15 1.15 1.15 1.19 1.15 1.16 1.15 0.98 1.15 0.94 1.15 6.46 1.15

1962 0.48 1.15 7.33 1.15 0.85 1.15 1.23 1.15 1.27 1.15 1.24 1.15 1.05 1.15 1.00 1.15 7.18 1.15

1963 0.50 1.15 7.41 1.15 0.90 1.15 1.27 1.15 1.31 1.15 1.28 1.15 1.08 1.15 1.04 1.15 7.22 1.15

1964 0.42 1.15 6.50 1.15 0.75 1.15 1.09 1.15 1.13 1.15 1.11 1.15 0.94 1.15 0.90 1.15 7.04 1.15

1965 0.39 1.15 5.76 1.15 0.70 1.15 0.99 1.15 1.03 1.15 1.01 1.15 0.86 1.15 0.82 1.15 6.28 1.15

1966 0.31 1.15 5.62 1.15 0.63 1.15 0.84 1.15 0.84 1.15 0.74 1.15 0.61 1.15 0.56 1.15 0.72 1.15

1967 0.36 1.15 6.43 1.15 0.72 1.15 0.96 1.15 0.96 1.15 0.85 1.15 0.70 1.15 0.64 1.15 0.82 1.15

1968 0.26 1.15 6.11 1.15 0.53 1.15 0.77 1.15 0.74 1.15 0.62 1.15 0.51 1.15 0.46 1.15 0.57 1.15

1969 0.17 1.15 3.84 1.15 0.34 1.15 0.49 1.15 0.47 1.15 0.39 1.15 0.32 1.15 0.29 1.15 0.35 1.15

1970 0.13 1.15 2.83 1.15 0.26 1.15 0.37 1.15 0.36 1.15 0.31 1.15 0.25 1.15 0.23 1.15 0.29 1.15

1971 0.16 1.15 3.27 1.15 0.31 1.15 0.45 1.15 0.44 1.15 0.38 1.15 0.32 1.15 0.29 1.15 0.39 1.15

1972 0.19 1.15 4.11 1.15 0.40 1.15 0.55 1.15 0.53 1.15 0.46 1.15 0.38 1.15 0.34 1.15 0.44 1.15

1973 0.21 1.15 3.60 1.15 0.43 1.15 0.55 1.15 0.54 1.15 0.48 1.15 0.40 1.15 0.37 1.15 0.50 1.15

1974 0.12 1.15 2.05 1.15 0.23 1.15 0.31 1.15 0.31 1.15 0.28 1.15 0.23 1.15 0.22 1.15 0.32 1.15

1975 0.08 1.15 1.81 1.15 0.15 1.15 0.23 1.15 0.22 1.15 0.18 1.15 0.15 1.15 0.14 1.15 0.17 1.15

1976 0.10 1.15 2.43 1.15 0.20 1.15 0.30 1.15 0.29 1.15 0.24 1.15 0.19 1.15 0.18 1.15 0.22 1.15

1977 0.08 1.15 2.28 1.15 0.16 1.15 0.25 1.15 0.24 1.15 0.20 1.15 0.16 1.15 0.14 1.15 0.16 1.15

1978 0.06 1.15 1.33 1.15 0.13 1.15 0.18 1.15 0.17 1.15 0.15 1.15 0.12 1.15 0.11 1.15 0.14 1.15

1979 0.06 1.15 1.58 1.15 0.13 1.15 0.19 1.15 0.18 1.15 0.15 1.15 0.13 1.15 0.11 1.15 0.14 1.15

1980 0.09 1.15 2.35 1.15 0.16 1.15 0.27 1.15 0.25 1.15 0.21 1.15 0.17 1.15 0.15 1.15 0.18 1.15

1981 0.07 1.15 1.72 1.15 0.15 1.15 0.22 1.15 0.21 1.15 0.18 1.15 0.14 1.15 0.13 1.15 0.16 1.15

1982 0.07 1.15 1.50 1.15 0.14 1.15 0.20 1.15 0.19 1.15 0.17 1.15 0.14 1.15 0.12 1.15 0.17 1.15

1983 0.04 1.15 0.50 1.15 0.08 1.15 0.10 1.15 0.10 1.15 0.10 1.15 0.08 1.15 0.08 1.15 0.13 1.15

1984 0.04 1.15 0.54 1.15 0.07 1.15 0.09 1.15 0.10 1.15 0.09 1.15 0.08 1.15 0.07 1.15 0.12 1.15

1985 0.05 1.15 0.84 1.15 0.10 1.15 0.14 1.15 0.14 1.15 0.13 1.15 0.11 1.15 0.10 1.15 0.17 1.15

1986 0.06 1.15 0.19 1.15 0.12 1.15 0.13 1.15 0.14 1.15 0.15 1.15 0.13 1.15 0.13 1.15 0.24 1.15

1987 0.09 0.86 0.33 0.97 0.21 0.98 0.19 0.89 0.20 0.87 0.21 0.83 0.17 0.82 0.17 0.81 0.26 0.66

1988 0.03 0.79 0.11 0.91 0.07 0.91 0.07 0.82 0.08 0.80 0.08 0.76 0.07 0.75 0.06 0.74 0.11 0.60

1989 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 185 of 248

Table C-20 (Continued).

Year

RR 

Yard GSD

P

Area GSD

L

Area GSD

KArea GSD FS GSD CS GSD Avg. GSD Max. GSD

1955 0.67 1.15 4.21 1.15 2.67 1.15 2.56 1.15 0.18 1.15 0.56 1.15 1.15 1.15 4.21 1.15

1956 0.78 1.15 4.09 1.15 4.27 1.15 4.23 1.15 0.26 1.15 1.02 1.15 1.67 1.15 4.27 1.15

1957 0.96 1.15 4.95 1.15 5.13 1.15 5.04 1.15 0.31 1.15 1.14 1.15 2.02 1.15 5.13 1.15

1958 1.13 1.15 5.78 1.15 5.99 1.15 5.90 1.15 0.37 1.15 1.35 1.15 2.39 1.15 5.99 1.15

1959 1.44 1.15 7.41 1.15 7.69 1.15 7.58 1.15 0.47 1.15 1.75 1.15 3.04 1.15 7.69 1.15

1960 1.38 1.15 6.72 1.15 6.93 1.15 6.84 1.15 0.47 1.15 1.72 1.15 2.97 1.15 7.12 1.15

1961 1.43 1.15 7.38 1.15 6.96 1.15 6.84 1.15 0.47 1.15 1.72 1.15 2.97 1.15 7.38 1.15

1962 1.48 1.15 7.44 1.15 7.01 1.15 6.91 1.15 0.49 1.15 1.85 1.15 3.12 1.15 7.44 1.15

1963 1.52 1.15 7.52 1.15 7.77 1.15 7.67 1.15 0.51 1.15 1.91 1.15 3.26 1.15 7.77 1.15

1964 1.26 1.15 5.95 1.15 6.12 1.15 6.05 1.15 0.44 1.15 1.64 1.15 2.76 1.15 7.04 1.15

1965 1.14 1.15 5.19 1.15 6.02 1.15 5.96 1.15 0.40 1.15 1.49 1.15 2.54 1.15 6.28 1.15

1966 0.94 1.15 5.56 1.15 5.88 1.15 5.84 1.15 0.29 1.15 1.34 1.15 2.05 1.15 5.88 1.15

1967 1.08 1.15 6.36 1.15 6.72 1.15 6.68 1.15 0.34 1.15 1.54 1.15 2.34 1.15 6.72 1.15

1968 0.77 1.15 5.23 1.15 1.01 1.15 5.49 1.15 0.24 1.15 1.28 1.15 1.64 1.15 6.11 1.15

1969 0.46 1.15 3.02 1.15 0.65 1.15 3.86 1.15 0.15 1.15 0.81 1.15 1.04 1.15 3.86 1.15

1970 0.40 1.15 2.73 1.15 0.51 1.15 2.85 1.15 0.12 1.15 0.61 1.15 0.82 1.15 2.85 1.15

1971 0.58 1.15 4.31 1.15 0.66 1.15 3.16 1.15 0.15 1.15 0.72 1.15 1.04 1.15 4.31 1.15

1972 0.60 1.15 4.18 1.15 0.79 1.15 4.45 1.15 0.18 1.15 0.90 1.15 1.23 1.15 4.45 1.15

1973 0.72 1.15 5.13 1.15 0.90 1.15 5.34 1.15 0.19 1.15 0.86 1.15 1.35 1.15 5.34 1.15

1974 0.49 1.15 3.77 1.15 0.52 1.15 2.36 1.15 0.11 1.15 0.49 1.15 0.79 1.15 3.77 1.15

1975 0.24 1.15 1.70 1.15 0.30 1.15 1.44 1.15 0.07 1.15 0.38 1.15 0.48 1.15 1.81 1.15

1976 0.29 1.15 1.99 1.15 0.37 1.15 1.90 1.15 0.09 1.15 0.50 1.15 0.62 1.15 2.43 1.15

1977 0.21 1.15 1.39 1.15 0.28 1.15 1.31 1.15 0.07 1.15 0.44 1.15 0.49 1.15 2.28 1.15

1978 0.19 1.15 1.30 1.15 0.25 1.15 1.42 1.15 0.06 1.15 0.29 1.15 0.39 1.15 1.42 1.15

1979 0.18 1.15 1.24 1.15 0.24 1.15 1.24 1.15 0.06 1.15 0.32 1.15 0.40 1.15 1.58 1.15

1980 0.24 1.15 1.67 1.15 0.30 1.15 1.31 1.15 0.08 1.15 0.46 1.15 0.53 1.15 2.35 1.15

1981 0.22 1.15 1.47 1.15 0.28 1.15 1.53 1.15 0.07 1.15 0.36 1.15 0.46 1.15 1.72 1.15

1982 0.24 1.15 1.76 1.15 0.28 1.15 1.33 1.15 0.07 1.15 0.32 1.15 0.45 1.15 1.76 1.15

1983 0.21 1.15 1.76 1.15 0.20 1.15 0.83 1.15 0.04 1.15 0.14 1.15 0.29 1.15 1.76 1.15

1984 0.20 1.15 1.67 1.15 0.17 1.15 0.50 1.15 0.04 1.15 0.14 1.15 0.26 1.15 1.67 1.15

1985 0.27 1.15 2.23 1.15 0.25 1.15 0.84 1.15 0.05 1.15 0.21 1.15 0.38 1.15 2.23 1.15

1986 0.44 1.15 3.76 1.15 0.36 1.15 1.27 1.15 0.07 1.15 0.15 1.15 0.49 1.15 3.76 1.15

1987 0.40 0.50 2.89 0.20 1.40 1.04 3.03 1.12 0.09 0.82 0.23 0.90 0.66 0.83 3.03 1.12

1988 0.18 0.45 1.31 0.18 0.71 1.05 0.80 1.11 0.03 0.75 0.09 0.83 0.25 0.75 1.31 0.82

1989 0.00 1.15 0.01 1.15 0.11 1.15 0.01 1.15 0.00 1.15 0.00 1.15 0.01 1.15 0.11 1.15

a. Doses estimated based on 2,000 hours per year exposure and FGR #12 dose conversion factor of 1.01E-13 Sv m3 per 

sec Bq.

b. Dose to skin should be assessed as 30% from >15 KeV electrons and 70% from >250 KeV photons.

c. Dose to bone surfaces should be assumed to be 100% from >250 KeV photons. 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 186 of 248

Table C-21. Argon-41 50th-percentile dosea,b to whole body and other organs resulting from reactor 

releases (mrem per year) and geometric standard deviation.

Year

A

Area GSD

C

Area GSD

D

Area GSD

F

Area GSD BG GSD

H

Area GSD

S

Area GSD

Z

Area GSD

R 

Area GSD

1955 0.12 1.15 1.28 1.15 0.21 1.15 0.28 1.15 0.30 1.15 0.30 1.15 0.26 1.15 0.25 1.15 1.52 1.15

1956 0.18 1.15 2.89 1.15 0.34 1.15 0.47 1.15 0.48 1.15 0.45 1.15 0.38 1.15 0.36 1.15 1.85 1.15

1957 0.22 1.15 2.99 1.15 0.40 1.15 0.54 1.15 0.56 1.15 0.55 1.15 0.47 1.15 0.45 1.15 2.90 1.15

1958 0.26 1.15 3.58 1.15 0.47 1.15 0.64 1.15 0.67 1.15 0.65 1.15 0.55 1.15 0.53 1.15 3.47 1.15

1959 0.33 1.15 4.73 1.15 0.61 1.15 0.83 1.15 0.86 1.15 0.83 1.15 0.70 1.15 0.67 1.15 4.11 1.15

1960 0.32 1.15 4.68 1.15 0.58 1.15 0.82 1.15 0.85 1.15 0.84 1.15 0.71 1.15 0.68 1.15 5.05 1.15

1961 0.32 1.15 4.68 1.15 0.58 1.15 0.81 1.15 0.84 1.15 0.83 1.15 0.70 1.15 0.67 1.15 4.58 1.15

1962 0.34 1.15 5.20 1.15 0.60 1.15 0.87 1.15 0.90 1.15 0.88 1.15 0.74 1.15 0.71 1.15 5.09 1.15

1963 0.35 1.15 5.25 1.15 0.64 1.15 0.90 1.15 0.93 1.15 0.91 1.15 0.77 1.15 0.74 1.15 5.12 1.15

1964 0.30 1.15 4.61 1.15 0.53 1.15 0.77 1.15 0.80 1.15 0.79 1.15 0.67 1.15 0.64 1.15 4.99 1.15

1965 0.28 1.15 4.09 1.15 0.50 1.15 0.71 1.15 0.73 1.15 0.72 1.15 0.61 1.15 0.58 1.15 4.45 1.15

1966 0.22 1.15 3.99 1.15 0.45 1.15 0.60 1.15 0.59 1.15 0.52 1.15 0.43 1.15 0.40 1.15 0.51 1.15

1967 0.25 1.15 4.56 1.15 0.51 1.15 0.68 1.15 0.68 1.15 0.60 1.15 0.49 1.15 0.45 1.15 0.58 1.15

1968 0.19 1.15 4.33 1.15 0.37 1.15 0.54 1.15 0.52 1.15 0.44 1.15 0.36 1.15 0.33 1.15 0.40 1.15

1969 0.12 1.15 2.72 1.15 0.24 1.15 0.35 1.15 0.33 1.15 0.28 1.15 0.23 1.15 0.21 1.15 0.25 1.15

1970 0.09 1.15 2.00 1.15 0.19 1.15 0.26 1.15 0.25 1.15 0.22 1.15 0.18 1.15 0.16 1.15 0.20 1.15

1971 0.11 1.15 2.32 1.15 0.22 1.15 0.32 1.15 0.31 1.15 0.27 1.15 0.22 1.15 0.21 1.15 0.28 1.15

1972 0.14 1.15 2.91 1.15 0.28 1.15 0.39 1.15 0.38 1.15 0.32 1.15 0.27 1.15 0.24 1.15 0.31 1.15

1973 0.15 1.15 2.55 1.15 0.31 1.15 0.39 1.15 0.39 1.15 0.34 1.15 0.28 1.15 0.26 1.15 0.35 1.15

1974 0.08 1.15 1.45 1.15 0.16 1.15 0.22 1.15 0.22 1.15 0.20 1.15 0.17 1.15 0.15 1.15 0.22 1.15

1975 0.05 1.15 1.29 1.15 0.11 1.15 0.16 1.15 0.15 1.15 0.13 1.15 0.11 1.15 0.10 1.15 0.12 1.15

1976 0.07 1.15 1.73 1.15 0.14 1.15 0.21 1.15 0.20 1.15 0.17 1.15 0.14 1.15 0.12 1.15 0.15 1.15

1977 0.06 1.15 1.62 1.15 0.11 1.15 0.18 1.15 0.17 1.15 0.14 1.15 0.11 1.15 0.10 1.15 0.12 1.15

1978 0.04 1.15 0.95 1.15 0.09 1.15 0.13 1.15 0.12 1.15 0.10 1.15 0.09 1.15 0.08 1.15 0.10 1.15

1979 0.05 1.15 1.12 1.15 0.09 1.15 0.14 1.15 0.13 1.15 0.11 1.15 0.09 1.15 0.08 1.15 0.10 1.15

1980 0.06 1.15 1.66 1.15 0.12 1.15 0.19 1.15 0.18 1.15 0.15 1.15 0.12 1.15 0.11 1.15 0.13 1.15

1981 0.05 1.15 1.22 1.15 0.10 1.15 0.15 1.15 0.15 1.15 0.12 1.15 0.10 1.15 0.09 1.15 0.11 1.15

1982 0.05 1.15 1.06 1.15 0.10 1.15 0.14 1.15 0.14 1.15 0.12 1.15 0.10 1.15 0.09 1.15 0.12 1.15

1983 0.03 1.15 0.35 1.15 0.06 1.15 0.07 1.15 0.07 1.15 0.07 1.15 0.06 1.15 0.06 1.15 0.09 1.15

1984 0.03 1.15 0.39 1.15 0.05 1.15 0.07 1.15 0.07 1.15 0.06 1.15 0.05 1.15 0.05 1.15 0.08 1.15

1985 0.04 1.15 0.60 1.15 0.07 1.15 0.10 1.15 0.10 1.15 0.09 1.15 0.08 1.15 0.07 1.15 0.12 1.15

1986 0.04 1.15 0.14 1.15 0.08 1.15 0.09 1.15 0.10 1.15 0.11 1.15 0.09 1.15 0.09 1.15 0.17 1.15

1987 0.06 0.86 0.24 0.97 0.15 0.98 0.13 0.89 0.14 0.87 0.15 0.83 0.12 0.82 0.12 0.81 0.19 0.66

1988 0.02 0.79 0.08 0.91 0.05 0.91 0.05 0.82 0.05 0.80 0.06 0.76 0.05 0.75 0.05 0.74 0.08 0.60

1989 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 187 of 248

Table C-21 (Continued).

Year

RR

Yard GSD

P 

Area GSD

L 

Area GSD

KArea GSD FS GSD CS GSD Avg. GSD Max. GSD

1955 0.47 1.15 2.99 1.15 1.90 1.15 1.81 1.15 0.12 1.15 0.40 1.15 0.81 1.15 2.99 1.15

1956 0.55 1.15 2.90 1.15 3.03 1.15 3.00 1.15 0.18 1.15 0.72 1.15 1.19 1.15 3.03 1.15

1957 0.68 1.15 3.51 1.15 3.63 1.15 3.58 1.15 0.22 1.15 0.81 1.15 1.43 1.15 3.63 1.15

1958 0.80 1.15 4.10 1.15 4.25 1.15 4.18 1.15 0.26 1.15 0.96 1.15 1.69 1.15 4.25 1.15

1959 1.02 1.15 5.26 1.15 5.45 1.15 5.38 1.15 0.33 1.15 1.24 1.15 2.16 1.15 5.45 1.15

1960 0.98 1.15 4.76 1.15 4.91 1.15 4.85 1.15 0.33 1.15 1.22 1.15 2.10 1.15 5.05 1.15

1961 1.01 1.15 5.23 1.15 4.93 1.15 4.85 1.15 0.33 1.15 1.22 1.15 2.11 1.15 5.23 1.15

1962 1.05 1.15 5.28 1.15 4.97 1.15 4.90 1.15 0.35 1.15 1.31 1.15 2.21 1.15 5.28 1.15

1963 1.08 1.15 5.33 1.15 5.51 1.15 5.44 1.15 0.36 1.15 1.35 1.15 2.31 1.15 5.51 1.15

1964 0.90 1.15 4.22 1.15 4.34 1.15 4.29 1.15 0.31 1.15 1.16 1.15 1.96 1.15 4.99 1.15

1965 0.81 1.15 3.68 1.15 4.27 1.15 4.23 1.15 0.29 1.15 1.05 1.15 1.80 1.15 4.45 1.15

1966 0.67 1.15 3.94 1.15 4.17 1.15 4.14 1.15 0.21 1.15 0.95 1.15 1.45 1.15 4.17 1.15

1967 0.76 1.15 4.51 1.15 4.76 1.15 4.73 1.15 0.24 1.15 1.09 1.15 1.66 1.15 4.76 1.15

1968 0.54 1.15 3.71 1.15 0.72 1.15 3.89 1.15 0.17 1.15 0.91 1.15 1.16 1.15 4.33 1.15

1969 0.33 1.15 2.14 1.15 0.46 1.15 2.73 1.15 0.11 1.15 0.57 1.15 0.74 1.15 2.73 1.15

1970 0.28 1.15 1.94 1.15 0.36 1.15 2.02 1.15 0.08 1.15 0.43 1.15 0.58 1.15 2.02 1.15

1971 0.41 1.15 3.06 1.15 0.47 1.15 2.24 1.15 0.11 1.15 0.51 1.15 0.74 1.15 3.06 1.15

1972 0.43 1.15 2.96 1.15 0.56 1.15 3.16 1.15 0.13 1.15 0.64 1.15 0.87 1.15 3.16 1.15

1973 0.51 1.15 3.64 1.15 0.64 1.15 3.79 1.15 0.14 1.15 0.61 1.15 0.96 1.15 3.79 1.15

1974 0.35 1.15 2.67 1.15 0.37 1.15 1.67 1.15 0.08 1.15 0.35 1.15 0.56 1.15 2.67 1.15

1975 0.17 1.15 1.20 1.15 0.21 1.15 1.02 1.15 0.05 1.15 0.27 1.15 0.34 1.15 1.29 1.15

1976 0.21 1.15 1.41 1.15 0.27 1.15 1.35 1.15 0.06 1.15 0.35 1.15 0.44 1.15 1.73 1.15

1977 0.15 1.15 0.98 1.15 0.20 1.15 0.93 1.15 0.05 1.15 0.31 1.15 0.35 1.15 1.62 1.15

1978 0.13 1.15 0.92 1.15 0.18 1.15 1.00 1.15 0.04 1.15 0.20 1.15 0.28 1.15 1.00 1.15

1979 0.13 1.15 0.88 1.15 0.17 1.15 0.88 1.15 0.04 1.15 0.23 1.15 0.28 1.15 1.12 1.15

1980 0.17 1.15 1.19 1.15 0.21 1.15 0.93 1.15 0.06 1.15 0.33 1.15 0.37 1.15 1.66 1.15

1981 0.15 1.15 1.04 1.15 0.20 1.15 1.09 1.15 0.05 1.15 0.26 1.15 0.33 1.15 1.22 1.15

1982 0.17 1.15 1.25 1.15 0.20 1.15 0.94 1.15 0.05 1.15 0.23 1.15 0.32 1.15 1.25 1.15

1983 0.15 1.15 1.24 1.15 0.14 1.15 0.59 1.15 0.03 1.15 0.10 1.15 0.21 1.15 1.24 1.15

1984 0.14 1.15 1.18 1.15 0.12 1.15 0.35 1.15 0.03 1.15 0.10 1.15 0.18 1.15 1.18 1.15

1985 0.19 1.15 1.58 1.15 0.18 1.15 0.60 1.15 0.04 1.15 0.15 1.15 0.27 1.15 1.58 1.15

1986 0.31 1.15 2.66 1.15 0.26 1.15 0.90 1.15 0.05 1.15 0.11 1.15 0.35 1.15 2.66 1.15

1987 0.29 0.50 2.05 0.20 0.99 1.04 2.15 1.12 0.06 0.82 0.16 0.90 0.47 0.83 2.15 1.12

1988 0.12 0.45 0.93 0.18 0.50 1.05 0.57 1.11 0.02 0.75 0.06 0.83 0.18 0.75 0.93 0.82

1989 0.00 1.15 0.01 1.15 0.08 1.15 0.01 1.15 0.00 1.15 0.00 1.15 0.01 1.15 0.08 1.15

a. Doses estimated based on 2,000 hours per year exposure and FGR #12 dose conversion factor of 7.16E-14 Sv m3 per sec Bq.

b. Dose to whole body and other underlying organs should be assumed to be 100% from >250 KeV photons.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 188 of 248

C.3 Maximum Plausible Annual Doses From Ingestion of Foodstuffs at the Savannah 

River Site

C.3.1

The purpose of this calculation is to estimate the maximum plausible organ dose that a energy 

employee could have received by consuming radioactively contaminated foodstuffs (e.g., nuts, 

berries, fruits) gathered from the Savannah River Site (SRS). These potential doses have been 

bounded by the use of claimant favorable assumptions related to soil radionuclide concentrations and 

soil-to-vegetation transfer factors (TF) and, in the case of tritium, maximum vegetation concentrations.

Introduction

Radionuclides of Concern

The major radionuclides of concern (i.e., those which would provide 95% of the dose potential) 

through the soil pathway were identified in Section 1.3 and Attachment A of this Site Profile. These 

radionuclides include strontium-90, cesium-137, and plutonium-238 and 239.

Uptake from Soil and Ingestion Intakes

Soil analyses conducted in the 1970s and 1980s indicated no measurable SRS-released plutonium at 

off-site locations. However, elevated levels of certain radionuclides, including plutonium, were 

detected at several locations near the release points in both F- and H-Areas. Careful review of these 

data over several years concluded that the measured soil concentrations in F and H Areas were due 

almost entirely to deposition during the years of the greatest releases. Since the great majority of 

239Pu was released from F Area in a single year, 1955, and the majority of 238Pu was released from H 

Area in 1969, the measured soil concentrations for 239Pu could be conservatively assumed to have 

persisted in F Area since 1955 and for 238Pu in H Area since 1969 (RAC Report (CDC 2001, p. 12.2-

7)). Because the atmospheric releases of nonvolatile radionuclides from other on-site locations were 

much less than those from F and H Areas, only the soil measurements in F and H Area were deemed 

important for purposes of bounding the potential dose from the ingestion pathway. The maximum 

onsite concentrations of the radionuclides of concern are provided in Table C-22.

Table C-22. Radionuclide soil concentrations, pCi/g, (1 sigma) dry weight (0-8 cm depth).

Location Sr-90 σ Cs-137 σ Pu-238 σ Pu-239 σ

F-Area 0.0216 0.0182 1.05 0.05 0.403 0.014 0.531 0.014

H-Area 0.0289 0.0183 1.32 0.05 0.0213 0.0016 0.0554 0.0023

Source: CDC 2001, Table 12.2-8, on p. 12.2-8, Arnett 1993.

If the assumption is made that all of the foodstuffs were grown in the areas of highest radionuclide 

concentrations, the content of the radionuclide in the plant mass can be calculated by use of Transfer 

Factors. These factors provide the ratio of the plant mass to the soil concentration; dry mass to dry 

mass. Transfer factors for various plant types have been recently tabulated from multiple sources in 

PNNL-13421 (2003) and provided in Tables C-23, C-24, and C-25 below. 

An inspection of the values in Table C-23, C-24, and C-25 would provide the conservative, upper 

bound TF values for all food stuffs shown in Table C-26. If the assumption is made that the foodstuffs 

are 80% water by weight, then the radionuclide intake per kilogram consumed (wet weight) can be 

estimated with the following expression:

Intake (pCi/year) = Soil Concentration (pCi/g) × TF × Consumption Rate (kg/year) ×

0.20 (wet to dry ratio)
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 189 of 248

Table C-23. Transfer factors for leafy vegetables (Bq/kg leafy vegetables per Bq/kg soil).

Element TF

Be 1.00E-02

C 7.00E-01

N 5.50E-02

F 6.00E-02

Na 3.00E-01

Mg 1.00E+00

Si 3.50E-01

P 3.50E+00

S 1.50E+00

Cl 7.00E+01

K 1.00E+00

Ca 3.50E+00

Sc 6.00E-03

Cr 7.50E-03

Mn 7.00E-01

Fe 5.00E-02

Co 2.30E-01

Ni 2.80E-01

Cu 4.00E-01

Zn 1.30E+00

Element TF

Ga 4.00E-03

As 4.00E-02

Se 2.50E-01

Br 1.50E+00

Rb 9.00E-01

Sr 3.00E+00

Y 1.00E-02

Zr 1.00E-03

Nb 2.50E-02

Mo 8.00E-01

Tc 2.10E+02

Ru 4.00E-02

Rh 1.50E-01

Pd 1.50E-01

Ag 2.70E-04

Cd 5.50E-01

In 4.00E-03

Sn 3.00E-02

Sb 1.30E-04

Te 2.50E-02

Element TF

I 4.00E-02

Cs 4.60E-01

Ba 1.50E-01

La 5.20E-03

Ce 2.00E-02

Pr 2.00E-02

Nd 2.00E-02

Pm 2.00E-02

Sm 2.00E-02

Eu 2.00E-02

Gd 2.00E-02

Tb 2.00E-02

Dy 2.00E-02

Ho 2.00E-02

Er 2.00E-02

Hf 1.00E-03

Ta 2.50E-02

W 3.00E+00

Re 1.50E+00

Os 1.50E-02

Element TF

Ir 5.50E-02

Au 1.00E-02

Hg 8.50E-01

Tl 4.00E-03

Pb 1.00E-02

Bi 5.00E-01

Po 1.20E-03

Ra 4.90E-02

Ac 4.70E-04

Th 1.80E-03

Pa 4.70E-04

U 8.30E-03

Np 3.20E-02

Pu 6.00E-05

Am 4.70E-04

Cm 7.70E-04

Cf 4.70E-04

Cf 2.50E-04

Table C-24. Transfer factors for fruit (Bq/kg fruit per Bq/kg soil).

Element TF

Be 1.50E-03

C 7.00E-01

N 3.00E-02

F 6.00E-03

Na 3.00E-01

Mg 5.50E-01

Si 7.00E-02

P 3.50E+00

S 1.50E+00

Cl 7.00E+01

K 5.50E-01

Ca 3.50E-01

Sc 1.00E-03

Cr 4.50E-03

Mn 5.00E-02

Fe 5.00E-02

Co 7.00E-03

Ni 6.00E-02

Cu 2.50E-01

Zn 9.00E-01

Element TF

Ga 4.00E-04

As 6.00E-03

Se 5.00E-02

Br 1.50E+00

Rb 9.00E-01

Sr 2.00E-01

Y 1.00E-02

Zr 1.00E-03

Nb 2.50E-02

Mo 5.00E-02

Tc 1.50E+00

Ru 4.00E-02

Rh 4.00E-02

Pd 4.00E-02

Ag 8.00E-04

Cd 1.50E-01

In 4.00E-04

Sn 6.00E-03

Sb 8.00E-05

Te 4.00E-03

Element TF

I 4.00E-02

Cs 2.20E-01

Ba 1.50E-02

La 4.00E-03

Ce 2.00E-02

Pr 2.00E-02

Nd 2.00E-02

Pm 2.00E-02

Sm 2.00E-02

Eu 2.00E-02

Gd 2.00E-02

Tb 2.00E-02

Dy 2.00E-02

Ho 2.00E-02

Er 2.00E-02

Hf 1.00E-03

Ta 2.50E-02

W 3.00E+00

Re 3.50E-01

Os 4.50E-02

Element TF

Ir 1.50E-02

Au 1.40E-02

Hg 3.70E-01

Tl 4.00E-04

Pb 1.00E-02

Bi 5.00E-01

Po 1.20E-03

Ra 6.10E-03

Ac 2.50E-04

Th 2.50E-04

Pa 2.50E-04

U 4.00E-03

Np 1.00E-02

Pu 4.50E-05

Am 2.50E-04

Cm 1.50E-05

Cf 2.50E-04

The Georgia Department of Natural Resources (GDNR 2004) estimates that, in Georgia, the annual 

consumption of wild foodstuffs is 60 kilograms. Therefore, using the expression above, the maximum 

plausible intake (per kilogram and per year) of foodstuff consumed can be calculated. The 

radionuclide specific intakes are provided in Table C-26.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 190 of 248

Table C-25. Transfer factors for root vegetables (Bq/kg root vegetable per Bq/kg soil).

Element TF

C 7.00E-01

N 4.90E-02

F 6.00E-03

Na 3.00E-01

Mg 5.50E-01

Si 7.00E-02

P 3.50E+00

S 1.50E+00

Cl 7.00E+01

K 5.50E-01

Ca 3.50E-01

Sc 1.00E-03

Cr 4.50E-03

Mn 2.00E-01

Fe 5.00E-02

Co 6.70E-02

Ni 6.00E-02

Cu 2.50E-01

Zn 3.50E-01

Element TF

Ga 4.00E-04

As 6.00E-03

Se 5.00E-02

Br 1.50E+00

Rb 9.00E-01

Sr 5.00E-01

Y 1.00E-02

Zr 1.00E-03

Nb 2.50E-02

Mo 8.00E-01

Tc 2.40E-01

Ru 4.00E-02

Rh 4.00E-02

Pd 4.00E-02

Ag 1.30E-03

Cd 1.50E-01

In 4.00E-04

Sn 6.00E-03

Sb 5.60E-04

Element TF

Te 4.00E-03

I 4.00E-02

Cs 1.30E-01

Ba 1.50E-02

La 3.50E-04

Ce 2.00E-02

Pr 2.00E-02

Nd 2.00E-02

Pm 2.00E-02

Sm 2.00E-02

Eu 2.00E-02

Gd 2.00E-02

Tb 2.00E-02

Dy 2.00E-02

Ho 2.00E-02

Er 2.00E-02

Hf 1.00E-03

Ta 2.50E-02

W 3.00E+00

Element TF

Re 3.50E-01

Os 3.50E-03

Ir 1.50E-02

Au 1.80E-02

Hg 2.00E-01

Tl 4.00E-04

Pb 6.00E-03

Bi 5.00E-01

Po 7.00E-03

Ra 2.00E-03

Ac 3.50E-04

Th 3.30E-04

Pa 3.50E-04

U 1.20E-02

Np 1.30E-02

Pu 1.10E-03

Am 3.50E-04

Cm 4.30E-04

Cf 3.50E-04

Table C-26. Parameters used for calculation of 50-year equivalent dose.

Radionuclide TF

Maximum soil 

concentration, pCi/g

Maximum 

intake, pCi/kg

Total annual intake,

pCi, for 60 kg

Absorption 

factor, f1

Sr-90 3 0.0289 17.34 1,040.0 0.3

Cs-137 0.4 1.32 105.6 6,336.0 1.0

Pu-238 0.0011 0.403 0.08866 5.3 0.0005

Pu-239 0.0011 0.531 0.11682 7.0 0.0005

It is recognized that the maximum strontium-90 concentration (i.e., intake) shown in Table C-26 is 

lower than the concentrations reported for some onsite vegetation samples reported in the SRS 

annual environmental reports. However, most of the strontium found in vegetation (both onsite and 

offsite) is from global fallout resulting from worldwide weapons testing and is therefore not eligible for 

consideration under the Energy Employee Occupational Illness Compensation Program Act 

(EEOICPA). This fact was clearly shown in a report by Rollins (2003) which supported development 

of the Occupational Environmental Dose section of this Site Profile). The report documented that fact 

that the onsite, nonvolatile, atmospheric beta concentrations correlated very closely the offsite control 

station concentrations and very poorly with SRS nonvolatile beta source terms. However, it is 

recognized that there are areas onsite at the SRS where gross contamination of soil and vegetation is 

known to exist (e.g., burial ground and seepage basins) but these areas have been fenced off and 

thus made inaccessible to workers. For this reason, the higher concentrations reported for selected 

vegetations samples were not deemed appropriate for use for dose reconstruction under the 

EEOICPA.

Tritium Intake

In terms of curies released and physical properties, tritium, or more specifically, tritium oxide (HTO) is 

by far the most available, most mobile, and most easily taken up by vegetation. It readily takes on the 

chemical form of water. In the environmental sampling conducted by SRS, and the states of Georgia 

and South Carolina, HTO is by far the most widely found contaminate attributed to SRS operations.
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The interest in HTO has increased over the years, due to the increasing releases and improved 

monitoring techniques. In the middle 70’s, HTO was reported in units of pCi/ml free water obtained by 

freeze drying the vegetation sample. By the 80’s, HTO was reported in pCi/g wet weight of the 

sample. The factor connecting the two is the moisture content of the sample. Most edible vegetation 

has moisture content between 60 and 80% of the wet weight at time of harvest. In converting from 

pCi/ml to pCi/gm, moisture content of 60% was used, providing higher HTO concentrations.

The range of maximum reported HTO concentration each year is 2.8 to 380 pCi/ml with an average 

maximum of 130 pCi/ml. The assumption of 400 pCi/ml includes the worst case reported in the 

annual reports (see reference section for details). Table C-27 provides the maximum values found in 

the SRS environmental reports; note there are years that did not report HTO.

Table C-27. Maximum HTO concentration measured in onsite 

vegetation at the SRS.

Year pCi/ml

2000 4

1999 2

1997 18

1996 106

1995 31

1994 52

Year pCi/ml

1993 38

1992 44

1991 90

1990 240

1989 81

1988 160

Year pCi/ml

1987 380

1986 180

1983 330

1982 150

1979 170

1978 179

The maximum plausible HTO intake can be calculated using the claimant favorable vegetation 

concentration of 400 pCi/ml, 80% water content of the vegetation, and a consumption of 60 kg per 

year. Using these assumptions, the maximum plausible HTO intake is 1.92 x 107 pCi per year.

Dose

Using the IMBA computer code, the 50-year organ equivalent doses can be calculated for these 

intakes. The f1 absorption factors used for this evaluation were 0.0005, 0.3, and 1.0 for plutonium, 

strontium, and cesium, respectively. The values provided in Table C-28 are 50-year equivalent 

doses, mrem, per 60 kilograms (wet mass) of foodstuffs consumed.

Table C-28. 50-year equivalent organ doses, mrem, based on 60 kilogram per 

year consumption. 

Target organs Sr-90 Cs-137 Pu-239 Pu-238 HTO Total

Adrenals 2.55E-03 3.30E-01 3.76E-09 2.54E-09 1.30E+00 1.63E+00

Urinary bladder 5.68E-03 3.40E-01 3.76E-09 2.54E-09 1.30E+00 1.65E+00

Brain 2.55E-03 2.77E-01 3.76E-09 2.54E-09 1.30E+00 1.58E+00

Breast 2.55E-03 2.64E-01 3.76E-09 2.54E-09 1.30E+00 1.57E+00

Gall bladder 2.55E-03 3.26E-01 3.76E-09 2.54E-09 1.30E+00 1.63E+00

Heart wall 2.55E-03 3.25E-01 3.76E-09 2.54E-09 1.30E+00 1.63E+00

Kidneys 2.55E-03 3.17E-01 8.78E-09 6.30E-09 1.30E+00 1.62E+00

Liver 2.55E-03 3.19E-01 4.48E-07 3.07E-07 1.30E+00 1.62E+00

Muscle 2.55E-03 2.95E-01 3.76E-09 2.54E-09 1.30E+00 1.60E+00

Ovaries 2.55E-03 3.37E-01 2.80E-08 1.91E-08 1.30E+00 1.64E+00

Pancreas 2.55E-03 3.38E-01 3.76E-09 2.54E-09 1.30E+00 1.64E+00

Testes 2.55E-03 2.96E-01 2.85E-08 1.95E-08 1.30E+00 1.60E+00

Thyroid 2.55E-03 3.07E-01 3.76E-09 2.54E-09 1.30E+00 1.61E+00

R.B.M. 6.87E-01 3.08E-01 1.01E-07 7.19E-08 1.30E+00 2.30E+00

Bone surface 1.57E+00 3.22E-01 2.13E-06 1.45E-06 1.30E+00 3.19E+00

Stomach 3.46E-03 3.15E-01 4.07E-09 2.79E-09 1.30E+00 1.62E+00

S.I. 4.35E-03 3.47E-01 4.54E-09 3.17E-09 1.30E+00 1.65E+00
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Table C-28 (Continued).

Target organs Sr-90 Cs-137 Pu-239 Pu-238 HTO Total

U.L.I. 2.24E-02 3.39E-01 8.45E-09 6.34E-09 1.30E+00 1.66E+00

L.L.I. 8.40E-02 3.90E-01 1.75E-08 1.37E-08 1.30E+00 1.77E+00

Skin 2.55E-03 2.52E-01 3.76E-09 2.54E-09 1.30E+00 1.55E+00

Spleen 2.55E-03 3.17E-01 3.76E-09 2.54E-09 1.30E+00 1.62E+00

Thymus 2.55E-03 3.08E-01 3.76E-09 2.54E-09 1.30E+00 1.61E+00

Uterus 2.55E-03 3.40E-01 3.76E-09 2.54E-09 1.30E+00 1.64E+00

ET 2.55E-03 3.07E-01 3.76E-09 2.54E-09 1.30E+00 1.61E+00

Lung 2.55E-03 2.98E-01 3.76E-09 2.54E-09 1.30E+00 1.60E+00

Colon 4.91E-02 3.61E-01 1.24E-08 9.52E-09 1.30E+00 1.71E+00

ET1 2.55E-03 3.07E-01 3.76E-09 2.54E-09 1.30E+00 1.61E+00

ET2 2.55E-03 3.07E-01 3.76E-09 2.54E-09 1.30E+00 1.61E+00

LN(ET) 2.55E-03 3.07E-01 3.76E-09 2.54E-09 1.30E+00 1.61E+00

BBsec 2.55E-03 2.98E-01 3.76E-09 2.54E-09 1.30E+00 1.60E+00

BBbas 2.55E-03 2.98E-01 3.76E-09 2.54E-09 1.30E+00 1.60E+00

bb 2.55E-03 2.98E-01 3.76E-09 2.54E-09 1.30E+00 1.60E+00

AI 2.55E-03 2.98E-01 3.76E-09 2.54E-09 1.30E+00 1.60E+00

LN(TH) 2.55E-03 2.98E-01 3.76E-09 2.54E-09 1.30E+00 1.60E+00

Esophagus 2.55E-03 3.08E-01 3.76E-09 2.54E-09 1.30E+00 1.61E+00

Gonads 2.55E-03 3.37E-01 2.85E-08 1.95E-08 1.30E+00 1.64E+00

Remainder 2.58E-03 2.96E-01 3.83E-09 2.59E-09 1.30E+00 1.60E+00

Doses in Table C-28 are the 50-year equivalent organ doses for a one year chronic ingestion. 

However, dose from a long term chronic ingestion must be considered. For Cs-137, Pu-238, and Pu239, annual doses from a long-term chronic ingestion of foodstuffs are insignificant and may be 

omitted from dose reconstructions. 

For tritium, the doses in Table C-28 are comparable to annual doses from a long term chronic intake, 

therefore, 0.013 rem should be assigned for each year that the worker had potential dose from 

ingestion of onsite foodstuffs, or use the entire employment period if the specific years are not known. 

The tritium dose should be assigned as a chronic exposure to electrons less than 15 keV with a 

constant distribution.

For Sr-90 in foodstuffs, a chronic intake over a period of years results in gradually increasing annual 

doses. Doses to the bone surfaces and red bone marrow need to be considered, however, foodstuff 

doses from Sr-90 to other organs and tissues are insignificant and may be omitted from dose 

reconstructions. For a continuous Sr-90 ingestion of 2.85 pCi/day, dose increases from less than 

0.001 rem in year 1 to about 0.003 rem per year in year 65. Therefore, 0.003 rem per year provides 

an upper bound to the annual bone surface dose and may be assigned to all years up to diagnosis. 

For red bone marrow, dose from a long term chronic intake does not exceed 0.001 rem in a year. 

Therefore, 0.001 rem per year provides an upper bound and may be assigned to all years up to 

diagnosis. The Sr-90 doses should be assigned as a chronic exposure to electrons greater than 15 

keV with a constant distribution.
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ATTACHMENT D

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

TABLE OF CONTENTS

Section

D.0 Occupational Internal Dose.....................................................................................................

Page

195

D.1 In Vitro Minimum Detectable Activities (MDAs) and Reporting Levels ..................................... 195

D.1.1 In Vitro Urine Analysis .................................................................................................... 195

D.1.2 In Vitro Urine Analyses for Individual Radionuclides....................................................... 196

D.1.3 In Vitro Fecal Analysis.................................................................................................... 197

D.2 In Vivo MDAs and Reporting Practices at SRS ....................................................................... 197

D.2.1 Whole Body Counting..................................................................................................... 197

D.2.2 Chest Counting .............................................................................................................. 198

D.3 Tables of Chronic Intakes of Various Fission Product and Activation Product 

Radionuclides Based on MDA Whole Body Count on Last Day of the Period ......................... 198
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LIST OF TABLES

Table

D-1 Limits of detection for urinalysis................................................................................

Page

196

D-2 MDAs for whole body counting ................................................................................. 198

D-3 MDAs for chest counting........................................................................................... 199

D-4 95Zr chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 199

D-5 95Nb chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 200

D-6 54Mn chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 200

D-7 131I chronic inhalation intake assessment based on MDA whole body count on 

last day of the period ................................................................................................ 201

D-8 154Eu chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 201

D-9 65Zn chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 202

D-10 140Ba chronic inhalation intake assessment based on MDA whole body count 

on last day of the period ........................................................................................... 202

D-11 Ba-140, Type F, f1=0.1, 50-year equivalent dose for years of chronic intake 

(Sv) MDA = 9.3 nCi.................................................................................................. 204

D-12 Ce-144, Type W, 50-year equivalent dose for years of chronic intake (Sv) 

(MDA = 29 nCi)......................................................................................................... 206

D-13 Co-60, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 1 nCi............................................................................................................. 208

D-14 Co-60, Type S, 50-year equivalent dose for years of chronic intake (Sv) MDA 

= 1 nCi...................................................................................................................... 210

D-15 Cs-137, Type F 50-year equivalent dose for years of chronic intake (Sv) 

(MDA = 1 nCi)........................................................................................................... 212

D-16 Eu-154, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 2.4 nCi.......................................................................................................... 214

D-17 I-131, Type F, 50-year equivalent dose for years of chronic intake (Sv) MDA 

= 1.4 nCi................................................................................................................... 216

D-18 La-140, Type F, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 9.3 nCi.......................................................................................................... 218

D-19 La-140, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 9.3 nCi.......................................................................................................... 220

D-20 Mn-54, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 1 nCi............................................................................................................. 222

D-21 Nb-95, Type M, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 2.2 nCi.......................................................................................................... 224

D-22 Nb-95, Type W, 50-year equivalent dose for years of chronic intake (Sv) 

MDA = 2.2 nCi.......................................................................................................... 226

D-23 Nb-95, Type S, 50-year equivalent dose for years of chronic intake (Sv) (MDA 

= 2.2 nCi).................................................................................................................. 228

D-24 Ru-106, Type F, f1=0.05, 50-year equivalent dose for years of chronic intake 

(Sv) MDA = 1 nCi .................................................................................................... 230

D-25 Ru-106, Type S, f1=0.05, 50-year equivalent dose for years of chronic intake 

(Sv) MDA = 1 nCi .................................................................................................... 232
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D-26 Zn-65, Type S, f1=0.49999, 50-year equivalent dose for years of chronic 

intake (Sv) MDA = 5.1 nCi ....................................................................................... 234

D-27 Zr-95, Type M, 50-year equivalent dose for years of chronic intake (Sv) MDA 

= 2.2 nCi................................................................................................................... 236

D-28 Zr-95, Type S, 50-year equivalent dose for years of chronic intake (Sv) MDA 

= 2.2 nCi................................................................................................................... 238

D.0 OCCUPATIONAL INTERNAL DOSE

The bioassay program monitored internal uptake of radionuclides using both in vitro analysis of urine 

and feces and in vivo monitoring using whole body and chest counting.

D.1 In Vitro Minimum Detectable Activities (MDAs) and Reporting Levels

The decision level is the level at which activity is considered present in a sample with a 95% 

confidence level. The reporting level is the minimally acceptable decision level. It appears from the 

database that at some times and certain radionuclides the reporting level was also the level below 

which a result was reported as “less than”, regardless of the actual decision level for a given sample. 

Since 1990, the MDA has been defined as the quantity of material (or activity) that has a 5% chance 

of not being detected. Prior to 1990, the term MDA may have had less rigorous or at least slightly 

different meanings. For some times and certain radionuclides, MDA and reporting level were 

synonymous. For the dosimetrist the reporting level is the relevant concept over the history of the 

SRS records. There is no consistent quantitative relationship between the reporting level and the 

MDA.

D.1.1 In Vitro Urine Analysis

The reporting levels and MDAs (if known and different than reporting levels), are shown in Table D-1. 

These values are applicable over the time period specified. The lack of an MDA in a given time period 

reflects the lack of documented information about urine analysis for that radionuclide. Activity fractions 

for depleted, recycled, and high enriched uranium are listed in the glossary section. Activity fractions 

for 6% and 12% plutonium mixtures are given in Chapter 4.

The following codes are found in the excreta records:

C routine urine sample, includes baselines and termination samples

A-A 24-hours special urine sample

A-O two 24-hour urine samples, one analyzed, one held 

A-W follow-up to positive routine sample

IA Inconclusive Analysis; analysis didn’t meet QC criteria; no result reported

B follow-up to an IA sample

A-F single void fecal sample

T tritium sample, only 50 mL collected

P-10 an old code for tritium

SP Neptunium

FP fission products; gamma scan.

FPIA fission products/induced activity. Same analysis as FP.

IA with a numerical result would be same as FP

LMF uranium
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Table D-1. Limits of detection for urinalysis.

a

Radionuclide Period MDA (pCi/L) Reporting level (pCi/L) 3

H Start to 1958 1 E+06 3

H 1958 to 1991 5 E+05 3

H 1992 to Present 2.0E+04 1E+05b

Sr Start to 1969 (see fission product) 90Sr Present 2.9 4

FP 1955 to 1960 30 dpm/750 mL

FP 1960-1965 100 dpm/1.5L (beta)

500 dpm/1.5L (gamma)

FP 1965 - ? 1 nCi/1.5L 234U 1990 to present 0.032

235U 1990 to present 0.036 238U 1990 to present 0.032

U-natural 1982 to 1986 1 μg/L 1 μg/L

DU 1954 to 1982 Not listed 5 μg/L 

DU 1982 to 1986 Same as U-nat same as U-nat

DU 1986-1994 0.5/1.5L 1 dpm/1.5L

RU 1990 to present Use 234U and isotopic ratios Use 234U and isotopic ratios

EU Start to mid-1960s 0.15 dpm/1.5 L

EU Mid-1960s to 1982 1 dpm/1.5 L

EU 1982 to 1986 1 dpm/L

EU 1986 to 1990 1dpm/1.5 L

HEU 1990 to present Use 235U and isotopic ratios Use 235U and isotopic ratios

SP (237Np) ~1959 to 1964 0.05 dpm/1.5L

SP 1964 to 2001 0.1 dpm/1.5L 237Np 2001 to present 0.035 c

238,239,240Pu 1954 to 1962 0.05dpm/1.5L

238,239,240Pu 1963 to 1981 0.1dpm/1.5 L 238Pu 1981 to 1988d 0.05 dpm/1.5 L 238Pu 1988 to present 0.07 dpm 239Pu, 240Pu 1981 to 1988d 0.07 dpm/1.5 L 239Pu, 240Pu 1988 to present 0.06 dpm 241Am Mid 1960s to1971 3 dpm/1.5 L 241Am 1971 to 1990 0.3 dpm/1.5 L 241Am 1990 to 1994 0.1 dpm/L 241Am 1994 to present 0.029 244Cm Mid 1960s to1971 3 dpm/1.5 L 244Cm 1971 to 1990 0.3 dpm/1.5 L 244Cm 1990 to 1994 0.1 dpm/L 244Cm 1994 to present 0.021 252Cf Mid 1960s to1971 3 dpm/1.5 L 252Cf 1971 to 1990 0.3 dpm/1.5 L 252Cf 1990 to 1994 0.1 dpm/L 252Cf 1994 to present 0.021

a. MDAs are given if less than reporting levels; generally the reporting level should be used as the limit of detection.

b. Measured values between 5E+4 and 1E+5 pCi/L are reported in the database as 0.1 μCi/L.

c. Some uncertainty about this value. See discussion in Chapter 4.

d. For routine samples; gross alpha method still used for special samples.

D.1.2 In Vitro Urine Analyses for Individual Radionuclides

Urine analysis for trivalent actinides (americium, curium, and californium)

Urinalysis for trivalent actinides began in the mid-1960s. From initiation to 1990, the reporting levels 

varied from about 1 dpm/1.5L to 0.3 dpm/L. In 1990, the reporting levels decreased to 0.1 dpm/L. The 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 197 of 248

MDAs since 1994 are shown in Table D-1. The counting method does not distinguish 241Am from 

244Cm from 252Cf. Interpretation of which radionuclide is being counted will have to come from 

particulars of the intake description. The claimant-favorable default is 241Am.

Urine analysis for plutonium

The reporting levels and MDAs are also listed in Table D-1. Results prior to 1981 (1988 for special 

samples) were for total alpha from plutonium isotopes. Plutonium-239 results are usually indicative of 

a mixture of plutonium isotopes. See 4.1.2 for discussion. 

Urine analysis for tritium

From startup until 1958, the reported MDA was 1 μCi/L. From 1958 until present, the MDA improved 

to the current level of 20 nCi/L; however, the reporting level remained at 1 μCi/ L, and decreased to 

the current value of 0.1 μCi/L. For tritium, the denominator has always been per liter of urine, and not 

1.5 L as for other radionuclides.

For current analyses, tritium results of ≤0.05 μCi/L are reported as <0.1μCi/L, from 0.05 to 0.1 μCi/L 

are reported as 0.1 μCi/L, and >0.1 μCi/ L are reported as measured. Only results greater than 5 

μCi/L were evaluated prior to computer evaluation of data.

Urine analysis for uranium

A variety of methods has been used historically to analyze uranium resulting in differences in the MDA 

over time, as shown in Table D-1. Activity fractions for depleted, recycled, and highly enriched 

uranium are listed in the glossary section. For earlier monitoring periods, the designations “enriched” 

and “depleted” analysis for uranium referred to the analysis and were not indicative of the degree of 

uranium enrichment. Natural background levels of uranium in excreta samples can interfere with 

interpretation of bioassay results; see the discussion in 4.1.2. 

Urine analysis for radiostrontium

The start date for urinalysis for radioisotopes of strontium was not specifically determined, other than 

“late 1950s.” This was referred to as the fission product analysis. From start until 1969 the same 

method was used, which involved separation of the strontium and beta counting. Both 89Sr and 90Sr 

would have been counted, but it is claimant-favorable to assume the result is all 90Sr. 

D.1.3 In Vitro Fecal Analysis

Fecal analysis has been done periodically since the 1950s for special purposes and was always done 

in conjunction with urine bioassay sampling. Usually the analysis was a gamma scan on the total 

sample with typical MDAs of 7 pCi for 241Am, 300 pCi for 238Pu, and 600 pCi for 238Pu plus 239Pu in 

weapons grade mixtures. If fecal samples were also analyzed by wet radiochemistry, the MDAs were 

approximately 1.5 times the same MDAs for urinalyses.

D.2 In Vivo MDAs and Reporting Practices at SRS

Whole body counting has been performed since 1960 and chest counting was initiated in the early 

1970s. In vivo detection capabilities are reviewed below.

D.2.1 Whole Body Counting

A variety of hardware and software applications have been used for whole body counting since 1960 

consistent with industry-wide improvements in the discipline. Whole body counting was not used for 

plutonium and americium bioassay due to their low energy photon emissions. Sometimes MDAs were 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 198 of 248

reported on the forms for the individual measurements and these should be used when available. 

Default MDAs for various radionuclides are shown in Table D-2. MDAs and reporting levels were 

synonymous. 

Table D-2. MDAs for whole body counting.

Radionuclide Period MDA (nCi)

54Mn Present 3.4

60Co Present 2.9

65Zn 1960 to ?? 5.1

65Zn Present 6.1

Zr/95Nb 1960 to ?? 2.2

106Ru 1960 to ?? 6.1

106Ru Present 36

125Sb Present 14

131I 1960 to ?? 1.4

134Cs Present 3.8

137Cs 1960 to ?? 1.0

137Cs Present 4.1

Ba/140La 1960 to ?? 9.3

144Ce 1960 to ?? 29

144Ce Present 69

152Eu Present 18

154Eu Present 8.4

235U Present 14

U-natural (235U) 1960 to ?? 62

237Np Present 14

For special investigational counts, the MDAs are approximately 0.71 times these values for whole 

body counting.

D.2.2 Chest Counting

No MDA information could be found for chest counting performed from 1966 to around 1971. From 

1971 to 1987 or 1988, using phoswich detectors, chest burdens of 4 to 10 nCi were reported for 238Pu. 

In 1987 or 1988, the 238Pu MDA is reported to be 60-70 nCi.

MDAs for later periods have been generated individually for counts by processing software, with 

typical values shown in Table D-3.

D.3 Tables of Chronic Intakes of Various Fission Product and Activation Product 

Radionuclides Based on MDA Whole Body Count on Last Day of the Period 

If either 95Zr or 95Nb is used as the radionuclide to maximize the dose to the organ of interest, then 

assume an equal intake of the other radionuclide to also have occurred.

If either 95Zr or 95Nb is used as the radionuclide to maximize the dose to the organ of interest, then 

assume an equal intake of the other radionuclide to also have occurred.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 199 of 248

Table D-3. MDAs for chest counting.

Radionuclide Time period MDA (nCi) 144Ce Present 0.31 152Eu Present 0.056 228Th Present 3.2 232Th Present 31 234U Present 30 235U Present 0.10 236U Present 89 238U Present 1.1

DU Present 1.2

RU Present 8.3

HEU Present 5.2 237Np Present 0.31 238Pu <1971 NA 238Pu 1971 to 1987 or 1988 4 to 10 238Pu 1987 to Present 130a

239Pu 1987 to Present 290a

240Pu 1987 to Present 110a

241Am 1987 to Present 0.13a

243Am 1987 to Present 0.12 244Cm 1987 to Present 78a

252Cf 1987 to Present 85a

a. Based on chest wall thickness of 50th male worker of 3.7 cm.

Note: For special investigational counts, the MDAs are approximately 

0.77 times the values for chest counting in Table D-3.

Table D-4. 95Zr chronic inhalation intake assessment based on MDA whole body count on 

last day of the period.

Solubility: Type M Type S

Particle size: 5 5

Analytical MDA: 2.2 nCi 2.2 nCi

Duration of intake Daily intake rate Cumulative intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

1 365 8.45E+02 3.08E+05 1.39E+02 9.57E+02 3.49E+05 1.57E+02

2 730 8.37E+02 6.11E+05 2.75E+02 9.48E+02 6.92E+05 3.12E+02

3 1,095 8.36E+02 9.15E+05 4.12E+02 9.48E+02 1.04E+06 4.68E+02

4 1,461 8.36E+02 1.22E+06 5.50E+02 9.48E+02 1.39E+06 6.24E+02

5 1,826 8.36E+02 1.53E+06 6.88E+02 9.48E+02 1.73E+06 7.80E+02

6 2,190 8.36E+02 1.83E+06 8.25E+02 9.48E+02 2.08E+06 9.35E+02

7 2,556 8.36E+02 2.14E+06 9.63E+02 9.48E+02 2.42E+06 1.09E+03

8 2,922 8.36E+02 2.44E+06 1.10E+03 9.48E+02 2.77E+06 1.25E+03

9 3,287 8.36E+02 2.75E+06 1.24E+03 9.48E+02 3.12E+06 1.40E+03

10 3,652 8.36E+02 3.05E+06 1.38E+03 9.48E+02 3.46E+06 1.56E+03

12 4,383 8.36E+02 3.66E+06 1.65E+03 9.48E+02 4.16E+06 1.87E+03

14 5,113 8.36E+02 4.27E+06 1.93E+03 9.48E+02 4.85E+06 2.18E+03

16 5,844 8.36E+02 4.89E+06 2.20E+03 9.48E+02 5.54E+06 2.50E+03

18 6,574 8.36E+02 5.50E+06 2.48E+03 9.48E+02 6.23E+06 2.81E+03

20 7,305 8.36E+02 6.11E+06 2.75E+03 9.48E+02 6.93E+06 3.12E+03

25 9,131 8.36E+02 7.63E+06 3.44E+03 9.48E+02 8.66E+06 3.90E+03

30 10,958 8.36E+02 9.16E+06 4.13E+03 9.48E+02 1.04E+07 4.68E+03

35 12,784 8.36E+02 1.07E+07 4.81E+03 9.48E+02 1.21E+07 5.46E+03

40 14,610 8.36E+02 1.22E+07 5.50E+03 9.48E+02 1.39E+07 6.24E+03

45 16,436 8.36E+02 1.37E+07 6.19E+03 9.48E+02 1.56E+07 7.02E+03

50 18,263 8.36E+02 1.53E+07 6.88E+03 9.48E+02 1.73E+07 7.80E+03
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 200 of 248

Table D-5. 95Nb chronic inhalation intake assessment based on MDA whole body count on 

last day of the period.

Solubility: Type M Type S

Particle size: 5 5

Analytical MDA: 2.2 nCi 2.2 nCi

Duration of intake Daily intake rate Cumulative intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

1 365 1.25E+03 4.55E+05 2.05E+02 1.34E+03 4.91E+05 2.21E+02

2 730 1.25E+03 9.10E+05 4.10E+02 1.34E+03 9.81E+05 4.42E+02

3 1,095 1.25E+03 1.37E+06 6.15E+02 1.34E+03 1.47E+06 6.63E+02

4 1,461 1.25E+03 1.82E+06 8.21E+02 1.34E+03 1.96E+06 8.84E+02

5 1,826 1.25E+03 2.28E+06 1.03E+03 1.34E+03 2.45E+06 1.11E+03

6 2,190 1.25E+03 2.73E+06 1.23E+03 1.34E+03 2.94E+06 1.33E+03

7 2,556 1.25E+03 3.19E+06 1.44E+03 1.34E+03 3.44E+06 1.55E+03

8 2,922 1.25E+03 3.64E+06 1.64E+03 1.34E+03 3.93E+06 1.77E+03

9 3,287 1.25E+03 4.10E+06 1.85E+03 1.34E+03 4.42E+06 1.99E+03

10 3,652 1.25E+03 4.55E+06 2.05E+03 1.34E+03 4.91E+06 2.21E+03

12 4,383 1.25E+03 5.47E+06 2.46E+03 1.34E+03 5.89E+06 2.65E+03

14 5,113 1.25E+03 6.38E+06 2.87E+03 1.34E+03 6.87E+06 3.10E+03

16 5,844 1.25E+03 7.29E+06 3.28E+03 1.34E+03 7.85E+06 3.54E+03

18 6,574 1.25E+03 8.20E+06 3.69E+03 1.34E+03 8.84E+06 3.98E+03

20 7,305 1.25E+03 9.11E+06 4.10E+03 1.34E+03 9.82E+06 4.42E+03

25 9,131 1.25E+03 1.14E+07 5.13E+03 1.34E+03 1.23E+07 5.53E+03

30 10,958 1.25E+03 1.37E+07 6.16E+03 1.34E+03 1.47E+07 6.63E+03

35 12,784 1.25E+03 1.59E+07 7.18E+03 1.34E+03 1.72E+07 7.74E+03

40 14,610 1.25E+03 1.82E+07 8.21E+03 1.34E+03 1.96E+07 8.84E+03

45 16,436 1.25E+03 2.05E+07 9.23E+03 1.34E+03 2.21E+07 9.95E+03

50 18,263 1.25E+03 2.28E+07 1.03E+04 1.34E+03 2.45E+07 1.11E+04

Table D-6. 54Mn chronic inhalation intake assessment 

based on MDA whole body count on last day of the 

period.

Solubility: Type M

Particle size: 5

Analytical MDA: 1 nCi

Duration of intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi)

1 365 2.37E+02 8.65E+04 3.90E+01

2 730 2.30E+02 1.68E+05 7.56E+01

3 1,095 2.30E+02 2.52E+05 1.13E+02

4 1,461 2.30E+02 3.36E+05 1.51E+02

5 1,826 2.30E+02 4.20E+05 1.89E+02

6 2,190 2.30E+02 5.04E+05 2.27E+02

7 2,556 2.30E+02 5.88E+05 2.65E+02

8 2,922 2.30E+02 6.72E+05 3.03E+02

9 3,287 2.30E+02 7.56E+05 3.41E+02

10 3,652 2.30E+02 8.40E+05 3.78E+02

12 4,383 2.30E+02 1.01E+06 4.54E+02

14 5,113 2.30E+02 1.18E+06 5.30E+02

16 5,844 2.30E+02 1.34E+06 6.05E+02

18 6,574 2.30E+02 1.51E+06 6.81E+02

20 7,305 2.30E+02 1.68E+06 7.57E+02

25 9,131 2.30E+02 2.10E+06 9.46E+02

30 10,958 2.30E+02 2.52E+06 1.14E+03

35 12,784 2.30E+02 2.94E+06 1.32E+03

40 14,610 2.30E+02 3.36E+06 1.51E+03

45 16,436 2.30E+02 3.78E+06 1.70E+03

50 18,263 2.30E+02 4.20E+06 1.89E+03
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 201 of 248

Table D-7. 131I chronic inhalation intake assessment based 

on MDA whole body count on last day of the period.

Solubility: Type F

Particle size: 5

Analytical MDA: 1.4 nCi

Duration of intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi)

1 365 1.48E+03 5.40E+05 2.43E+02

2 730 1.48E+03 1.08E+06 4.87E+02

3 1,095 1.48E+03 1.62E+06 7.30E+02

4 1,461 1.48E+03 2.16E+06 9.74E+02

5 1,826 1.48E+03 2.70E+06 1.22E+03

6 2,190 1.48E+03 3.24E+06 1.46E+03

7 2,556 1.48E+03 3.78E+06 1.70E+03

8 2,922 1.48E+03 4.32E+06 1.95E+03

9 3,287 1.48E+03 4.86E+06 2.19E+03

10 3,652 1.48E+03 5.40E+06 2.43E+03

12 4,383 1.48E+03 6.49E+06 2.92E+03

14 5,113 1.48E+03 7.57E+06 3.41E+03

16 5,844 1.48E+03 8.65E+06 3.90E+03

18 6,574 1.48E+03 9.73E+06 4.38E+03

20 7,305 1.48E+03 1.08E+07 4.87E+03

25 9,131 1.48E+03 1.35E+07 6.09E+03

30 10,958 1.48E+03 1.62E+07 7.31E+03

35 12,784 1.48E+03 1.89E+07 8.52E+03

40 14,610 1.48E+03 2.16E+07 9.74E+03

45 16,436 1.48E+03 2.43E+07 1.10E+04

50 18,263 1.48E+03 2.70E+07 1.22E+04

Table D-8. 154Eu chronic inhalation intake assessment 

based on MDA whole body count on last day of the period.

Solubility: Type M

Particle size: 5

Analytical MDA: 2.4 nCi

Duration of intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi)

1 365 2.63E+02 9.60E+04 4.32E+01

2 730 1.52E+02 1.11E+05 5.00E+01

3 1,095 1.12E+02 1.23E+05 5.52E+01

4 1,461 9.12E+01 1.33E+05 6.00E+01

5 1,826 7.87E+01 1.44E+05 6.47E+01

6 2,190 7.04E+01 1.54E+05 6.94E+01

7 2,556 6.46E+01 1.65E+05 7.44E+01

8 2,922 6.02E+01 1.76E+05 7.92E+01

9 3,287 5.70E+01 1.87E+05 8.44E+01

10 3,652 5.44E+01 1.99E+05 8.95E+01

12 4,383 5.08E+01 2.23E+05 1.00E+02

14 5,113 4.84E+01 2.47E+05 1.11E+02

16 5,844 4.68E+01 2.73E+05 1.23E+02

18 6,574 4.57E+01 3.00E+05 1.35E+02

20 7,305 4.49E+01 3.28E+05 1.48E+02

25 9,131 4.38E+01 4.00E+05 1.80E+02

30 10,958 4.32E+01 4.73E+05 2.13E+02

35 12,784 4.30E+01 5.50E+05 2.48E+02

40 14,610 4.29E+01 6.27E+05 2.82E+02

45 16,436 4.28E+01 7.03E+05 3.17E+02

50 18,263 4.28E+01 7.82E+05 3.52E+02
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 202 of 248

Table D-9. 65Zn chronic inhalation intake assessment 

based on MDA whole body count on last day of the period.

Solubility: Type S

Particle size: 5

Analytical MDA: 5.1nCi

Duration of intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi)

1 365 TBD

2 730

3 1,095

4 1,461

5 1,826

6 2,190

7 2,556

8 2,922

9 3,287

10 3,652

12 4,383

14 5,113

16 5,844

18 6,574

20 7,305

25 9,131

30 10,958

35 12,784

40 14,610

45 16,436

50 18,263

Table D-10. 140Ba chronic inhalation intake assessment 

based on MDA whole body count on last day of the period.

a

Solubility: Type F

Particle size: 5

Analytical MDA: 9.3 nCi

Duration of intake Daily intake rate Cumulative intake

(Years) (Days) (dpm/d) (dpm) (nCi)

1 365 TBD

2 730

3 1,095

4 1,461

5 1,826

6 2,190

7 2,556

8 2,922

9 3,287

10 3,652

12 4,383

14 5,113

16 5,844

18 6,574

20 7,305

25 9,131

30 10,958

35 12,784

40 14,610

45 16,436

50 18,263

a. Assume equal intake of La-140.
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 203 of 248

D.4 Tables of 50-Yr Committed Dose from Chronic Intakes at Whole Body Count MDA 

Values
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 204 of 248

If either 140Ba or 140La is used as the radionuclide to maximize the dose to the organ of interest, then assume an equal intake of the 

other radionuclide to also have occurred.

Table D-11. Ba-140, Type F, f1=0.1, 50-year equivalent dose for years of chronic intake (Sv) MDA = 9.3 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 1.75E-05 3.50E-05 5.25E-05 7.00E-05 8.75E-05 1.05E-04 1.23E-04 1.40E-04 1.58E-04 1.75E-04 2.10E-04

Urinary Bladder 2.16E-05 4.32E-05 6.48E-05 8.64E-05 1.08E-04 1.30E-04 1.51E-04 1.73E-04 1.94E-04 2.16E-04 2.59E-04

Brain 1.15E-05 2.31E-05 3.45E-05 4.60E-05 5.75E-05 6.90E-05 8.05E-05 9.20E-05 1.04E-04 1.15E-04 1.38E-04

Breast 7.18E-06 1.44E-05 2.15E-05 2.87E-05 3.59E-05 4.31E-05 5.03E-05 5.74E-05 6.46E-05 7.18E-05 8.62E-05

Gall Bladder 1.54E-05 3.07E-05 4.62E-05 6.16E-05 7.70E-05 9.24E-05 1.08E-04 1.23E-04 1.39E-04 1.54E-04 1.85E-04

Heart Wall 9.93E-06 1.99E-05 2.98E-05 3.97E-05 4.97E-05 5.96E-05 6.95E-05 7.94E-05 8.94E-05 9.93E-05 1.19E-04

Kidneys 1.50E-05 3.01E-05 4.50E-05 6.00E-05 7.50E-05 9.00E-05 1.05E-04 1.20E-04 1.35E-04 1.50E-04 1.80E-04

Liver 1.03E-05 2.06E-05 3.09E-05 4.12E-05 5.15E-05 6.18E-05 7.21E-05 8.24E-05 9.27E-05 1.03E-04 1.24E-04

Muscle 1.36E-05 2.71E-05 4.08E-05 5.44E-05 6.80E-05 8.16E-05 9.52E-05 1.09E-04 1.22E-04 1.36E-04 1.63E-04

Ovaries 4.32E-05 8.65E-05 1.30E-04 1.73E-04 2.16E-04 2.59E-04 3.02E-04 3.46E-04 3.89E-04 4.32E-04 5.18E-04

Pancreas 1.32E-05 2.65E-05 3.96E-05 5.28E-05 6.60E-05 7.92E-05 9.24E-05 1.06E-04 1.19E-04 1.32E-04 1.58E-04

Testes 8.91E-06 1.78E-05 2.67E-05 3.56E-05 4.46E-05 5.35E-05 6.24E-05 7.13E-05 8.02E-05 8.91E-05 1.07E-04

Thyroid 1.01E-05 2.02E-05 3.03E-05 4.04E-05 5.05E-05 6.06E-05 7.07E-05 8.08E-05 9.09E-05 1.01E-04 1.21E-04

R.B.M. 1.13E-04 2.26E-04 3.39E-04 4.52E-04 5.65E-04 6.78E-04 7.91E-04 9.04E-04 1.02E-03 1.13E-03 1.36E-03

Bone Surface 1.59E-04 3.17E-04 4.77E-04 6.36E-04 7.95E-04 9.54E-04 1.11E-03 1.27E-03 1.43E-03 1.59E-03 1.91E-03

Stomach 1.76E-05 3.53E-05 5.28E-05 7.04E-05 8.80E-05 1.06E-04 1.23E-04 1.41E-04 1.58E-04 1.76E-04 2.11E-04

S.I. 4.63E-05 9.26E-05 1.39E-04 1.85E-04 2.32E-04 2.78E-04 3.24E-04 3.70E-04 4.17E-04 4.63E-04 5.56E-04

U.L.I. 2.69E-04 5.38E-04 8.07E-04 1.08E-03 1.35E-03 1.61E-03 1.88E-03 2.15E-03 2.42E-03 2.69E-03 3.23E-03

L.L.I. 8.61E-04 1.72E-03 2.58E-03 3.44E-03 4.31E-03 5.17E-03 6.03E-03 6.89E-03 7.75E-03 8.61E-03 1.03E-02

Skin 8.28E-06 1.66E-05 2.48E-05 3.31E-05 4.14E-05 4.97E-05 5.80E-05 6.62E-05 7.45E-05 8.28E-05 9.94E-05

Spleen 1.06E-05 2.12E-05 3.18E-05 4.24E-05 5.30E-05 6.36E-05 7.42E-05 8.48E-05 9.54E-05 1.06E-04 1.27E-04

Thymus 9.98E-06 2.00E-05 2.99E-05 3.99E-05 4.99E-05 5.99E-05 6.99E-05 7.98E-05 8.98E-05 9.98E-05 1.20E-04

Uterus 2.18E-05 4.36E-05 6.54E-05 8.72E-05 1.09E-04 1.31E-04 1.53E-04 1.74E-04 1.96E-04 2.18E-04 2.62E-04

ET 5.81E-04 1.16E-03 1.74E-03 2.32E-03 2.91E-03 3.49E-03 4.07E-03 4.65E-03 5.23E-03 5.81E-03 6.97E-03

ET1 1.51E-01 3.02E-01 4.53E-01 6.04E-01 7.55E-01 9.06E-01 1.06E+00 1.21E+00 1.36E+00 1.51E+00 1.81E+00

ET2 4.31E-04 8.62E-04 1.29E-03 1.72E-03 2.16E-03 2.59E-03 3.02E-03 3.45E-03 3.88E-03 4.31E-03 5.17E-03

Lung 1.23E-05 2.46E-05 3.69E-05 4.92E-05 6.15E-05 7.38E-05 8.61E-05 9.84E-05 1.11E-04 1.23E-04 1.48E-04

Colon 5.25E-04 1.05E-03 1.58E-03 2.10E-03 2.63E-03 3.15E-03 3.68E-03 4.20E-03 4.73E-03 5.25E-03 6.30E-03

LN(ET) 4.08E-04 8.17E-04 1.22E-03 1.63E-03 2.04E-03 2.45E-03 2.86E-03 3.26E-03 3.67E-03 4.08E-03 4.90E-03

BBsec 1.53E-05 3.07E-05 4.59E-05 6.12E-05 7.65E-05 9.18E-05 1.07E-04 1.22E-04 1.38E-04 1.53E-04 1.84E-04

BBbas 1.45E-05 2.90E-05 4.35E-05 5.80E-05 7.25E-05 8.70E-05 1.02E-04 1.16E-04 1.31E-04 1.45E-04 1.74E-04

bb 1.12E-05 2.23E-05 3.36E-05 4.48E-05 5.60E-05 6.72E-05 7.84E-05 8.96E-05 1.01E-04 1.12E-04 1.34E-04

AI 1.08E-05 2.15E-05 3.24E-05 4.32E-05 5.40E-05 6.48E-05 7.56E-05 8.64E-05 9.72E-05 1.08E-04 1.30E-04

LN(TH) 1.06E-05 2.12E-05 3.18E-05 4.24E-05 5.30E-05 6.36E-05 7.42E-05 8.48E-05 9.54E-05 1.06E-04 1.27E-04

Esophagus 9.98E-06 2.00E-05 2.99E-05 3.99E-05 4.99E-05 5.99E-05 6.99E-05 7.98E-05 8.98E-05 9.98E-05 1.20E-04

Gonads 4.32E-05 8.65E-05 1.30E-04 1.73E-04 2.16E-04 2.59E-04 3.02E-04 3.46E-04 3.89E-04 4.32E-04 5.18E-04

Intake Rate (pCi/d) 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2

Days 365 730 1095 1461 1826 2190 2556 2922 3287 3652 4383

Intake (pCi) 1.80E+06 3.60E+06 5.40E+06 7.21E+06 9.01E+06 1.08E+07 1.26E+07 1.44E+07 1.62E+07 1.80E+07 2.16E+07
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 205 of 248

Table D-11 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 2.45E-04 2.80E-04 3.15E-04 3.50E-04 4.38E-04 5.25E-04 6.13E-04 7.00E-04 7.88E-04 8.75E-04

Urinary Bladder 3.02E-04 3.46E-04 3.89E-04 4.32E-04 5.40E-04 6.48E-04 7.56E-04 8.64E-04 9.72E-04 1.08E-03

Brain 1.61E-04 1.84E-04 2.07E-04 2.30E-04 2.88E-04 3.45E-04 4.03E-04 4.60E-04 5.18E-04 5.75E-04

Breast 1.01E-04 1.15E-04 1.29E-04 1.44E-04 1.80E-04 2.15E-04 2.51E-04 2.87E-04 3.23E-04 3.59E-04

Gall Bladder 2.16E-04 2.46E-04 2.77E-04 3.08E-04 3.85E-04 4.62E-04 5.39E-04 6.16E-04 6.93E-04 7.70E-04

Heart Wall 1.39E-04 1.59E-04 1.79E-04 1.99E-04 2.48E-04 2.98E-04 3.48E-04 3.97E-04 4.47E-04 4.97E-04

Kidneys 2.10E-04 2.40E-04 2.70E-04 3.00E-04 3.75E-04 4.50E-04 5.25E-04 6.00E-04 6.75E-04 7.50E-04

Liver 1.44E-04 1.65E-04 1.85E-04 2.06E-04 2.58E-04 3.09E-04 3.61E-04 4.12E-04 4.64E-04 5.15E-04

Muscle 1.90E-04 2.18E-04 2.45E-04 2.72E-04 3.40E-04 4.08E-04 4.76E-04 5.44E-04 6.12E-04 6.80E-04

Ovaries 6.05E-04 6.91E-04 7.78E-04 8.64E-04 1.08E-03 1.30E-03 1.51E-03 1.73E-03 1.94E-03 2.16E-03

Pancreas 1.85E-04 2.11E-04 2.38E-04 2.64E-04 3.30E-04 3.96E-04 4.62E-04 5.28E-04 5.94E-04 6.60E-04

Testes 1.25E-04 1.43E-04 1.60E-04 1.78E-04 2.23E-04 2.67E-04 3.12E-04 3.56E-04 4.01E-04 4.46E-04

Thyroid 1.41E-04 1.62E-04 1.82E-04 2.02E-04 2.53E-04 3.03E-04 3.54E-04 4.04E-04 4.55E-04 5.05E-04

R.B.M. 1.58E-03 1.81E-03 2.03E-03 2.26E-03 2.83E-03 3.39E-03 3.96E-03 4.52E-03 5.09E-03 5.65E-03

Bone Surface 2.23E-03 2.54E-03 2.86E-03 3.18E-03 3.98E-03 4.77E-03 5.57E-03 6.36E-03 7.16E-03 7.95E-03

Stomach 2.46E-04 2.82E-04 3.17E-04 3.52E-04 4.40E-04 5.28E-04 6.16E-04 7.04E-04 7.92E-04 8.80E-04

S.I. 6.48E-04 7.41E-04 8.33E-04 9.26E-04 1.16E-03 1.39E-03 1.62E-03 1.85E-03 2.08E-03 2.32E-03

U.L.I. 3.77E-03 4.30E-03 4.84E-03 5.38E-03 6.73E-03 8.07E-03 9.42E-03 1.08E-02 1.21E-02 1.35E-02

L.L.I. 1.21E-02 1.38E-02 1.55E-02 1.72E-02 2.15E-02 2.58E-02 3.01E-02 3.44E-02 3.87E-02 4.31E-02

Skin 1.16E-04 1.32E-04 1.49E-04 1.66E-04 2.07E-04 2.48E-04 2.90E-04 3.31E-04 3.73E-04 4.14E-04

Spleen 1.48E-04 1.70E-04 1.91E-04 2.12E-04 2.65E-04 3.18E-04 3.71E-04 4.24E-04 4.77E-04 5.30E-04

Thymus 1.40E-04 1.60E-04 1.80E-04 2.00E-04 2.50E-04 2.99E-04 3.49E-04 3.99E-04 4.49E-04 4.99E-04

Uterus 3.05E-04 3.49E-04 3.92E-04 4.36E-04 5.45E-04 6.54E-04 7.63E-04 8.72E-04 9.81E-04 1.09E-03

ET 8.13E-03 9.30E-03 1.05E-02 1.16E-02 1.45E-02 1.74E-02 2.03E-02 2.32E-02 2.61E-02 2.91E-02

ET1 2.11E+00 2.42E+00 2.72E+00 3.02E+00 3.78E+00 4.53E+00 5.29E+00 6.04E+00 6.80E+00 7.55E+00

ET2 6.03E-03 6.90E-03 7.76E-03 8.62E-03 1.08E-02 1.29E-02 1.51E-02 1.72E-02 1.94E-02 2.16E-02

Lung 1.72E-04 1.97E-04 2.21E-04 2.46E-04 3.08E-04 3.69E-04 4.31E-04 4.92E-04 5.54E-04 6.15E-04

Colon 7.35E-03 8.40E-03 9.45E-03 1.05E-02 1.31E-02 1.58E-02 1.84E-02 2.10E-02 2.36E-02 2.63E-02

LN(ET) 5.71E-03 6.53E-03 7.34E-03 8.16E-03 1.02E-02 1.22E-02 1.43E-02 1.63E-02 1.84E-02 2.04E-02

BBsec 2.14E-04 2.45E-04 2.75E-04 3.06E-04 3.83E-04 4.59E-04 5.36E-04 6.12E-04 6.89E-04 7.65E-04

BBbas 2.03E-04 2.32E-04 2.61E-04 2.90E-04 3.63E-04 4.35E-04 5.08E-04 5.80E-04 6.53E-04 7.25E-04

bb 1.57E-04 1.79E-04 2.02E-04 2.24E-04 2.80E-04 3.36E-04 3.92E-04 4.48E-04 5.04E-04 5.60E-04

AI 1.51E-04 1.73E-04 1.94E-04 2.16E-04 2.70E-04 3.24E-04 3.78E-04 4.32E-04 4.86E-04 5.40E-04

LN(TH) 1.48E-04 1.70E-04 1.91E-04 2.12E-04 2.65E-04 3.18E-04 3.71E-04 4.24E-04 4.77E-04 5.30E-04

Esophagus 1.40E-04 1.60E-04 1.80E-04 2.00E-04 2.50E-04 2.99E-04 3.49E-04 3.99E-04 4.49E-04 4.99E-04

Gonads 6.05E-04 6.91E-04 7.78E-04 8.64E-04 1.08E-03 1.30E-03 1.51E-03 1.73E-03 1.94E-03 2.16E-03

Intake Rate (pCi/d) 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2 4934.2

Days 5113 5844 6574 7305 9131 10958 12784 14610 16436 18263

Intake (pCi) 2.52E+07 2.88E+07 3.24E+07 3.60E+07 4.51E+07 5.41E+07 6.31E+07 7.21E+07 8.11E+07 9.01E+07
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 206 of 248

Table D-12. Ce-144, Type W, 50-year equivalent dose for years of chronic intake (Sv) (MDA = 29 nCi). Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 3.24E-05 4.85E-05 6.64E-05 8.63E-05 1.06E-04 1.26E-04 1.47E-04 1.68E-04 1.89E-04 2.10E-04 2.53E-04

Urinary Bladder 2.82E-05 4.22E-05 5.78E-05 7.51E-05 9.23E-05 1.10E-04 1.28E-04 1.46E-04 1.65E-04 1.83E-04 2.20E-04

Brain 2.80E-05 4.20E-05 5.75E-05 7.46E-05 9.18E-05 1.09E-04 1.27E-04 1.46E-04 1.64E-04 1.82E-04 2.18E-04

Breast 2.90E-05 4.35E-05 5.95E-05 7.74E-05 9.51E-05 1.13E-04 1.32E-04 1.51E-04 1.70E-04 1.89E-04 2.26E-04

Gall Bladder 3.46E-05 5.19E-05 7.11E-05 9.23E-05 1.14E-04 1.35E-04 1.58E-04 1.80E-04 2.03E-04 2.25E-04 2.70E-04

Heart Wall 3.11E-05 4.66E-05 6.38E-05 8.29E-05 1.02E-04 1.21E-04 1.41E-04 1.62E-04 1.82E-04 2.02E-04 2.43E-04

Kidneys 3.05E-05 4.56E-05 6.25E-05 8.12E-05 9.98E-05 1.19E-04 1.39E-04 1.58E-04 1.78E-04 1.98E-04 2.38E-04

Liver 2.33E-03 3.48E-03 4.77E-03 6.20E-03 7.62E-03 9.06E-03 1.06E-02 1.21E-02 1.36E-02 1.51E-02 1.81E-02

Muscle 2.89E-05 4.33E-05 5.92E-05 7.70E-05 9.46E-05 1.13E-04 1.31E-04 1.50E-04 1.69E-04 1.88E-04 2.25E-04

Ovaries 2.91E-05 4.36E-05 5.97E-05 7.76E-05 9.54E-05 1.13E-04 1.32E-04 1.51E-04 1.70E-04 1.89E-04 2.27E-04

Pancreas 3.13E-05 4.70E-05 6.43E-05 8.35E-05 1.03E-04 1.22E-04 1.43E-04 1.63E-04 1.83E-04 2.04E-04 2.44E-04

Testes 2.77E-05 4.15E-05 5.69E-05 7.39E-05 9.08E-05 1.08E-04 1.26E-04 1.44E-04 1.62E-04 1.80E-04 2.16E-04

Thyroid 2.83E-05 4.24E-05 5.80E-05 7.54E-05 9.27E-05 1.10E-04 1.29E-04 1.47E-04 1.66E-04 1.84E-04 2.21E-04

R.B.M. 4.45E-04 6.67E-04 9.13E-04 1.19E-03 1.46E-03 1.73E-03 2.02E-03 2.31E-03 2.60E-03 2.89E-03 3.47E-03

Bone Surface 7.87E-04 1.18E-03 1.62E-03 2.10E-03 2.58E-03 3.07E-03 3.58E-03 4.09E-03 4.61E-03 5.12E-03 6.14E-03

Stomach 4.15E-05 6.22E-05 8.52E-05 1.11E-04 1.36E-04 1.62E-04 1.89E-04 2.16E-04 2.43E-04 2.70E-04 3.24E-04

S.I. 6.50E-05 9.74E-05 1.33E-04 1.73E-04 2.13E-04 2.53E-04 2.96E-04 3.38E-04 3.80E-04 4.23E-04 5.07E-04

U.L.I. 2.50E-04 3.75E-04 5.13E-04 6.66E-04 8.19E-04 9.74E-04 1.14E-03 1.30E-03 1.46E-03 1.62E-03 1.95E-03

L.L.I. 6.73E-04 1.01E-03 1.38E-03 1.79E-03 2.21E-03 2.62E-03 3.06E-03 3.50E-03 3.94E-03 4.38E-03 5.25E-03

Skin 2.81E-05 4.21E-05 5.76E-05 7.49E-05 9.20E-05 1.09E-04 1.28E-04 1.46E-04 1.64E-04 1.83E-04 2.19E-04

Spleen 2.90E-05 4.35E-05 5.96E-05 7.74E-05 9.51E-05 1.13E-04 1.32E-04 1.51E-04 1.70E-04 1.89E-04 2.26E-04

Thymus 2.93E-05 4.39E-05 6.02E-05 7.82E-05 9.61E-05 1.14E-04 1.33E-04 1.53E-04 1.72E-04 1.91E-04 2.29E-04

Uterus 2.85E-05 4.28E-05 5.86E-05 7.61E-05 9.35E-05 1.11E-04 1.30E-04 1.48E-04 1.67E-04 1.86E-04 2.23E-04

Colon 4.33E-04 6.49E-04 8.88E-04 1.15E-03 1.42E-03 1.69E-03 1.97E-03 2.25E-03 2.53E-03 2.81E-03 3.38E-03

ET 3.56E-04 5.33E-04 7.30E-04 9.49E-04 1.17E-03 1.39E-03 1.62E-03 1.85E-03 2.08E-03 2.31E-03 2.78E-03

ET1 6.63E-02 9.94E-02 1.36E-01 1.77E-01 2.17E-01 2.58E-01 3.02E-01 3.45E-01 3.88E-01 4.31E-01 5.17E-01

ET2 2.90E-04 4.35E-04 5.95E-04 7.74E-04 9.51E-04 1.13E-03 1.32E-03 1.51E-03 1.70E-03 1.89E-03 2.26E-03

Lung 2.40E-03 3.59E-03 4.92E-03 6.39E-03 7.86E-03 9.35E-03 1.09E-02 1.25E-02 1.40E-02 1.56E-02 1.87E-02

LN(ET) 1.19E-04 1.78E-04 2.43E-04 3.16E-04 3.89E-04 4.62E-04 5.40E-04 6.17E-04 6.94E-04 7.71E-04 9.25E-04

BBsec 3.66E-03 5.49E-03 7.51E-03 9.76E-03 1.20E-02 1.43E-02 1.67E-02 1.90E-02 2.14E-02 2.38E-02 2.86E-02

BBbas 3.20E-03 4.80E-03 6.58E-03 8.54E-03 1.05E-02 1.25E-02 1.46E-02 1.67E-02 1.88E-02 2.08E-02 2.50E-02

bb 2.00E-03 3.00E-03 4.11E-03 5.34E-03 6.57E-03 7.81E-03 9.12E-03 1.04E-02 1.17E-02 1.30E-02 1.56E-02

AI 1.77E-03 2.65E-03 3.62E-03 4.71E-03 5.78E-03 6.88E-03 8.03E-03 9.18E-03 1.03E-02 1.15E-02 1.38E-02

LN(TH) 2.52E-04 3.78E-04 5.17E-04 6.72E-04 8.26E-04 9.82E-04 1.15E-03 1.31E-03 1.47E-03 1.64E-03 1.97E-03

Esophagus 2.93E-05 4.39E-05 6.02E-05 7.82E-05 9.61E-05 1.14E-04 1.33E-04 1.53E-04 1.72E-04 1.91E-04 2.29E-04

Gonads 2.91E-05 4.36E-05 5.97E-05 7.76E-05 9.54E-05 1.13E-04 1.32E-04 1.51E-04 1.70E-04 1.89E-04 2.27E-04
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 207 of 248

Table D-12 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 2.95E-04 3.37E-04 3.79E-04 4.21E-04 5.26E-04 6.32E-04 7.37E-04 8.42E-04 9.47E-04 1.05E-03

Urinary Bladder 2.56E-04 2.93E-04 3.30E-04 3.66E-04 4.58E-04 5.49E-04 6.41E-04 7.32E-04 8.24E-04 9.15E-04

Brain 2.55E-04 2.91E-04 3.28E-04 3.64E-04 4.55E-04 5.46E-04 6.37E-04 7.28E-04 8.19E-04 9.10E-04

Breast 2.64E-04 3.02E-04 3.40E-04 3.77E-04 4.72E-04 5.66E-04 6.60E-04 7.55E-04 8.49E-04 9.43E-04

Gall Bladder 3.15E-04 3.60E-04 4.05E-04 4.50E-04 5.63E-04 6.76E-04 7.88E-04 9.01E-04 1.01E-03 1.13E-03

Heart Wall 2.83E-04 3.24E-04 3.64E-04 4.04E-04 5.05E-04 6.07E-04 7.08E-04 8.09E-04 9.10E-04 1.01E-03

Kidneys 2.77E-04 3.17E-04 3.56E-04 3.96E-04 4.95E-04 5.94E-04 6.93E-04 7.92E-04 8.91E-04 9.90E-04

Liver 2.12E-02 2.42E-02 2.72E-02 3.02E-02 3.78E-02 4.54E-02 5.29E-02 6.05E-02 6.80E-02 7.56E-02

Muscle 2.63E-04 3.00E-04 3.38E-04 3.75E-04 4.69E-04 5.63E-04 6.57E-04 7.51E-04 8.45E-04 9.38E-04

Ovaries 2.65E-04 3.03E-04 3.41E-04 3.78E-04 4.73E-04 5.68E-04 6.62E-04 7.57E-04 8.51E-04 9.46E-04

Pancreas 2.85E-04 3.26E-04 3.67E-04 4.07E-04 5.09E-04 6.11E-04 7.13E-04 8.15E-04 9.17E-04 1.02E-03

Testes 2.52E-04 2.88E-04 3.24E-04 3.60E-04 4.50E-04 5.41E-04 6.31E-04 7.21E-04 8.11E-04 9.01E-04

Thyroid 2.58E-04 2.94E-04 3.31E-04 3.68E-04 4.60E-04 5.52E-04 6.44E-04 7.36E-04 8.28E-04 9.20E-04

R.B.M. 4.05E-03 4.63E-03 5.21E-03 5.78E-03 7.23E-03 8.68E-03 1.01E-02 1.16E-02 1.30E-02 1.45E-02

Bone Surface 7.16E-03 8.19E-03 9.21E-03 1.02E-02 1.28E-02 1.54E-02 1.79E-02 2.05E-02 2.30E-02 2.56E-02

Stomach 3.78E-04 4.32E-04 4.86E-04 5.40E-04 6.75E-04 8.10E-04 9.45E-04 1.08E-03 1.21E-03 1.35E-03

S.I. 5.92E-04 6.76E-04 7.61E-04 8.45E-04 1.06E-03 1.27E-03 1.48E-03 1.69E-03 1.90E-03 2.11E-03

U.L.I. 2.27E-03 2.60E-03 2.92E-03 3.25E-03 4.06E-03 4.88E-03 5.69E-03 6.50E-03 7.31E-03 8.13E-03

L.L.I. 6.13E-03 7.00E-03 7.88E-03 8.75E-03 1.09E-02 1.31E-02 1.53E-02 1.75E-02 1.97E-02 2.19E-02

Skin 2.56E-04 2.92E-04 3.29E-04 3.65E-04 4.57E-04 5.48E-04 6.39E-04 7.30E-04 8.22E-04 9.13E-04

Spleen 2.64E-04 3.02E-04 3.40E-04 3.77E-04 4.72E-04 5.66E-04 6.61E-04 7.55E-04 8.49E-04 9.44E-04

Thymus 2.67E-04 3.05E-04 3.43E-04 3.81E-04 4.77E-04 5.72E-04 6.67E-04 7.63E-04 8.58E-04 9.53E-04

Uterus 2.60E-04 2.97E-04 3.34E-04 3.71E-04 4.64E-04 5.57E-04 6.50E-04 7.42E-04 8.35E-04 9.28E-04

Colon 3.94E-03 4.50E-03 5.07E-03 5.63E-03 7.04E-03 8.44E-03 9.85E-03 1.13E-02 1.27E-02 1.41E-02

ET 3.24E-03 3.70E-03 4.17E-03 4.63E-03 5.79E-03 6.94E-03 8.10E-03 9.26E-03 1.04E-02 1.16E-02

ET1 6.03E-01 6.90E-01 7.76E-01 8.62E-01 1.08E+00 1.29E+00 1.51E+00 1.72E+00 1.94E+00 2.16E+00

ET2 2.64E-03 3.02E-03 3.40E-03 3.77E-03 4.72E-03 5.66E-03 6.60E-03 7.55E-03 8.49E-03 9.43E-03

Lung 2.18E-02 2.50E-02 2.81E-02 3.12E-02 3.90E-02 4.68E-02 5.46E-02 6.24E-02 7.02E-02 7.80E-02

LN(ET) 1.08E-03 1.23E-03 1.39E-03 1.54E-03 1.93E-03 2.31E-03 2.70E-03 3.09E-03 3.47E-03 3.86E-03

BBsec 3.33E-02 3.81E-02 4.29E-02 4.76E-02 5.95E-02 7.14E-02 8.33E-02 9.52E-02 1.07E-01 1.19E-01

BBbas 2.92E-02 3.33E-02 3.75E-02 4.17E-02 5.21E-02 6.25E-02 7.29E-02 8.33E-02 9.38E-02 1.04E-01

bb 1.82E-02 2.08E-02 2.35E-02 2.61E-02 3.26E-02 3.91E-02 4.56E-02 5.21E-02 5.86E-02 6.52E-02

AI 1.61E-02 1.84E-02 2.07E-02 2.30E-02 2.87E-02 3.44E-02 4.02E-02 4.59E-02 5.16E-02 5.74E-02

LN(TH) 2.29E-03 2.62E-03 2.95E-03 3.28E-03 4.10E-03 4.92E-03 5.73E-03 6.55E-03 7.37E-03 8.19E-03

Esophagus 2.67E-04 3.05E-04 3.43E-04 3.81E-04 4.77E-04 5.72E-04 6.67E-04 7.63E-04 8.58E-04 9.53E-04

Gonads 2.65E-04 3.03E-04 3.41E-04 3.78E-04 4.73E-04 5.68E-04 6.62E-04 7.57E-04 8.51E-04 9.46E-04
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 208 of 248

Table D-13. Co-60, Type M, 50-year equivalent dose for years of chronic intake (Sv) MDA = 1 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 5.58E-06 8.98E-06 1.22E-05 1.54E-05 1.86E-05 2.18E-05 2.51E-05 2.85E-05 3.19E-05 3.53E-05 4.22E-05

Urinary Bladder 2.89E-06 4.66E-06 6.33E-06 7.98E-06 9.63E-06 1.13E-05 1.30E-05 1.48E-05 1.65E-05 1.83E-05 2.19E-05

Brain 2.08E-06 3.34E-06 4.54E-06 5.73E-06 6.92E-06 8.12E-06 9.35E-06 1.06E-05 1.19E-05 1.31E-05 1.57E-05

Breast 4.53E-06 7.28E-06 9.89E-06 1.25E-05 1.51E-05 1.77E-05 2.04E-05 2.31E-05 2.58E-05 2.86E-05 3.42E-05

Gall Bladder 4.69E-06 7.55E-06 1.03E-05 1.29E-05 1.56E-05 1.83E-05 2.11E-05 2.39E-05 2.68E-05 2.97E-05 3.55E-05

Heart Wall 7.01E-06 1.13E-05 1.53E-05 1.93E-05 2.33E-05 2.74E-05 3.15E-05 3.58E-05 4.00E-05 4.43E-05 5.30E-05

Kidneys 3.79E-06 6.10E-06 8.29E-06 1.04E-05 1.26E-05 1.48E-05 1.71E-05 1.93E-05 2.16E-05 2.40E-05 2.87E-05

Liver 7.80E-06 1.25E-05 1.70E-05 2.15E-05 2.60E-05 3.05E-05 3.51E-05 3.98E-05 4.45E-05 4.93E-05 5.90E-05

Muscle 3.45E-06 5.55E-06 7.54E-06 9.50E-06 1.15E-05 1.35E-05 1.55E-05 1.76E-05 1.97E-05 2.18E-05 2.61E-05

Ovaries 3.99E-06 6.42E-06 8.72E-06 1.10E-05 1.33E-05 1.56E-05 1.80E-05 2.04E-05 2.28E-05 2.52E-05 3.02E-05

Pancreas 4.95E-06 7.97E-06 1.08E-05 1.37E-05 1.65E-05 1.94E-05 2.23E-05 2.53E-05 2.83E-05 3.13E-05 3.75E-05

Testes 2.21E-06 3.56E-06 4.83E-06 6.09E-06 7.36E-06 8.64E-06 9.95E-06 1.13E-05 1.26E-05 1.40E-05 1.67E-05

Thyroid 3.40E-06 5.47E-06 7.43E-06 9.37E-06 1.13E-05 1.33E-05 1.53E-05 1.73E-05 1.94E-05 2.15E-05 2.57E-05

R.B.M. 3.88E-06 6.25E-06 8.49E-06 1.07E-05 1.29E-05 1.52E-05 1.75E-05 1.98E-05 2.22E-05 2.45E-05 2.94E-05

Bone Surface 3.61E-06 5.81E-06 7.89E-06 9.95E-06 1.20E-05 1.41E-05 1.62E-05 1.84E-05 2.06E-05 2.28E-05 2.73E-05

Stomach 4.13E-06 6.64E-06 9.02E-06 1.14E-05 1.37E-05 1.61E-05 1.86E-05 2.11E-05 2.36E-05 2.61E-05 3.12E-05

S.I. 4.09E-06 6.58E-06 8.94E-06 1.13E-05 1.36E-05 1.60E-05 1.84E-05 2.09E-05 2.33E-05 2.58E-05 3.09E-05

U.L.I. 5.34E-06 8.60E-06 1.17E-05 1.47E-05 1.78E-05 2.09E-05 2.41E-05 2.73E-05 3.05E-05 3.38E-05 4.04E-05

L.L.I. 7.86E-06 1.27E-05 1.72E-05 2.17E-05 2.62E-05 3.07E-05 3.54E-05 4.01E-05 4.49E-05 4.97E-05 5.95E-05

Skin 2.23E-06 3.58E-06 4.87E-06 6.14E-06 7.41E-06 8.70E-06 1.00E-05 1.14E-05 1.27E-05 1.41E-05 1.68E-05

Spleen 4.41E-06 7.09E-06 9.64E-06 1.22E-05 1.47E-05 1.72E-05 1.98E-05 2.25E-05 2.52E-05 2.79E-05 3.33E-05

Thymus 5.51E-06 8.87E-06 1.21E-05 1.52E-05 1.84E-05 2.15E-05 2.48E-05 2.81E-05 3.15E-05 3.48E-05 4.17E-05

Uterus 3.32E-06 5.35E-06 7.27E-06 9.16E-06 1.11E-05 1.30E-05 1.50E-05 1.70E-05 1.90E-05 2.10E-05 2.51E-05

Colon 6.43E-06 1.04E-05 1.41E-05 1.77E-05 2.14E-05 2.51E-05 2.90E-05 3.28E-05 3.67E-05 4.07E-05 4.86E-05

ET 3.84E-05 6.19E-05 8.41E-05 1.06E-04 1.28E-04 1.50E-04 1.73E-04 1.96E-04 2.20E-04 2.43E-04 2.91E-04

ET1 1.70E-03 2.74E-03 3.72E-03 4.69E-03 5.66E-03 6.65E-03 7.66E-03 8.68E-03 9.71E-03 1.08E-02 1.29E-02

ET2 3.68E-05 5.92E-05 8.04E-05 1.01E-04 1.23E-04 1.44E-04 1.66E-04 1.88E-04 2.10E-04 2.33E-04 2.78E-04

Lung 4.28E-05 6.89E-05 9.37E-05 1.18E-04 1.43E-04 1.68E-04 1.93E-04 2.19E-04 2.45E-04 2.71E-04 3.24E-04

LN(ET) 2.95E-05 4.74E-05 6.44E-05 8.12E-05 9.81E-05 1.15E-04 1.33E-04 1.50E-04 1.68E-04 1.86E-04 2.23E-04

BBsec 8.42E-05 1.36E-04 1.84E-04 2.32E-04 2.80E-04 3.29E-04 3.79E-04 4.30E-04 4.81E-04 5.33E-04 6.37E-04

BBbas 6.38E-05 1.03E-04 1.40E-04 1.76E-04 2.12E-04 2.49E-04 2.87E-04 3.26E-04 3.64E-04 4.03E-04 4.82E-04

bb 3.05E-05 4.91E-05 6.67E-05 8.41E-05 1.02E-04 1.19E-04 1.37E-04 1.56E-04 1.74E-04 1.93E-04 2.31E-04

AI 2.41E-05 3.88E-05 5.27E-05 6.64E-05 8.02E-05 9.42E-05 1.08E-04 1.23E-04 1.38E-04 1.52E-04 1.82E-04

LN(TH) 1.62E-05 2.61E-05 3.55E-05 4.47E-05 5.40E-05 6.35E-05 7.31E-05 8.28E-05 9.27E-05 1.03E-04 1.23E-04

Esophagus 5.51E-06 8.87E-06 1.21E-05 1.52E-05 1.84E-05 2.15E-05 2.48E-05 2.81E-05 3.15E-05 3.48E-05 4.17E-05

Gonads 3.99E-06 6.42E-06 8.72E-06 1.10E-05 1.33E-05 1.56E-05 1.80E-05 2.04E-05 2.28E-05 2.52E-05 3.02E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 209 of 248

Table D-13 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 4.92E-05 5.63E-05 6.33E-05 7.04E-05 8.80E-05 1.06E-04 1.23E-04 1.41E-04 1.58E-04 1.76E-04

Urinary Bladder 2.55E-05 2.92E-05 3.28E-05 3.65E-05 4.56E-05 5.47E-05 6.38E-05 7.29E-05 8.20E-05 9.12E-05

Brain 1.83E-05 2.09E-05 2.36E-05 2.62E-05 3.27E-05 3.93E-05 4.58E-05 5.24E-05 5.89E-05 6.55E-05

Breast 3.99E-05 4.56E-05 5.13E-05 5.70E-05 7.13E-05 8.56E-05 9.98E-05 1.14E-04 1.28E-04 1.43E-04

Gall Bladder 4.14E-05 4.73E-05 5.32E-05 5.91E-05 7.39E-05 8.87E-05 1.03E-04 1.18E-04 1.33E-04 1.48E-04

Heart Wall 6.18E-05 7.06E-05 7.95E-05 8.83E-05 1.10E-04 1.32E-04 1.55E-04 1.77E-04 1.99E-04 2.21E-04

Kidneys 3.34E-05 3.82E-05 4.30E-05 4.78E-05 5.97E-05 7.16E-05 8.36E-05 9.55E-05 1.07E-04 1.19E-04

Liver 6.88E-05 7.86E-05 8.84E-05 9.83E-05 1.23E-04 1.47E-04 1.72E-04 1.97E-04 2.21E-04 2.46E-04

Muscle 3.04E-05 3.48E-05 3.91E-05 4.34E-05 5.43E-05 6.52E-05 7.60E-05 8.69E-05 9.77E-05 1.09E-04

Ovaries 3.52E-05 4.02E-05 4.52E-05 5.03E-05 6.28E-05 7.54E-05 8.79E-05 1.01E-04 1.13E-04 1.26E-04

Pancreas 4.37E-05 5.00E-05 5.62E-056.24E-05 7.81E-05 9.37E-05 1.09E-04 1.25E-04 1.40E-04 1.56E-04

Testes 1.95E-05 2.23E-05 2.51E-05 2.79E-05 3.48E-05 4.18E-05 4.87E-05 5.57E-05 6.27E-05 6.96E-05

Thyroid 3.00E-05 3.43E-05 3.86E-05 4.28E-05 5.35E-05 6.43E-05 7.50E-05 8.57E-05 9.64E-05 1.07E-04

R.B.M. 3.42E-05 3.91E-05 4.40E-05 4.89E-05 6.12E-05 7.34E-05 8.56E-05 9.78E-05 1.10E-04 1.22E-04

Bone Surface 3.18E-05 3.64E-05 4.09E-05 4.55E-05 5.68E-05 6.82E-05 7.96E-05 9.10E-05 1.02E-04 1.14E-04

Stomach 3.64E-05 4.16E-05 4.68E-05 5.20E-05 6.50E-05 7.80E-05 9.10E-05 1.04E-04 1.17E-04 1.30E-04

S.I. 3.61E-05 4.12E-05 4.64E-05 5.15E-05 6.44E-05 7.73E-05 9.02E-05 1.03E-04 1.16E-04 1.29E-04

U.L.I. 4.71E-05 5.39E-05 6.06E-05 6.74E-05 8.42E-05 1.01E-04 1.18E-04 1.35E-04 1.52E-04 1.68E-04

L.L.I. 6.94E-05 7.93E-05 8.92E-05 9.91E-05 1.24E-04 1.49E-04 1.73E-04 1.98E-04 2.23E-04 2.48E-04

Skin 1.96E-05 2.24E-05 2.53E-05 2.81E-05 3.51E-05 4.21E-05 4.91E-05 5.61E-05 6.31E-05 7.02E-05

Spleen 3.89E-05 4.44E-05 5.00E-05 5.56E-05 6.94E-05 8.33E-05 9.72E-05 1.11E-04 1.25E-04 1.39E-04

Thymus 4.86E-05 5.56E-05 6.25E-05 6.95E-05 8.68E-05 1.04E-04 1.22E-04 1.39E-04 1.56E-04 1.74E-04

Uterus 2.93E-05 3.35E-05 3.77E-05 4.19E-05 5.24E-05 6.28E-05 7.33E-05 8.38E-05 9.42E-05 1.05E-04

Colon 5.67E-05 6.49E-05 7.30E-05 8.11E-05 1.01E-04 1.22E-04 1.42E-04 1.62E-04 1.82E-04 2.03E-04

ET 3.39E-04 3.88E-04 4.36E-04 4.85E-04 6.06E-04 7.27E-04 8.48E-04 9.69E-04 1.09E-03 1.21E-03

ET1 1.50E-02 1.72E-02 1.93E-02 2.14E-02 2.68E-02 3.22E-02 3.75E-02 4.29E-02 4.82E-02 5.36E-02

ET2 3.25E-04 3.71E-04 4.17E-04 4.64E-04 5.80E-04 6.96E-04 8.11E-04 9.27E-04 1.04E-03 1.16E-03

Lung 3.78E-04 4.32E-04 4.86E-04 5.40E-04 6.75E-04 8.10E-04 9.45E-04 1.08E-03 1.22E-03 1.35E-03

LN(ET) 2.60E-04 2.97E-04 3.34E-04 3.71E-04 4.64E-04 5.57E-04 6.50E-04 7.43E-04 8.35E-04 9.28E-04

BBsec 7.43E-04 8.49E-04 9.55E-04 1.06E-03 1.33E-03 1.59E-03 1.86E-03 2.12E-03 2.39E-03 2.65E-03

BBbas 5.63E-04 6.43E-04 7.24E-04 8.04E-04 1.01E-03 1.21E-03 1.41E-03 1.61E-03 1.81E-03 2.01E-03

bb 2.69E-04 3.08E-04 3.46E-04 3.85E-04 4.81E-04 5.77E-04 6.73E-04 7.69E-04 8.65E-04 9.62E-04

AI 2.12E-04 2.43E-04 2.73E-04 3.04E-04 3.79E-04 4.55E-04 5.31E-04 6.07E-04 6.83E-04 7.59E-04

LN(TH) 1.43E-04 1.64E-04 1.84E-04 2.05E-04 2.56E-04 3.07E-04 3.58E-04 4.09E-04 4.60E-04 5.11E-04

Esophagus 4.86E-05 5.56E-05 6.25E-05 6.95E-05 8.68E-05 1.04E-04 1.22E-04 1.39E-04 1.56E-04 1.74E-04

Gonads 3.52E-05 4.02E-05 4.52E-05 5.03E-05 6.28E-05 7.54E-05 8.79E-05 1.01E-04 1.13E-04 1.26E-04
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 210 of 248

Table D-14. Co-60, Type S, 50-year equivalent dose for years of chronic intake (Sv) MDA = 1 nCi.

Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 1.03E-05 1.36E-05 1.67E-05 2.00E-05 2.33E-05 2.68E-05 3.03E-05 3.39E-05 3.76E-05 4.13E-05 4.93E-05

Urinary Bladder 1.40E-06 1.84E-06 2.27E-06 2.71E-06 3.16E-06 3.62E-06 4.11E-06 4.60E-06 5.09E-06 5.59E-06 5.11E-06

Brain 1.55E-06 2.04E-06 2.51E-06 3.00E-06 3.50E-06 4.02E-06 4.55E-06 5.09E-06 5.64E-06 6.20E-06 6.35E-06

Breast 1.09E-05 1.44E-05 1.77E-05 2.11E-05 2.46E-05 2.83E-05 3.20E-05 3.58E-05 3.97E-05 4.36E-05 5.33E-05

Gall Bladder 4.83E-06 6.36E-06 7.83E-06 9.34E-06 1.09E-05 1.25E-05 1.42E-05 1.59E-05 1.76E-05 1.93E-05 2.15E-05

Heart Wall 1.77E-05 2.33E-05 2.87E-05 3.43E-05 4.00E-05 4.59E-05 5.20E-05 5.82E-05 6.45E-05 7.08E-05 8.68E-05

Kidneys 4.53E-06 5.97E-06 7.34E-06 8.76E-06 1.02E-05 1.17E-05 1.33E-05 1.49E-05 1.65E-05 1.81E-05 2.05E-05

Liver 1.01E-05 1.32E-05 1.63E-05 1.94E-05 2.27E-05 2.61E-05 2.95E-05 3.30E-05 3.66E-05 4.02E-05 4.59E-05

Muscle 4.98E-06 6.56E-06 8.07E-06 9.63E-06 1.12E-05 1.29E-05 1.46E-05 1.64E-05 1.81E-05 1.99E-05 2.32E-05

Ovaries 2.40E-06 3.17E-06 3.89E-06 4.65E-06 5.42E-06 6.22E-06 7.05E-06 7.89E-06 8.74E-06 9.60E-06 1.02E-05

Pancreas 7.82E-06 1.03E-05 1.27E-05 1.51E-05 1.76E-05 2.02E-05 2.29E-05 2.57E-05 2.84E-05 3.12E-05 3.67E-05

Testes 9.24E-07 1.22E-06 1.50E-06 1.79E-06 2.08E-06 2.39E-06 2.71E-06 3.03E-06 3.36E-06 3.69E-06 2.99E-06

Thyroid 4.96E-06 6.53E-06 8.03E-06 9.58E-06 1.12E-05 1.28E-05 1.45E-05 1.63E-05 1.80E-05 1.98E-05 2.30E-05

R.B.M. 6.05E-06 7.97E-06 9.81E-06 1.17E-05 1.37E-05 1.57E-05 1.78E-05 1.99E-05 2.20E-05 2.42E-05 2.85E-05

Bone Surface 4.79E-06 6.31E-06 7.76E-06 9.26E-06 1.08E-05 1.24E-05 1.41E-05 1.57E-05 1.74E-05 1.91E-05 2.20E-05

Stomach 6.21E-06 8.18E-06 1.01E-05 1.20E-05 1.40E-05 1.61E-05 1.82E-05 2.04E-05 2.26E-05 2.48E-05 2.92E-05

S.I. 2.91E-06 3.83E-06 4.71E-06 5.63E-06 6.57E-06 7.54E-06 8.54E-06 9.56E-06 1.06E-05 1.16E-05 1.27E-05

U.L.I. 3.98E-06 5.24E-06 6.44E-06 7.69E-06 8.97E-06 1.03E-05 1.17E-05 1.31E-05 1.45E-05 1.59E-05 1.83E-05

L.L.I. 5.34E-06 7.03E-06 8.65E-06 1.03E-05 1.20E-05 1.38E-05 1.57E-05 1.75E-05 1.94E-05 2.13E-05 2.57E-05

Skin 2.88E-06 3.80E-06 4.67E-06 5.57E-06 6.51E-06 7.47E-06 8.46E-06 9.47E-06 1.05E-05 1.15E-05 1.32E-05

Spleen 7.86E-06 1.03E-05 1.27E-05 1.52E-05 1.77E-05 2.03E-05 2.30E-05 2.58E-05 2.86E-05 3.14E-05 3.74E-05

Thymus 1.27E-05 1.67E-05 2.06E-05 2.45E-05 2.86E-05 3.29E-05 3.72E-05 4.17E-05 4.62E-05 5.07E-05 6.18E-05

Uterus 1.80E-06 2.37E-06 2.92E-06 3.49E-06 4.07E-06 4.67E-06 5.29E-06 5.92E-06 6.56E-06 7.21E-06 6.98E-06

Colon 4.57E-06 6.01E-06 7.39E-06 8.82E-06 1.03E-05 1.18E-05 1.34E-05 1.50E-05 1.66E-05 1.82E-05 2.15E-05

ET 6.84E-05 9.01E-05 1.11E-04 1.32E-04 1.54E-04 1.77E-04 2.01E-04 2.25E-04 2.49E-04 2.73E-04 3.41E-04

ET1 1.22E-03 1.61E-03 1.97E-03 2.36E-03 2.75E-03 3.16E-03 3.58E-03 4.00E-03 4.43E-03 4.87E-03 6.10E-03

ET2 6.73E-05 8.86E-05 1.09E-04 1.30E-04 1.52E-04 1.74E-04 1.97E-04 2.21E-04 2.45E-04 2.69E-04 3.35E-04

Lung 8.69E-05 1.14E-04 1.41E-04 1.68E-04 1.96E-04 2.25E-04 2.55E-04 2.85E-04 3.16E-04 3.47E-04 4.33E-04

LN(ET) 6.58E-05 8.66E-05 1.07E-04 1.27E-04 1.48E-04 1.70E-04 1.93E-04 2.16E-04 2.39E-04 2.63E-04 3.28E-04

BBsec 1.11E-04 1.46E-04 1.80E-04 2.15E-04 2.51E-04 2.88E-04 3.26E-04 3.65E-04 4.04E-04 4.44E-04 5.54E-04

BBbas 9.25E-05 1.22E-04 1.50E-04 1.79E-04 2.09E-04 2.40E-04 2.71E-04 3.04E-04 3.36E-04 3.70E-04 4.61E-04

bb 7.65E-05 1.01E-04 1.24E-04 1.48E-04 1.73E-04 1.98E-04 2.24E-04 2.51E-04 2.78E-04 3.06E-04 3.81E-04

AI 8.23E-05 1.08E-04 1.33E-04 1.59E-04 1.86E-04 2.13E-04 2.41E-04 2.70E-04 2.99E-04 3.29E-04 4.10E-04

LN(TH) 7.98E-05 1.05E-04 1.29E-04 1.54E-04 1.80E-04 2.07E-04 2.34E-04 2.62E-04 2.90E-04 3.19E-04 3.98E-04

Esophagus 1.27E-05 1.67E-05 2.06E-05 2.45E-05 2.86E-05 3.29E-05 3.72E-05 4.17E-05 4.62E-05 5.07E-05 6.18E-05

Gonads 2.40E-06 3.17E-06 3.89E-06 4.65E-06 5.42E-06 6.22E-06 7.05E-06 7.89E-06 8.74E-06 9.60E-06 1.02E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 211 of 248

Table D-14 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 5.70E-05 6.47E-05 7.24E-05 8.04E-05 1.00E-04 1.20E-04 1.40E-04 1.60E-04 1.80E-04 2.00E-04

Urinary Bladder 5.90E-06 6.70E-06 7.50E-06 8.32E-06 1.04E-05 1.24E-05 1.45E-05 1.65E-05 1.86E-05 2.07E-05

Brain 7.33E-06 8.33E-06 9.32E-06 1.03E-05 1.29E-05 1.54E-05 1.80E-05 2.05E-05 2.31E-05 2.57E-05

Breast 6.15E-05 6.99E-05 7.82E-05 8.68E-05 1.08E-04 1.30E-04 1.51E-04 1.72E-04 1.94E-04 2.15E-04

Gall Bladder 2.48E-05 2.82E-05 3.15E-05 3.50E-05 4.36E-05 5.22E-05 6.09E-05 6.95E-05 7.82E-05 8.69E-05

Heart Wall 1.00E-04 1.14E-04 1.27E-04 1.41E-04 1.76E-04 2.11E-04 2.46E-04 2.81E-04 3.16E-04 3.51E-04

Kidneys 2.37E-05 2.69E-05 3.02E-05 3.35E-05 4.17E-05 4.99E-05 5.83E-05 6.65E-05 7.48E-05 8.31E-05

Liver 5.31E-05 6.02E-05 6.74E-05 7.48E-05 9.32E-05 1.12E-04 1.30E-04 1.49E-04 1.67E-04 1.86E-04

Muscle 2.69E-05 3.05E-05 3.41E-05 3.79E-05 4.72E-05 5.65E-05 6.59E-05 7.52E-05 8.46E-05 9.40E-05

Ovaries 1.17E-05 1.33E-05 1.49E-05 1.66E-05 2.06E-05 2.47E-05 2.88E-05 3.29E-05 3.70E-05 4.11E-05

Pancreas 4.24E-05 4.82E-05 5.39E-05 5.98E-05 7.45E-05 8.93E-05 1.04E-04 1.19E-04 1.34E-04 1.49E-04

Testes 3.45E-06 3.92E-06 4.38E-06 4.86E-06 6.06E-06 7.26E-06 8.47E-06 9.66E-06 1.09E-05 1.21E-05

Thyroid 2.66E-05 3.02E-05 3.38E-05 3.75E-05 4.67E-05 5.60E-05 6.53E-05 7.45E-05 8.39E-05 9.32E-05

R.B.M. 3.30E-05 3.74E-05 4.19E-05 4.65E-05 5.79E-05 6.94E-05 8.09E-05 9.23E-05 1.04E-04 1.15E-04

Bone Surface 2.55E-05 2.89E-05 3.24E-05 3.59E-05 4.47E-05 5.36E-05 6.25E-05 7.13E-05 8.02E-05 8.92E-05

Stomach 3.38E-05 3.84E-05 4.29E-05 4.76E-05 5.93E-05 7.11E-05 8.29E-05 9.46E-05 1.06E-04 1.18E-04

S.I. 1.47E-05 1.66E-05 1.86E-05 2.07E-05 2.57E-05 3.08E-05 3.60E-05 4.10E-05 4.62E-05 5.13E-05

U.L.I. 2.12E-05 2.40E-05 2.69E-05 2.98E-05 3.72E-05 4.45E-05 5.19E-05 5.93E-05 6.67E-05 7.41E-05

L.L.I. 2.97E-05 3.37E-05 3.77E-05 4.19E-05 5.22E-05 6.25E-05 7.29E-05 8.32E-05 9.36E-05 1.04E-04

Skin 1.53E-05 1.73E-05 1.94E-05 2.15E-05 2.68E-05 3.21E-05 3.75E-05 4.28E-05 4.81E-05 5.35E-05

Spleen 4.32E-05 4.91E-05 5.49E-05 6.09E-05 7.59E-05 9.09E-05 1.06E-04 1.21E-04 1.36E-04 1.51E-04

Thymus 7.13E-05 8.10E-05 9.07E-05 1.01E-04 1.25E-04 1.50E-04 1.75E-04 2.00E-04 2.25E-04 2.50E-04

Uterus 8.06E-06 9.15E-06 1.02E-05 1.14E-05 1.42E-05 1.70E-05 1.98E-05 2.26E-05 2.54E-05 2.82E-05

Colon 2.48E-05 2.82E-05 3.16E-05 3.50E-05 4.36E-05 5.23E-05 6.10E-05 6.96E-05 7.83E-05 8.70E-05

ET 3.94E-04 4.47E-04 5.00E-04 5.55E-04 6.91E-04 8.28E-04 9.67E-04 1.10E-03 1.24E-03 1.38E-03

ET1 7.05E-03 8.00E-03 8.96E-03 9.94E-03 1.24E-02 1.48E-02 1.73E-02 1.97E-02 2.22E-02 2.47E-02

ET2 3.87E-04 4.39E-04 4.92E-04 5.46E-04 6.80E-04 8.14E-04 9.50E-04 1.08E-03 1.22E-03 1.36E-03

Lung 5.00E-04 5.68E-04 6.36E-04 7.05E-04 8.79E-04 1.05E-03 1.23E-03 1.40E-03 1.58E-03 1.75E-03

LN(ET) 3.78E-04 4.30E-04 4.81E-04 5.33E-04 6.65E-04 7.96E-04 9.29E-04 1.06E-03 1.19E-03 1.32E-03

BBsec 6.40E-04 7.27E-04 8.14E-04 9.02E-04 1.12E-03 1.35E-03 1.57E-03 1.79E-03 2.02E-03 2.24E-03

BBbas 5.33E-04 6.05E-04 6.77E-04 7.51E-04 9.36E-04 1.12E-03 1.31E-03 1.49E-03 1.68E-03 1.87E-03

bb 4.40E-04 5.00E-04 5.60E-04 6.21E-04 7.74E-04 9.27E-04 1.08E-03 1.23E-03 1.39E-03 1.54E-03

AI 4.74E-04 5.38E-04 6.02E-04 6.68E-04 8.32E-04 9.97E-04 1.16E-03 1.33E-03 1.49E-03 1.66E-03

LN(TH) 4.59E-04 5.22E-04 5.84E-04 6.48E-04 8.07E-04 9.67E-04 1.13E-03 1.29E-03 1.45E-03 1.61E-03

Esophagus 7.13E-05 8.10E-05 9.07E-05 1.01E-04 1.25E-04 1.50E-04 1.75E-04 2.00E-04 2.25E-04 2.50E-04

Gonads 1.17E-05 1.33E-05 1.49E-05 1.66E-05 2.06E-05 2.47E-05 2.88E-05 3.29E-05 3.70E-05 4.11E-05
Effective Date: 0

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Table D-15. Cs-137, Type F 50-year equivalent dose for years of chronic intake (Sv) (MDA = 1 nCi). Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 1.47E-06 2.68E-06 3.98E-06 5.31E-06 6.63E-06 7.95E-06 9.28E-06 1.06E-05 1.19E-05 1.33E-05 1.59E-05

Urinary Bladder 1.51E-06 2.76E-06 4.11E-06 5.47E-06 6.84E-06 8.20E-06 9.56E-06 1.09E-05 1.23E-05 1.37E-05 1.64E-05

Brain 1.24E-06 2.26E-06 3.35E-06 4.47E-06 5.59E-06 6.69E-06 7.81E-06 8.93E-06 1.00E-05 1.12E-05 1.34E-05

Breast 1.18E-06 2.15E-06 3.19E-06 4.26E-06 5.32E-06 6.37E-06 7.44E-06 8.50E-06 9.56E-06 1.06E-05 1.28E-05

Gall Bladder 1.45E-06 2.65E-06 3.93E-06 5.24E-06 6.55E-06 7.85E-06 9.16E-06 1.05E-05 1.18E-05 1.31E-05 1.57E-05

Heart Wall 1.45E-06 2.64E-06 3.93E-06 5.24E-06 6.55E-06 7.85E-06 9.16E-06 1.05E-05 1.18E-05 1.31E-05 1.57E-05

Kidneys 1.41E-06 2.57E-06 3.83E-06 5.10E-06 6.38E-06 7.64E-06 8.92E-06 1.02E-05 1.15E-05 1.27E-05 1.53E-05

Liver 1.42E-06 2.59E-06 3.85E-06 5.14E-06 6.42E-06 7.69E-06 8.97E-06 1.03E-05 1.15E-05 1.28E-05 1.54E-05

Muscle 1.32E-06 2.40E-06 3.57E-06 4.76E-06 5.95E-06 7.13E-06 8.32E-06 9.51E-06 1.07E-05 1.19E-05 1.43E-05

Ovaries 1.50E-06 2.73E-06 4.06E-06 5.41E-06 6.76E-06 8.10E-06 9.45E-06 1.08E-05 1.22E-05 1.35E-05 1.62E-05

Pancreas 1.51E-06 2.75E-06 4.08E-06 5.44E-06 6.80E-06 8.15E-06 9.51E-06 1.09E-05 1.22E-05 1.36E-05 1.63E-05

Testes 1.32E-06 2.40E-06 3.57E-06 4.76E-06 5.95E-06 7.13E-06 8.32E-06 9.51E-06 1.07E-05 1.19E-05 1.43E-05

Thyroid 1.37E-06 2.50E-06 3.72E-06 4.96E-06 6.20E-06 7.42E-06 8.67E-06 9.91E-06 1.11E-05 1.24E-05 1.49E-05

R.B.M. 1.37E-06 2.50E-06 3.72E-06 4.96E-06 6.20E-06 7.42E-06 8.67E-06 9.91E-06 1.11E-05 1.24E-05 1.49E-05

Bone Surface 1.44E-06 2.62E-06 3.90E-06 5.20E-06 6.50E-06 7.78E-06 9.08E-06 1.04E-05 1.17E-05 1.30E-05 1.56E-05

Stomach 1.38E-06 2.52E-06 3.75E-06 4.99E-06 6.24E-06 7.48E-06 8.73E-06 9.97E-06 1.12E-05 1.25E-05 1.50E-05

S.I. 1.47E-06 2.67E-06 3.98E-06 5.30E-06 6.62E-06 7.93E-06 9.26E-06 1.06E-05 1.19E-05 1.32E-05 1.59E-05

U.L.I. 1.51E-06 2.74E-06 4.08E-06 5.44E-06 6.80E-06 8.14E-06 9.51E-06 1.09E-05 1.22E-05 1.36E-05 1.63E-05

L.L.I. 1.73E-06 3.16E-06 4.70E-06 6.27E-06 7.83E-06 9.38E-06 1.10E-05 1.25E-05 1.41E-05 1.56E-05 1.88E-05

Skin 1.13E-06 2.05E-06 3.05E-06 4.07E-06 5.09E-06 6.10E-06 7.11E-06 8.13E-06 9.15E-06 1.02E-05 1.22E-05

Spleen 1.41E-06 2.57E-06 3.83E-06 5.10E-06 6.38E-06 7.64E-06 8.92E-06 1.02E-05 1.15E-05 1.27E-05 1.53E-05

Thymus 1.38E-06 2.51E-06 3.73E-06 4.97E-06 6.22E-06 7.45E-06 8.69E-06 9.94E-06 1.12E-05 1.24E-05 1.49E-05

Uterus 1.51E-06 2.75E-06 4.09E-06 5.46E-06 6.82E-06 8.17E-06 9.54E-06 1.09E-05 1.23E-05 1.36E-05 1.64E-05

Colon 1.60E-06 2.92E-06 4.35E-06 5.80E-06 7.25E-06 8.68E-06 1.01E-05 1.16E-05 1.30E-05 1.45E-05 1.74E-05

ET 2.82E-06 5.15E-06 7.65E-06 1.02E-05 1.27E-05 1.53E-05 1.78E-05 2.04E-05 2.29E-05 2.55E-05 3.06E-05

ET1 4.16E-04 7.58E-04 1.13E-03 1.50E-03 1.88E-03 2.25E-03 2.62E-03 3.00E-03 3.38E-03 3.75E-03 4.50E-03

ET2 2.41E-06 4.39E-06 6.53E-06 8.71E-06 1.09E-05 1.30E-05 1.52E-05 1.74E-05 1.96E-05 2.17E-05 2.61E-05

Lung 1.33E-06 2.43E-06 3.62E-06 4.82E-06 6.03E-06 7.22E-06 8.43E-06 9.64E-06 1.08E-05 1.20E-05 1.45E-05

LN(ET) 2.33E-06 4.26E-06 6.33E-06 8.44E-06 1.05E-05 1.26E-05 1.47E-05 1.69E-05 1.90E-05 2.11E-05 2.53E-05

BBsec 1.35E-06 2.45E-06 3.65E-06 4.86E-06 6.08E-06 7.28E-06 8.49E-06 9.71E-06 1.09E-05 1.21E-05 1.46E-05

BBbas 1.34E-06 2.45E-06 3.64E-06 4.85E-06 6.06E-06 7.26E-06 8.47E-06 9.69E-06 1.09E-05 1.21E-05 1.45E-05

bb 1.33E-06 2.43E-06 3.61E-06 4.81E-06 6.01E-06 7.20E-06 8.41E-06 9.61E-06 1.08E-05 1.20E-05 1.44E-05

AI 1.33E-06 2.42E-06 3.60E-06 4.81E-06 6.01E-06 7.19E-06 8.40E-06 9.60E-06 1.08E-05 1.20E-05 1.44E-05

LN(TH) 1.33E-06 2.42E-06 3.60E-06 4.80E-06 6.00E-06 7.19E-06 8.39E-06 9.60E-06 1.08E-05 1.20E-05 1.44E-05

Esophagus 1.38E-06 2.51E-06 3.73E-06 4.97E-06 6.22E-06 7.45E-06 8.69E-06 9.94E-06 1.12E-05 1.24E-05 1.49E-05

Gonads 1.50E-06 2.73E-06 4.06E-06 5.41E-06 6.76E-06 8.10E-06 9.45E-06 1.08E-05 1.22E-05 1.35E-05 1.62E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 213 of 248

Table D-15 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 1.86E-05 2.12E-05 2.39E-05 2.65E-05 3.31E-05 3.98E-05 4.64E-05 5.30E-05 5.97E-05 6.63E-05

Urinary Bladder 1.91E-05 2.19E-05 2.46E-05 2.73E-05 3.42E-05 4.10E-05 4.78E-05 5.47E-05 6.15E-05 6.83E-05

Brain 1.56E-05 1.79E-05 2.01E-05 2.23E-05 2.79E-05 3.35E-05 3.91E-05 4.47E-05 5.02E-05 5.58E-05

Breast 1.49E-05 1.70E-05 1.91E-05 2.13E-05 2.66E-05 3.19E-05 3.72E-05 4.25E-05 4.78E-05 5.31E-05

Gall Bladder 1.83E-05 2.10E-05 2.36E-05 2.62E-05 3.27E-05 3.93E-05 4.58E-05 5.24E-05 5.89E-05 6.55E-05

Heart Wall 1.83E-05 2.09E-05 2.35E-05 2.62E-05 3.27E-05 3.93E-05 4.58E-05 5.23E-05 5.89E-05 6.54E-05

Kidneys 1.78E-05 2.04E-05 2.29E-05 2.55E-05 3.19E-05 3.82E-05 4.46E-05 5.10E-05 5.73E-05 6.37E-05

Liver 1.80E-05 2.05E-05 2.31E-05 2.57E-05 3.21E-05 3.85E-05 4.49E-05 5.13E-05 5.77E-05 6.41E-05

Muscle 1.66E-05 1.90E-05 2.14E-05 2.38E-05 2.97E-05 3.57E-05 4.16E-05 4.76E-05 5.35E-05 5.95E-05

Ovaries 1.89E-05 2.16E-05 2.43E-05 2.70E-05 3.38E-05 4.05E-05 4.73E-05 5.40E-05 6.08E-05 6.75E-05

Pancreas 1.90E-05 2.17E-05 2.45E-05 2.72E-05 3.40E-05 4.08E-05 4.76E-05 5.44E-05 6.11E-05 6.79E-05

Testes 1.66E-05 1.90E-05 2.14E-05 2.38E-05 2.97E-05 3.57E-05 4.16E-05 4.76E-05 5.35E-05 5.94E-05

Thyroid 1.73E-05 1.98E-05 2.23E-05 2.48E-05 3.10E-05 3.72E-05 4.33E-05 4.95E-05 5.57E-05 6.19E-05

R.B.M. 1.73E-05 1.98E-05 2.23E-05 2.48E-05 3.10E-05 3.72E-05 4.33E-05 4.95E-05 5.57E-05 6.19E-05

Bone Surface 1.82E-05 2.08E-05 2.34E-05 2.60E-05 3.24E-05 3.89E-05 4.54E-05 5.19E-05 5.84E-05 6.49E-05

Stomach 1.75E-05 2.00E-05 2.24E-05 2.49E-05 3.12E-05 3.74E-05 4.36E-05 4.99E-05 5.61E-05 6.23E-05

S.I. 1.85E-05 2.12E-05 2.38E-05 2.65E-05 3.31E-05 3.97E-05 4.63E-05 5.29E-05 5.95E-05 6.62E-05

U.L.I. 1.90E-05 2.17E-05 2.44E-05 2.72E-05 3.40E-05 4.08E-05 4.75E-05 5.43E-05 6.11E-05 6.79E-05

L.L.I. 2.19E-05 2.50E-05 2.82E-05 3.13E-05 3.91E-05 4.70E-05 5.48E-05 6.26E-05 7.04E-05 7.83E-05

Skin 1.42E-05 1.63E-05 1.83E-05 2.03E-05 2.54E-05 3.05E-05 3.56E-05 4.07E-05 4.57E-05 5.08E-05

Spleen 1.78E-05 2.04E-05 2.29E-05 2.55E-05 3.19E-05 3.82E-05 4.46E-05 5.10E-05 5.73E-05 6.37E-05

Thymus 1.74E-05 1.99E-05 2.24E-05 2.48E-05 3.11E-05 3.73E-05 4.35E-05 4.97E-05 5.59E-05 6.21E-05

Uterus 1.91E-05 2.18E-05 2.45E-05 2.73E-05 3.41E-05 4.09E-05 4.77E-05 5.45E-05 6.13E-05 6.82E-05

Colon 2.03E-05 2.32E-05 2.61E-05 2.90E-05 3.62E-05 4.34E-05 5.07E-05 5.79E-05 6.51E-05 7.24E-05

ET 3.57E-05 4.08E-05 4.58E-05 5.09E-05 6.37E-05 7.64E-05 8.91E-05 1.02E-04 1.15E-04 1.27E-04

ET1 5.25E-03 6.00E-03 6.75E-03 7.50E-03 9.38E-03 1.13E-02 1.31E-02 1.50E-02 1.69E-02 1.88E-02

ET2 3.04E-05 3.48E-05 3.91E-05 4.35E-05 5.44E-05 6.52E-05 7.61E-05 8.70E-05 9.78E-05 1.09E-04

Lung 1.69E-05 1.93E-05 2.17E-05 2.41E-05 3.01E-05 3.61E-05 4.22E-05 4.82E-05 5.42E-05 6.02E-05

LN(ET) 2.95E-05 3.37E-05 3.79E-05 4.21E-05 5.27E-05 6.32E-05 7.37E-05 8.43E-05 9.48E-05 1.05E-04

BBsec 1.70E-05 1.94E-05 2.18E-05 2.43E-05 3.03E-05 3.64E-05 4.25E-05 4.86E-05 5.46E-05 6.07E-05

BBbas 1.70E-05 1.94E-05 2.18E-05 2.42E-05 3.03E-05 3.63E-05 4.24E-05 4.84E-05 5.45E-05 6.06E-05

bb 1.68E-05 1.92E-05 2.16E-05 2.40E-05 3.00E-05 3.60E-05 4.20E-05 4.80E-05 5.41E-05 6.01E-05

AI 1.68E-05 1.92E-05 2.16E-05 2.40E-05 3.00E-05 3.60E-05 4.20E-05 4.80E-05 5.40E-05 6.00E-05

LN(TH) 1.68E-05 1.92E-05 2.16E-05 2.40E-05 3.00E-05 3.60E-05 4.20E-05 4.80E-05 5.40E-05 6.00E-05

Esophagus 1.74E-05 1.99E-05 2.24E-05 2.48E-05 3.11E-05 3.73E-05 4.35E-05 4.97E-05 5.59E-05 6.21E-05

Gonads 1.89E-05 2.16E-05 2.43E-05 2.70E-05 3.38E-05 4.05E-05 4.73E-05 5.40E-05 6.08E-05 6.75E-05
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 214 of 248

Table D-16. Eu-154, Type M, 50-year equivalent dose for years of chronic intake (Sv) MDA = 2.4 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 4.35E-05 5.04E-05 5.57E-05 6.05E-05 6.52E-05 7.00E-05 7.49E-05 7.98E-05 8.50E-05 9.02E-05 1.01E-04

Urinary Bladder 6.28E-06 7.26E-06 8.03E-06 8.72E-06 9.41E-06 1.01E-05 1.08E-05 1.15E-05 1.23E-05 1.30E-05 1.46E-05

Brain 8.40E-06 9.72E-06 1.07E-05 1.17E-05 1.26E-05 1.35E-05 1.45E-05 1.54E-05 1.64E-05 1.74E-05 1.95E-05

Breast 1.14E-05 1.32E-05 1.46E-05 1.59E-05 1.71E-05 1.84E-05 1.97E-05 2.10E-05 2.23E-05 2.37E-05 2.65E-05

Gall Bladder 6.11E-05 7.07E-05 7.81E-05 8.49E-05 9.15E-05 9.82E-05 1.05E-04 1.12E-04 1.19E-04 1.27E-04 1.42E-04

Heart Wall 2.46E-05 2.85E-05 3.15E-05 3.42E-05 3.69E-05 3.96E-05 4.24E-05 4.51E-05 4.81E-05 5.10E-05 5.71E-05

Kidneys 2.99E-05 3.46E-05 3.82E-05 4.15E-05 4.48E-05 4.81E-05 5.15E-05 5.48E-05 5.84E-05 6.19E-05 6.94E-05

Liver 3.51E-04 4.06E-04 4.48E-04 4.87E-04 5.25E-04 5.64E-04 6.04E-04 6.43E-04 6.85E-04 7.26E-04 8.14E-04

Muscle 1.27E-05 1.47E-05 1.62E-05 1.76E-05 1.90E-05 2.04E-05 2.18E-05 2.33E-05 2.48E-05 2.63E-05 2.95E-05

Ovaries 1.36E-05 1.57E-05 1.74E-05 1.89E-05 2.03E-05 2.18E-05 2.34E-05 2.49E-05 2.65E-05 2.81E-05 3.15E-05

Pancreas 3.21E-05 3.72E-05 4.11E-05 4.47E-05 4.82E-05 5.17E-05 5.53E-05 5.90E-05 6.28E-05 6.66E-05 7.46E-05

Testes 3.49E-06 4.04E-06 4.46E-06 4.85E-06 5.23E-06 5.61E-06 6.01E-06 6.40E-06 6.82E-06 7.23E-06 8.10E-06

Thyroid 7.62E-06 8.81E-06 9.74E-06 1.06E-05 1.14E-05 1.22E-05 1.31E-05 1.40E-05 1.49E-05 1.58E-05 1.77E-05

R.B.M. 9.52E-05 1.10E-04 1.22E-04 1.32E-04 1.43E-04 1.53E-04 1.64E-04 1.75E-04 1.86E-04 1.97E-04 2.21E-04

Bone Surface 4.28E-04 4.95E-04 5.48E-04 5.95E-04 6.42E-04 6.88E-04 7.37E-04 7.85E-04 8.37E-04 8.87E-04 9.94E-04

Stomach 1.59E-05 1.84E-05 2.04E-05 2.21E-05 2.39E-05 2.56E-05 2.74E-05 2.92E-05 3.11E-05 3.30E-05 3.70E-05

S.I. 1.78E-05 2.06E-05 2.28E-05 2.48E-05 2.67E-05 2.86E-05 3.07E-05 3.27E-05 3.48E-05 3.69E-05 4.14E-05

U.L.I. 2.52E-05 2.92E-05 3.23E-05 3.50E-05 3.78E-05 4.06E-05 4.34E-05 4.63E-05 4.93E-05 5.23E-05 5.86E-05

L.L.I. 2.38E-05 2.76E-05 3.05E-05 3.31E-05 3.57E-05 3.83E-05 4.10E-05 4.37E-05 4.66E-05 4.94E-05 5.53E-05

Skin 8.09E-06 9.36E-06 1.03E-05 1.12E-05 1.21E-05 1.30E-05 1.39E-05 1.48E-05 1.58E-05 1.68E-05 1.88E-05

Spleen 1.21E-05 1.40E-05 1.55E-05 1.68E-05 1.81E-05 1.95E-05 2.08E-05 2.22E-05 2.36E-05 2.51E-05 2.81E-05

Thymus 1.21E-05 1.40E-05 1.55E-05 1.69E-05 1.82E-05 1.95E-05 2.09E-05 2.23E-05 2.37E-05 2.52E-05 2.82E-05

Uterus 9.75E-06 1.13E-05 1.25E-05 1.35E-05 1.46E-05 1.57E-05 1.68E-05 1.79E-05 1.90E-05 2.02E-05 2.26E-05

Colon 2.46E-05 2.85E-05 3.15E-05 3.42E-05 3.69E-05 3.96E-05 4.24E-05 4.52E-05 4.81E-05 5.10E-05 5.72E-05

ET 4.60E-05 5.32E-05 5.88E-05 6.39E-05 6.89E-05 7.39E-05 7.92E-05 8.43E-05 8.98E-05 9.52E-05 1.07E-04

ET1 3.85E-03 4.46E-03 4.92E-03 5.35E-03 5.77E-03 6.19E-03 6.63E-03 7.06E-03 7.52E-03 7.98E-03 8.94E-03

ET2 4.22E-05 4.88E-05 5.39E-05 5.86E-05 6.32E-05 6.78E-05 7.26E-05 7.74E-05 8.24E-05 8.74E-05 9.79E-05

Lung 1.04E-04 1.20E-04 1.33E-04 1.44E-04 1.56E-04 1.67E-04 1.79E-04 1.91E-04 2.03E-04 2.15E-04 2.41E-04

LN(ET) 2.65E-05 3.06E-05 3.38E-05 3.68E-05 3.97E-05 4.25E-05 4.56E-05 4.85E-05 5.17E-05 5.48E-05 6.14E-05

BBsec 1.93E-04 2.24E-04 2.47E-04 2.69E-04 2.90E-04 3.11E-04 3.33E-04 3.55E-04 3.78E-04 4.01E-04 4.49E-04

BBbas 1.49E-04 1.73E-04 1.91E-04 2.07E-04 2.24E-04 2.40E-04 2.57E-04 2.74E-04 2.92E-04 3.09E-04 3.47E-04

bb 7.67E-05 8.88E-05 9.81E-05 1.07E-04 1.15E-04 1.23E-04 1.32E-04 1.41E-04 1.50E-04 1.59E-04 1.78E-04

AI 6.37E-05 7.37E-05 8.15E-05 8.86E-05 9.55E-05 1.02E-04 1.10E-04 1.17E-04 1.25E-04 1.32E-04 1.48E-04

LN(TH) 3.38E-05 3.91E-05 4.32E-05 4.69E-05 5.06E-05 5.43E-05 5.82E-05 6.20E-05 6.60E-05 7.00E-05 7.84E-05

Esophagus 1.21E-05 1.40E-05 1.55E-05 1.69E-05 1.82E-05 1.95E-05 2.09E-05 2.23E-05 2.37E-05 2.52E-05 2.82E-05

Gonads 1.36E-05 1.57E-05 1.74E-05 1.89E-05 2.03E-05 2.18E-05 2.34E-05 2.49E-05 2.65E-05 2.81E-05 3.15E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 215 of 248

Table D-16 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 1.12E-04 1.24E-04 1.36E-04 1.49E-04 1.82E-04 2.15E-04 2.49E-04 2.84E-04 3.19E-04 3.55E-04

Urinary Bladder 1.62E-05 1.79E-05 1.97E-05 2.15E-05 2.62E-05 3.10E-05 3.60E-05 4.10E-05 4.60E-05 5.12E-05

Brain 2.17E-05 2.40E-05 2.63E-05 2.87E-05 3.50E-05 4.15E-05 4.81E-05 5.49E-05 6.16E-05 6.85E-05

Breast 2.95E-05 3.26E-05 3.58E-05 3.91E-05 4.77E-05 5.64E-05 6.55E-05 7.47E-05 8.38E-05 9.32E-05

Gall Bladder 1.58E-04 1.74E-04 1.91E-04 2.09E-04 2.55E-04 3.02E-04 3.50E-04 3.99E-04 4.48E-04 4.98E-04

Heart Wall 6.35E-05 7.02E-05 7.71E-05 8.42E-05 1.03E-04 1.22E-04 1.41E-04 1.61E-04 1.81E-04 2.01E-04

Kidneys 7.71E-05 8.53E-05 9.36E-05 1.02E-04 1.25E-04 1.48E-04 1.71E-04 1.95E-04 2.19E-04 2.44E-04

Liver 9.04E-04 1.00E-03 1.10E-03 1.20E-03 1.46E-03 1.73E-03 2.01E-03 2.29E-03 2.57E-03 2.86E-03

Muscle 3.27E-05 3.62E-05 3.97E-05 4.34E-05 5.29E-05 6.26E-05 7.27E-05 8.29E-05 9.30E-05 1.03E-04

Ovaries 3.50E-05 3.87E-05 4.25E-05 4.64E-05 5.66E-05 6.70E-05 7.78E-05 8.87E-05 9.96E-05 1.11E-04

Pancreas 8.29E-05 9.17E-05 1.01E-04 1.10E-04 1.34E-04 1.59E-04 1.84E-04 2.10E-04 2.36E-04 2.62E-04

Testes 9.01E-06 9.95E-06 1.09E-05 1.19E-05 1.46E-05 1.72E-05 2.00E-05 2.28E-05 2.56E-05 2.84E-05

Thyroid 1.97E-05 2.17E-05 2.39E-05 2.61E-05 3.18E-05 3.76E-05 4.37E-05 4.98E-05 5.59E-05 6.21E-05

R.B.M. 2.46E-04 2.71E-04 2.98E-04 3.25E-04 3.97E-04 4.70E-04 5.45E-04 6.22E-04 6.98E-04 7.76E-04

Bone Surface 1.10E-03 1.22E-03 1.34E-03 1.46E-03 1.79E-03 2.11E-03 2.45E-03 2.80E-03 3.14E-03 3.49E-03

Stomach 4.11E-05 4.54E-05 4.99E-05 5.45E-05 6.64E-05 7.86E-05 9.13E-05 1.04E-04 1.17E-04 1.30E-04

S.I. 4.60E-05 5.08E-05 5.58E-05 6.09E-05 7.43E-05 8.80E-05 1.02E-04 1.16E-04 1.31E-04 1.45E-04

U.L.I. 6.51E-05 7.19E-05 7.90E-05 8.63E-05 1.05E-04 1.25E-04 1.45E-04 1.65E-04 1.85E-04 2.06E-04

L.L.I. 6.15E-05 6.80E-05 7.47E-05 8.15E-05 9.94E-05 1.18E-04 1.37E-04 1.56E-04 1.75E-04 1.94E-04

Skin 2.09E-05 2.31E-05 2.54E-05 2.77E-05 3.37E-05 3.99E-05 4.64E-05 5.29E-05 5.94E-05 6.60E-05

Spleen 3.12E-05 3.45E-05 3.79E-05 4.14E-05 5.05E-05 5.97E-05 6.94E-05 7.91E-05 8.87E-05 9.86E-05

Thymus 3.13E-05 3.46E-05 3.80E-05 4.15E-05 5.06E-05 5.99E-05 6.96E-05 7.94E-05 8.91E-05 9.90E-05

Uterus 2.52E-05 2.78E-05 3.05E-05 3.33E-05 4.07E-05 4.81E-05 5.59E-05 6.37E-05 7.15E-05 7.95E-05

Colon 6.35E-05 7.02E-05 7.71E-05 8.42E-05 1.03E-04 1.22E-04 1.41E-04 1.61E-04 1.81E-04 2.01E-04

ET 1.19E-04 1.31E-04 1.44E-04 1.57E-04 1.92E-04 2.27E-04 2.64E-04 3.00E-04 3.37E-04 3.75E-04

ET1 9.94E-03 1.10E-02 1.21E-02 1.32E-02 1.61E-02 1.90E-02 2.21E-02 2.52E-02 2.82E-02 3.14E-02

ET2 1.09E-04 1.20E-04 1.32E-04 1.44E-04 1.76E-04 2.08E-04 2.42E-04 2.76E-04 3.09E-043.44E-04

Lung 2.68E-04 2.96E-04 3.25E-04 3.55E-04 4.33E-04 5.13E-04 5.95E-04 6.79E-04 7.62E-04 8.47E-04

LN(ET) 6.83E-05 7.55E-05 8.29E-05 9.05E-05 1.10E-04 1.31E-04 1.52E-04 1.73E-04 1.94E-04 2.16E-04

BBsec 4.99E-04 5.52E-04 6.06E-04 6.61E-04 8.07E-04 9.55E-04 1.11E-03 1.26E-03 1.42E-03 1.58E-03

BBbas 3.85E-04 4.26E-04 4.68E-04 5.11E-04 6.23E-04 7.37E-04 8.56E-04 9.76E-04 1.09E-03 1.22E-03

bb 1.98E-04 2.19E-04 2.40E-04 2.62E-04 3.20E-04 3.79E-04 4.40E-04 5.01E-04 5.63E-04 6.25E-04

AI 1.64E-04 1.82E-04 2.00E-04 2.18E-04 2.66E-04 3.15E-04 3.65E-04 4.17E-04 4.68E-04 5.20E-04

LN(TH) 8.72E-05 9.63E-05 1.06E-04 1.16E-04 1.41E-04 1.67E-04 1.94E-04 2.21E-04 2.48E-04 2.75E-04

Esophagus 3.13E-05 3.46E-05 3.80E-05 4.15E-05 5.06E-05 5.99E-05 6.96E-05 7.94E-05 8.91E-05 9.90E-05

Gonads 3.50E-05 3.87E-05 4.25E-05 4.64E-05 5.66E-05 6.70E-05 7.78E-05 8.87E-05 9.96E-05 1.11E-04
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 216 of 248

Table D-17. I-131, Type F, 50-year equivalent dose for years of chronic intake (Sv) MDA = 1.4 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 2.23E-07 4.46E-07 6.69E-07 8.93E-07 1.12E-06 1.34E-06 1.56E-06 1.79E-06 2.01E-06 2.23E-06 2.68E-06

Urinary Bladder 3.31E-06 6.62E-06 9.94E-06 1.33E-05 1.66E-05 1.99E-05 2.32E-05 2.65E-05 2.98E-05 3.31E-05 3.98E-05

Brain 7.36E-07 1.47E-06 2.21E-06 2.95E-06 3.68E-06 4.42E-06 5.16E-06 5.89E-06 6.63E-06 7.37E-06 8.84E-06

Breast 2.86E-07 5.72E-07 8.59E-07 1.15E-06 1.43E-06 1.72E-06 2.00E-06 2.29E-06 2.58E-06 2.86E-06 3.44E-06

Gall Bladder 2.00E-07 4.01E-07 6.01E-07 8.02E-07 1.00E-06 1.20E-06 1.40E-06 1.60E-06 1.80E-06 2.00E-06 2.41E-06

Heart Wall 3.60E-07 7.19E-07 1.08E-06 1.44E-06 1.80E-06 2.16E-06 2.52E-06 2.88E-06 3.24E-06 3.60E-06 4.32E-06

Kidneys 1.99E-07 3.99E-07 5.98E-07 7.98E-07 9.97E-07 1.20E-06 1.40E-06 1.60E-06 1.80E-06 1.99E-06 2.39E-06

Liver 2.19E-07 4.38E-07 6.57E-07 8.77E-07 1.10E-06 1.31E-06 1.53E-06 1.75E-06 1.97E-06 2.19E-06 2.63E-06

Muscle 6.32E-07 1.26E-06 1.90E-06 2.53E-06 3.16E-06 3.79E-06 4.43E-06 5.06E-06 5.69E-06 6.33E-06 7.59E-06

Ovaries 2.09E-07 4.19E-07 6.28E-07 8.38E-07 1.05E-06 1.26E-06 1.47E-06 1.68E-06 1.89E-06 2.10E-06 2.51E-06

Pancreas 2.44E-07 4.87E-07 7.31E-07 9.75E-07 1.22E-06 1.46E-06 1.71E-06 1.95E-06 2.19E-06 2.44E-06 2.93E-06

Testes 1.75E-07 3.50E-07 5.25E-07 7.01E-07 8.76E-07 1.05E-06 1.23E-06 1.40E-06 1.58E-06 1.75E-06 2.10E-06

Thyroid 1.90E-03 3.79E-03 5.69E-03 7.59E-03 9.48E-03 1.14E-02 1.33E-02 1.52E-02 1.71E-02 1.90E-02 2.28E-02

R.B.M. 4.97E-07 9.94E-07 1.49E-06 1.99E-06 2.49E-06 2.98E-06 3.48E-06 3.98E-06 4.47E-06 4.97E-06 5.97E-06

Bone Surface 6.25E-07 1.25E-06 1.88E-06 2.50E-06 3.13E-06 3.75E-06 4.38E-06 5.00E-06 5.63E-06 6.25E-06 7.51E-06

Stomach 6.62E-07 1.32E-06 1.99E-06 2.65E-06 3.31E-06 3.97E-06 4.64E-06 5.30E-06 5.96E-06 6.63E-06 7.95E-06

S.I. 1.92E-07 3.84E-07 5.76E-07 7.68E-07 9.60E-07 1.15E-06 1.34E-06 1.54E-06 1.73E-06 1.92E-06 2.30E-06

U.L.I. 2.20E-07 4.40E-07 6.60E-07 8.81E-07 1.10E-06 1.32E-06 1.54E-06 1.76E-06 1.98E-06 2.20E-06 2.64E-06

L.L.I. 2.92E-07 5.84E-07 8.76E-07 1.17E-06 1.46E-06 1.75E-06 2.04E-06 2.34E-06 2.63E-06 2.92E-06 3.51E-06

Skin 3.37E-07 6.74E-07 1.01E-06 1.35E-06 1.69E-06 2.02E-06 2.36E-06 2.70E-06 3.04E-06 3.37E-06 4.05E-06

Spleen 2.23E-07 4.46E-07 6.70E-07 8.93E-07 1.12E-06 1.34E-06 1.56E-06 1.79E-06 2.01E-06 2.23E-06 2.68E-06

Thymus 7.72E-07 1.54E-06 2.32E-06 3.09E-06 3.86E-06 4.63E-06 5.40E-06 6.18E-06 6.95E-06 7.72E-06 9.27E-06

Uterus 2.49E-07 4.98E-07 7.46E-07 9.96E-07 1.24E-06 1.49E-06 1.74E-06 1.99E-06 2.24E-06 2.49E-06 2.99E-06

Colon 2.51E-07 5.02E-07 7.53E-07 1.01E-06 1.26E-06 1.51E-06 1.76E-06 2.01E-06 2.26E-06 2.51E-06 3.02E-06

ET 4.34E-05 8.68E-05 1.30E-04 1.74E-04 2.17E-04 2.60E-04 3.04E-04 3.47E-04 3.91E-04 4.34E-04 5.21E-04

ET1 1.46E-02 2.91E-02 4.37E-02 5.83E-02 7.28E-02 8.73E-02 1.02E-01 1.17E-01 1.31E-01 1.46E-01 1.75E-01

ET2 2.89E-05 5.77E-05 8.66E-05 1.16E-04 1.44E-04 1.73E-04 2.02E-04 2.31E-04 2.60E-04 2.89E-04 3.47E-04

Lung 7.29E-07 1.46E-06 2.19E-06 2.92E-06 3.65E-06 4.37E-06 5.10E-06 5.83E-06 6.56E-06 7.29E-06 8.75E-06

LN(ET) 2.59E-05 5.18E-05 7.78E-05 1.04E-04 1.30E-04 1.56E-04 1.82E-04 2.08E-04 2.33E-04 2.59E-04 3.11E-04

BBsec 1.14E-06 2.28E-06 3.42E-06 4.57E-06 5.71E-06 6.85E-06 7.99E-06 9.14E-06 1.03E-05 1.14E-05 1.37E-05

BBbas 1.01E-06 2.02E-06 3.02E-06 4.04E-06 5.04E-06 6.05E-06 7.06E-06 8.07E-06 9.08E-06 1.01E-05 1.21E-05

bb 5.87E-07 1.17E-06 1.76E-06 2.35E-06 2.94E-06 3.52E-06 4.11E-06 4.70E-06 5.28E-06 5.87E-06 7.05E-06

AI 5.25E-07 1.05E-06 1.58E-06 2.10E-06 2.63E-06 3.15E-06 3.68E-06 4.21E-06 4.73E-06 5.26E-06 6.31E-06

LN(TH) 5.14E-07 1.03E-06 1.54E-06 2.06E-06 2.57E-06 3.08E-06 3.60E-06 4.11E-06 4.63E-06 5.14E-06 6.17E-06

Esophagus 7.72E-07 1.54E-06 2.32E-06 3.09E-06 3.86E-06 4.63E-06 5.40E-06 6.18E-06 6.95E-06 7.72E-06 9.27E-06

Gonads 2.09E-07 4.19E-07 6.28E-07 8.38E-07 1.05E-06 1.26E-06 1.47E-06 1.68E-06 1.89E-06 2.10E-06 2.51E-06
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 217 of 248

Table D-17 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 3.12E-06 3.57E-06 4.02E-06 4.46E-06 5.58E-06 6.69E-06 7.81E-06 8.93E-06 1.00E-05 1.12E-05

Urinary Bladder 4.64E-05 5.30E-05 5.96E-05 6.63E-05 8.29E-05 9.94E-05 1.16E-04 1.33E-04 1.49E-04 1.66E-04

Brain 1.03E-05 1.18E-05 1.33E-05 1.47E-05 1.84E-05 2.21E-05 2.58E-05 2.95E-05 3.32E-05 3.68E-05

Breast 4.01E-06 4.58E-06 5.15E-06 5.73E-06 7.16E-06 8.59E-06 1.00E-05 1.15E-05 1.29E-05 1.43E-05

Gall Bladder 2.81E-06 3.21E-06 3.61E-06 4.01E-06 5.01E-06 6.02E-06 7.02E-06 8.02E-06 9.02E-06 1.00E-05

Heart Wall 5.04E-06 5.76E-06 6.48E-067.20E-06 9.00E-06 1.08E-05 1.26E-05 1.44E-05 1.62E-05 1.80E-05

Kidneys 2.79E-06 3.19E-06 3.59E-06 3.99E-06 4.99E-06 5.98E-06 6.98E-06 7.98E-06 8.98E-06 9.97E-06

Liver 3.07E-06 3.51E-06 3.95E-06 4.39E-06 5.48E-06 6.58E-06 7.68E-06 8.77E-06 9.87E-06 1.10E-05

Muscle 8.86E-06 1.01E-05 1.14E-05 1.27E-05 1.58E-05 1.90E-05 2.21E-05 2.53E-05 2.85E-05 3.16E-05

Ovaries 2.93E-06 3.35E-06 3.77E-06 4.19E-06 5.24E-06 6.29E-06 7.34E-06 8.38E-06 9.43E-06 1.05E-05

Pancreas 3.41E-06 3.90E-06 4.39E-06 4.88E-06 6.10E-06 7.31E-06 8.53E-06 9.75E-06 1.10E-05 1.22E-05

Testes 2.45E-06 2.80E-06 3.15E-06 3.50E-06 4.38E-06 5.25E-06 6.13E-06 7.01E-06 7.88E-06 8.76E-06

Thyroid 2.65E-02 3.03E-02 3.41E-02 3.79E-02 4.74E-02 5.69E-02 6.64E-02 7.59E-02 8.53E-02 9.48E-02

R.B.M. 6.96E-06 7.95E-06 8.95E-06 9.94E-06 1.24E-05 1.49E-05 1.74E-05 1.99E-05 2.24E-05 2.49E-05

Bone Surface 8.76E-06 1.00E-05 1.13E-05 1.25E-05 1.56E-05 1.88E-05 2.19E-05 2.50E-05 2.81E-05 3.13E-05

Stomach 9.28E-06 1.06E-05 1.19E-05 1.33E-05 1.66E-05 1.99E-05 2.32E-05 2.65E-05 2.98E-05 3.31E-05

S.I. 2.69E-06 3.07E-06 3.46E-06 3.84E-06 4.80E-06 5.76E-06 6.72E-06 7.68E-06 8.64E-06 9.60E-06

U.L.I. 3.08E-06 3.52E-06 3.96E-06 4.40E-06 5.50E-06 6.61E-06 7.71E-06 8.81E-06 9.91E-06 1.10E-05

L.L.I. 4.09E-06 4.67E-06 5.26E-06 5.84E-06 7.30E-06 8.77E-06 1.02E-05 1.17E-05 1.31E-05 1.46E-05

Skin 4.72E-06 5.40E-06 6.07E-06 6.75E-06 8.43E-06 1.01E-05 1.18E-05 1.35E-05 1.52E-05 1.69E-05

Spleen 3.13E-06 3.57E-06 4.02E-06 4.47E-06 5.58E-06 6.70E-06 7.82E-06 8.93E-06 1.01E-05 1.12E-05

Thymus 1.08E-05 1.24E-05 1.39E-05 1.54E-05 1.93E-05 2.32E-05 2.70E-05 3.09E-05 3.48E-05 3.86E-05

Uterus 3.49E-06 3.98E-06 4.48E-06 4.98E-06 6.22E-06 7.47E-06 8.71E-06 9.96E-06 1.12E-05 1.25E-05

Colon 3.52E-06 4.02E-06 4.52E-06 5.03E-06 6.28E-06 7.54E-06 8.80E-06 1.01E-05 1.13E-05 1.26E-05

ET 6.08E-04 6.95E-04 7.81E-04 8.68E-04 1.09E-03 1.30E-03 1.52E-03 1.74E-03 1.95E-03 2.17E-03

ET1 2.04E-01 2.33E-01 2.62E-01 2.91E-01 3.64E-01 4.37E-01 5.10E-01 5.83E-01 6.56E-01 7.28E-01

ET2 4.04E-04 4.62E-04 5.20E-04 5.78E-04 7.22E-04 8.66E-04 1.01E-03 1.16E-03 1.30E-03 1.44E-03

Lung 1.02E-05 1.17E-05 1.31E-05 1.46E-05 1.82E-05 2.19E-05 2.55E-05 2.92E-05 3.28E-05 3.65E-05

LN(ET) 3.63E-04 4.15E-04 4.67E-04 5.19E-04 6.48E-04 7.78E-04 9.08E-04 1.04E-03 1.17E-03 1.30E-03

BBsec 1.60E-05 1.83E-05 2.06E-05 2.28E-05 2.86E-05 3.43E-05 4.00E-05 4.57E-05 5.14E-05 5.71E-05

BBbas 1.41E-05 1.61E-05 1.82E-05 2.02E-05 2.52E-05 3.03E-05 3.53E-05 4.04E-05 4.54E-05 5.04E-05

bb 8.22E-06 9.39E-06 1.06E-05 1.17E-05 1.47E-05 1.76E-05 2.06E-05 2.35E-05 2.64E-05 2.94E-05

AI 7.36E-06 8.41E-06 9.46E-06 1.05E-05 1.31E-05 1.58E-05 1.84E-05 2.10E-05 2.37E-05 2.63E-05

LN(TH) 7.20E-06 8.23E-06 9.25E-06 1.03E-05 1.29E-05 1.54E-05 1.80E-05 2.06E-05 2.31E-05 2.57E-05

Esophagus 1.08E-05 1.24E-05 1.39E-05 1.54E-05 1.93E-05 2.32E-05 2.70E-05 3.09E-05 3.48E-05 3.86E-05

Gonads 2.93E-06 3.35E-06 3.77E-06 4.19E-06 5.24E-06 6.29E-06 7.34E-06 8.38E-06 9.43E-06 1.05E-05
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 218 of 248

If either 140Ba or 140La is used as the radionuclide to maximize the dose to the organ of interest, then assume an equal intake of the 

other radionuclide to also have occurred.

Table D-18. La-140, Type F, 50-year equivalent dose for years of chronic intake (Sv) MDA = 9.3 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 3.37E-05 6.73E-05 1.01E-04 1.35E-04 1.69E-04 2.02E-04 2.36E-04 2.70E-04 3.03E-04 3.37E-04 4.04E-04

Urinary Bladder 1.80E-05 3.61E-05 5.40E-05 7.20E-05 9.00E-05 1.08E-04 1.26E-04 1.44E-04 1.62E-04 1.80E-04 2.16E-04

Brain 1.31E-05 2.61E-05 3.93E-05 5.24E-05 6.55E-05 7.86E-05 9.17E-05 1.05E-04 1.18E-04 1.31E-04 1.57E-04

Breast 1.28E-05 2.55E-05 3.84E-05 5.12E-05 6.40E-05 7.68E-05 8.96E-05 1.02E-04 1.15E-04 1.28E-04 1.54E-04

Gall Bladder 5.79E-05 1.16E-04 1.74E-04 2.32E-04 2.90E-04 3.47E-04 4.05E-04 4.63E-04 5.21E-04 5.79E-04 6.95E-04

Heart Wall 2.22E-05 4.45E-05 6.66E-05 8.88E-05 1.11E-04 1.33E-04 1.55E-04 1.78E-04 2.00E-04 2.22E-04 2.66E-04

Kidneys 2.79E-05 5.59E-05 8.37E-05 1.12E-04 1.40E-04 1.67E-04 1.95E-04 2.23E-04 2.51E-04 2.79E-04 3.35E-04

Liver 2.66E-04 5.33E-04 7.98E-04 1.06E-03 1.33E-03 1.60E-03 1.86E-03 2.13E-03 2.39E-03 2.66E-03 3.19E-03

Muscle 1.95E-05 3.90E-05 5.85E-05 7.80E-05 9.75E-05 1.17E-04 1.37E-04 1.56E-04 1.76E-04 1.95E-04 2.34E-04

Ovaries 4.20E-05 8.40E-05 1.26E-04 1.68E-04 2.10E-04 2.52E-04 2.94E-04 3.36E-04 3.78E-04 4.20E-04 5.04E-04

Pancreas 3.10E-05 6.20E-05 9.30E-05 1.24E-04 1.55E-04 1.86E-04 2.17E-04 2.48E-04 2.79E-04 3.10E-04 3.72E-04

Testes 1.00E-05 2.01E-05 3.00E-05 4.00E-05 5.00E-05 6.00E-05 7.00E-05 8.00E-05 9.00E-05 1.00E-04 1.20E-04

Thyroid 1.41E-05 2.82E-05 4.23E-05 5.64E-05 7.05E-05 8.46E-05 9.87E-05 1.13E-04 1.27E-04 1.41E-04 1.69E-04

R.B.M. 4.14E-05 8.27E-05 1.24E-04 1.66E-04 2.07E-04 2.48E-04 2.90E-04 3.31E-04 3.73E-04 4.14E-04 4.97E-04

Bone Surface 3.64E-05 7.29E-05 1.09E-04 1.46E-04 1.82E-04 2.18E-04 2.55E-04 2.91E-04 3.28E-04 3.64E-04 4.37E-04

Stomach 3.90E-05 7.80E-05 1.17E-04 1.56E-04 1.95E-04 2.34E-04 2.73E-04 3.12E-04 3.51E-04 3.90E-04 4.68E-04

S.I. 7.46E-05 1.49E-04 2.24E-04 2.98E-04 3.73E-04 4.48E-04 5.22E-04 5.97E-04 6.71E-04 7.46E-04 8.95E-04

U.L.I. 2.19E-04 4.38E-04 6.57E-04 8.76E-04 1.10E-03 1.31E-03 1.53E-03 1.75E-03 1.97E-03 2.19E-03 2.63E-03

L.L.I. 3.89E-04 7.79E-04 1.17E-03 1.56E-03 1.95E-03 2.33E-03 2.72E-03 3.11E-03 3.50E-03 3.89E-03 4.67E-03

Skin 1.20E-05 2.41E-05 3.60E-05 4.80E-05 6.00E-05 7.20E-05 8.40E-05 9.60E-05 1.08E-04 1.20E-04 1.44E-04

Spleen 1.59E-05 3.19E-05 4.77E-05 6.36E-05 7.95E-05 9.54E-05 1.11E-04 1.27E-04 1.43E-04 1.59E-04 1.91E-04

Thymus 1.78E-05 3.56E-05 5.34E-05 7.12E-05 8.90E-05 1.07E-04 1.25E-04 1.42E-04 1.60E-04 1.78E-04 2.14E-04

Uterus 2.50E-05 5.00E-05 7.50E-05 1.00E-04 1.25E-04 1.50E-04 1.75E-04 2.00E-04 2.25E-04 2.50E-04 3.00E-04

ET 1.47E-03 2.93E-03 4.41E-03 5.88E-03 7.35E-03 8.82E-03 1.03E-02 1.18E-02 1.32E-02 1.47E-02 1.76E-02

ET1 1.48E-01 2.97E-01 4.44E-01 5.92E-01 7.40E-01 8.88E-01 1.04E+00 1.18E+00 1.33E+00 1.48E+00 1.78E+00

ET2 1.32E-03 2.64E-03 3.96E-03 5.28E-03 6.60E-03 7.92E-03 9.24E-03 1.06E-02 1.19E-02 1.32E-02 1.58E-02

Lung 2.42E-05 4.84E-05 7.26E-05 9.68E-05 1.21E-04 1.45E-04 1.69E-04 1.94E-04 2.18E-04 2.42E-04 2.90E-04

Colon 2.93E-04 5.85E-04 8.79E-04 1.17E-03 1.47E-03 1.76E-03 2.05E-03 2.34E-03 2.64E-03 2.93E-03 3.52E-03

LN(ET) 1.28E-03 2.56E-03 3.84E-03 5.12E-03 6.40E-03 7.68E-03 8.96E-03 1.02E-02 1.15E-02 1.28E-02 1.54E-02

BBsec 2.90E-05 5.80E-05 8.70E-05 1.16E-04 1.45E-04 1.74E-04 2.03E-04 2.32E-04 2.61E-04 2.90E-04 3.48E-04

BBbas 2.81E-05 5.63E-05 8.43E-05 1.12E-04 1.41E-04 1.69E-04 1.97E-04 2.25E-04 2.53E-04 2.81E-04 3.37E-04

bb 2.23E-05 4.45E-05 6.69E-05 8.92E-05 1.12E-04 1.34E-04 1.56E-04 1.78E-04 2.01E-04 2.23E-04 2.68E-04

AI 2.17E-05 4.34E-05 6.51E-05 8.68E-05 1.09E-04 1.30E-04 1.52E-04 1.74E-04 1.95E-04 2.17E-04 2.60E-04

LN(TH) 2.13E-05 4.27E-05 6.39E-05 8.52E-05 1.07E-04 1.28E-04 1.49E-04 1.70E-04 1.92E-04 2.13E-04 2.56E-04

Esophagus 1.78E-05 3.56E-05 5.34E-05 7.12E-05 8.90E-05 1.07E-04 1.25E-04 1.42E-04 1.60E-04 1.78E-04 2.14E-04

Gonads 4.20E-05 8.40E-05 1.26E-04 1.68E-04 2.10E-04 2.52E-04 2.94E-04 3.36E-04 3.78E-04 4.20E-04 5.04E-04

Intake Rate pCi/d 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9

Intake (pCi) 365 730 1095 1461 1826 2190 2556 2922 3287 3652 4383

Days 2.96E+06 5.91E+06 8.87E+06 1.18E+07 1.48E+07 1.77E+07 2.07E+07 2.37E+07 2.66E+07 2.96E+07 3.55E+07
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 219 of 248

Table D-18 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 4.72E-04 5.39E-04 6.07E-04 6.74E-04 8.43E-04 1.01E-03 1.18E-03 1.35E-03 1.52E-03 1.69E-03

Urinary Bladder 2.52E-04 2.88E-04 3.24E-04 3.60E-04 4.50E-04 5.40E-04 6.30E-04 7.20E-04 8.10E-04 9.00E-04

Brain 1.83E-04 2.10E-04 2.36E-04 2.62E-04 3.28E-04 3.93E-04 4.59E-04 5.24E-04 5.90E-04 6.55E-04

Breast 1.79E-04 2.05E-04 2.30E-04 2.56E-04 3.20E-04 3.84E-04 4.48E-04 5.12E-04 5.76E-04 6.40E-04

Gall Bladder 8.11E-04 9.26E-04 1.04E-03 1.16E-03 1.45E-03 1.74E-03 2.03E-03 2.32E-03 2.61E-03 2.90E-03

Heart Wall 3.11E-04 3.55E-04 4.00E-04 4.44E-04 5.55E-04 6.66E-04 7.77E-04 8.88E-04 9.99E-04 1.11E-03

Kidneys 3.91E-04 4.46E-04 5.02E-04 5.58E-04 6.98E-04 8.37E-04 9.77E-04 1.12E-03 1.26E-03 1.40E-03

Liver 3.72E-03 4.26E-03 4.79E-03 5.32E-03 6.65E-03 7.98E-03 9.31E-03 1.06E-02 1.20E-02 1.33E-02

Muscle 2.73E-04 3.12E-04 3.51E-04 3.90E-04 4.88E-04 5.85E-04 6.83E-04 7.80E-04 8.78E-04 9.75E-04

Ovaries 5.88E-04 6.72E-04 7.56E-04 8.40E-04 1.05E-03 1.26E-03 1.47E-03 1.68E-03 1.89E-03 2.10E-03

Pancreas 4.34E-04 4.96E-04 5.58E-04 6.20E-04 7.75E-04 9.30E-04 1.09E-03 1.24E-03 1.40E-03 1.55E-03

Testes 1.40E-04 1.60E-04 1.80E-04 2.00E-04 2.50E-04 3.00E-04 3.50E-04 4.00E-04 4.50E-04 5.00E-04

Thyroid 1.97E-04 2.26E-04 2.54E-04 2.82E-04 3.53E-04 4.23E-04 4.94E-04 5.64E-04 6.35E-04 7.05E-04

R.B.M. 5.80E-04 6.62E-04 7.45E-04 8.28E-04 1.04E-03 1.24E-03 1.45E-03 1.66E-03 1.86E-03 2.07E-03

Bone Surface 5.10E-04 5.82E-04 6.55E-04 7.28E-04 9.10E-04 1.09E-03 1.27E-03 1.46E-03 1.64E-03 1.82E-03

Stomach 5.46E-04 6.24E-04 7.02E-04 7.80E-04 9.75E-04 1.17E-03 1.37E-03 1.56E-03 1.76E-03 1.95E-03

S.I. 1.04E-03 1.19E-03 1.34E-03 1.49E-03 1.87E-03 2.24E-03 2.61E-03 2.98E-03 3.36E-03 3.73E-03

U.L.I. 3.07E-03 3.50E-03 3.94E-03 4.38E-03 5.48E-03 6.57E-03 7.67E-03 8.76E-03 9.86E-03 1.10E-02

L.L.I. 5.45E-03 6.22E-03 7.00E-03 7.78E-03 9.73E-03 1.17E-02 1.36E-02 1.56E-02 1.75E-02 1.95E-02

Skin 1.68E-04 1.92E-04 2.16E-04 2.40E-04 3.00E-04 3.60E-04 4.20E-04 4.80E-04 5.40E-04 6.00E-04

Spleen 2.23E-04 2.54E-04 2.86E-04 3.18E-04 3.98E-04 4.77E-04 5.57E-04 6.36E-04 7.16E-04 7.95E-04

Thymus 2.49E-04 2.85E-04 3.20E-04 3.56E-04 4.45E-04 5.34E-04 6.23E-04 7.12E-04 8.01E-04 8.90E-04

Uterus 3.50E-04 4.00E-04 4.50E-04 5.00E-04 6.25E-04 7.50E-04 8.75E-04 1.00E-03 1.13E-03 1.25E-03

ET 2.06E-02 2.35E-02 2.65E-02 2.94E-02 3.68E-02 4.41E-02 5.15E-02 5.88E-02 6.62E-02 7.35E-02

ET1 2.07E+00 2.37E+00 2.66E+00 2.96E+00 3.70E+00 4.44E+00 5.18E+00 5.92E+00 6.66E+00 7.40E+00

ET2 1.85E-02 2.11E-02 2.38E-02 2.64E-02 3.30E-02 3.96E-02 4.62E-02 5.28E-02 5.94E-02 6.60E-02

Lung 3.39E-04 3.87E-04 4.36E-04 4.84E-04 6.05E-04 7.26E-04 8.47E-04 9.68E-04 1.09E-03 1.21E-03

Colon 4.10E-03 4.69E-03 5.27E-03 5.86E-03 7.33E-03 8.79E-03 1.03E-02 1.17E-02 1.32E-02 1.47E-02

LN(ET) 1.79E-02 2.05E-02 2.30E-02 2.56E-02 3.20E-02 3.84E-02 4.48E-02 5.12E-02 5.76E-02 6.40E-02

BBsec 4.06E-04 4.64E-04 5.22E-04 5.80E-04 7.25E-04 8.70E-04 1.02E-03 1.16E-03 1.31E-03 1.45E-03

BBbas 3.93E-04 4.50E-04 5.06E-04 5.62E-04 7.03E-04 8.43E-04 9.84E-04 1.12E-03 1.26E-03 1.41E-03

bb 3.12E-04 3.57E-04 4.01E-04 4.46E-04 5.58E-04 6.69E-04 7.81E-04 8.92E-04 1.00E-03 1.12E-03

AI 3.04E-04 3.47E-04 3.91E-04 4.34E-04 5.43E-04 6.51E-04 7.60E-04 8.68E-04 9.77E-04 1.09E-03

LN(TH) 2.98E-04 3.41E-04 3.83E-04 4.26E-04 5.33E-04 6.39E-04 7.46E-04 8.52E-04 9.59E-04 1.07E-03

Esophagus 2.49E-04 2.85E-04 3.20E-04 3.56E-04 4.45E-04 5.34E-04 6.23E-04 7.12E-04 8.01E-04 8.90E-04

Gonads 5.88E-04 6.72E-04 7.56E-04 8.40E-04 1.05E-03 1.26E-03 1.47E-03 1.68E-03 1.89E-03 2.10E-03

Intake Rate pCi/d 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9 8098.9

Intake (pCi) 5113 5844 6574 7305 9131 10958 12784 14610 16436 18263

Days 4.14E+07 4.73E+07 5.32E+07 5.92E+07 7.40E+07 8.87E+07 1.04E+08 1.18E+08 1.33E+08 1.48E+08
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 220 of 248

If either 140Ba or 140La is used as the radionuclide to maximize the dose to the organ of interest, then assume an equal intake of the 

other radionuclide to also have occurred.

Table D-19. La-140, Type M, 50-year equivalent dose for years of chronic intake (Sv) MDA = 9.3 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 1.61E-05 3.21E-05 4.83E-05 6.44E-05 8.05E-05 9.66E-05 1.13E-04 1.29E-04 1.45E-04 1.61E-04 1.93E-04

Urinary Bladder 2.43E-05 4.87E-05 7.29E-05 9.72E-05 1.22E-04 1.46E-04 1.70E-04 1.94E-04 2.19E-04 2.43E-04 2.92E-04

Brain 8.54E-06 1.71E-05 2.56E-05 3.42E-05 4.27E-05 5.12E-05 5.98E-05 6.83E-05 7.69E-05 8.54E-05 1.02E-04

Breast 1.10E-05 2.21E-05 3.30E-05 4.40E-05 5.50E-05 6.60E-05 7.70E-05 8.80E-05 9.90E-05 1.10E-04 1.32E-04

Gall Bladder 3.54E-05 7.08E-05 1.06E-04 1.42E-04 1.77E-04 2.12E-04 2.48E-04 2.83E-04 3.19E-04 3.54E-04 4.25E-04

Heart Wall 1.79E-05 3.58E-05 5.37E-05 7.16E-05 8.95E-05 1.07E-04 1.25E-04 1.43E-04 1.61E-04 1.79E-04 2.15E-04

Kidneys 1.77E-05 3.54E-05 5.31E-05 7.08E-05 8.85E-05 1.06E-04 1.24E-04 1.42E-04 1.59E-04 1.77E-04 2.12E-04

Liver 4.85E-05 9.70E-05 1.46E-04 1.94E-04 2.43E-04 2.91E-04 3.40E-04 3.88E-04 4.37E-04 4.85E-04 5.82E-04

Muscle 1.87E-05 3.73E-05 5.61E-05 7.48E-05 9.35E-05 1.12E-04 1.31E-04 1.50E-04 1.68E-04 1.87E-04 2.24E-04

Ovaries 7.77E-05 1.55E-04 2.33E-04 3.11E-04 3.89E-04 4.66E-04 5.44E-04 6.22E-04 6.99E-04 7.77E-04 9.32E-04

Pancreas 1.94E-05 3.89E-05 5.82E-05 7.76E-05 9.70E-05 1.16E-04 1.36E-04 1.55E-04 1.75E-04 1.94E-04 2.33E-04

Testes 8.15E-06 1.63E-05 2.45E-05 3.26E-05 4.08E-05 4.89E-05 5.71E-05 6.52E-05 7.34E-05 8.15E-05 9.78E-05

Thyroid 1.08E-05 2.17E-05 3.24E-05 4.32E-05 5.40E-05 6.48E-05 7.56E-05 8.64E-05 9.72E-05 1.08E-04 1.30E-04

R.B.M. 2.53E-05 5.06E-05 7.59E-05 1.01E-04 1.27E-04 1.52E-04 1.77E-04 2.02E-04 2.28E-04 2.53E-04 3.04E-04

Bone Surface 1.72E-05 3.43E-05 5.16E-05 6.88E-05 8.60E-05 1.03E-04 1.20E-04 1.38E-04 1.55E-04 1.72E-04 2.06E-04

Stomach 6.53E-05 1.31E-04 1.96E-04 2.61E-04 3.27E-04 3.92E-04 4.57E-04 5.22E-04 5.88E-04 6.53E-04 7.84E-04

S.I. 1.53E-04 3.05E-04 4.59E-04 6.12E-04 7.65E-04 9.18E-04 1.07E-03 1.22E-03 1.38E-03 1.53E-03 1.84E-03

U.L.I. 5.08E-04 1.02E-03 1.52E-03 2.03E-03 2.54E-03 3.05E-03 3.56E-03 4.06E-03 4.57E-03 5.08E-03 6.10E-03

L.L.I. 9.55E-04 1.91E-03 2.87E-03 3.82E-03 4.78E-03 5.73E-03 6.69E-03 7.64E-03 8.60E-03 9.55E-03 1.15E-02

Skin 8.85E-06 1.77E-05 2.66E-05 3.54E-05 4.43E-05 5.31E-05 6.20E-05 7.08E-05 7.97E-05 8.85E-05 1.06E-04

Spleen 1.50E-05 2.99E-05 4.50E-05 6.00E-05 7.50E-05 9.00E-05 1.05E-04 1.20E-04 1.35E-04 1.50E-04 1.80E-04

Thymus 1.75E-05 3.49E-05 5.25E-05 7.00E-05 8.75E-05 1.05E-04 1.23E-04 1.40E-04 1.58E-04 1.75E-04 2.10E-04

Uterus 3.69E-05 7.38E-05 1.11E-04 1.48E-04 1.85E-04 2.21E-04 2.58E-04 2.95E-04 3.32E-04 3.69E-04 4.43E-04

ET 1.94E-03 3.88E-03 5.82E-03 7.76E-03 9.70E-03 1.16E-02 1.36E-02 1.55E-02 1.75E-02 1.94E-02 2.33E-02

ET1 1.89E-01 3.78E-01 5.67E-01 7.56E-01 9.45E-01 1.13E+00 1.32E+00 1.51E+00 1.70E+00 1.89E+00 2.27E+00

ET2 1.75E-03 3.51E-03 5.25E-03 7.00E-03 8.75E-03 1.05E-02 1.23E-02 1.40E-02 1.58E-02 1.75E-02 2.10E-02

Lung 3.95E-04 7.90E-04 1.19E-03 1.58E-03 1.98E-03 2.37E-03 2.77E-03 3.16E-03 3.56E-03 3.95E-03 4.74E-03

Colon 7.01E-04 1.40E-03 2.10E-03 2.80E-03 3.51E-03 4.21E-03 4.91E-03 5.61E-03 6.31E-03 7.01E-03 8.41E-03

LN(ET) 1.64E-03 3.28E-03 4.92E-03 6.56E-03 8.20E-03 9.84E-03 1.15E-02 1.31E-02 1.48E-02 1.64E-02 1.97E-02

BBsec 8.73E-04 1.75E-03 2.62E-03 3.49E-03 4.37E-03 5.24E-03 6.11E-03 6.98E-03 7.86E-03 8.73E-03 1.05E-02

BBbas 7.83E-04 1.57E-03 2.35E-03 3.13E-03 3.92E-03 4.70E-03 5.48E-03 6.26E-03 7.05E-03 7.83E-03 9.40E-03

bb 2.16E-04 4.32E-04 6.48E-04 8.64E-04 1.08E-03 1.30E-03 1.51E-03 1.73E-03 1.94E-03 2.16E-03 2.59E-03

AI 1.42E-04 2.84E-04 4.26E-04 5.68E-04 7.10E-04 8.52E-04 9.94E-04 1.14E-03 1.28E-03 1.42E-03 1.70E-03

LN(TH) 3.62E-05 7.23E-05 1.09E-04 1.45E-04 1.81E-04 2.17E-04 2.53E-04 2.90E-04 3.26E-04 3.62E-04 4.34E-04

Esophagus 1.75E-05 3.49E-05 5.25E-05 7.00E-05 8.75E-05 1.05E-04 1.23E-04 1.40E-04 1.58E-04 1.75E-04 2.10E-04

Gonads 7.77E-05 1.55E-04 2.33E-04 3.11E-04 3.89E-04 4.66E-04 5.44E-04 6.22E-04 6.99E-04 7.77E-04 9.32E-04

Intake Rate pCi/d 10331 10331 10331 10331 10331 10331 10331 10331 10331 10331 10331

Intake (pCi) 365 730 1095 1461 1826 2190 2556 2922 3287 3652 4383

Days 3.77E+06 7.54E+06 1.13E+07 1.51E+07 1.89E+07 2.26E+07 2.64E+07 3.02E+07 3.40E+07 3.77E+07 4.53E+07
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 221 of 248

Table D-19 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 2.25E-04 2.58E-04 2.90E-04 3.22E-04 4.03E-04 4.83E-04 5.64E-04 6.44E-04 7.25E-04 8.05E-04

Urinary Bladder 3.40E-04 3.89E-04 4.37E-04 4.86E-04 6.08E-04 7.29E-04 8.51E-04 9.72E-04 1.09E-03 1.22E-03

Brain 1.20E-04 1.37E-04 1.54E-04 1.71E-04 2.14E-04 2.56E-04 2.99E-04 3.42E-04 3.84E-04 4.27E-04

Breast 1.54E-04 1.76E-04 1.98E-04 2.20E-04 2.75E-04 3.30E-04 3.85E-04 4.40E-04 4.95E-04 5.50E-04

Gall Bladder 4.96E-04 5.66E-04 6.37E-04 7.08E-04 8.85E-04 1.06E-03 1.24E-03 1.42E-03 1.59E-03 1.77E-03

Heart Wall 2.51E-04 2.86E-04 3.22E-04 3.58E-04 4.48E-04 5.37E-04 6.27E-04 7.16E-04 8.06E-04 8.95E-04

Kidneys 2.48E-04 2.83E-04 3.19E-04 3.54E-04 4.43E-04 5.31E-04 6.20E-04 7.08E-04 7.97E-04 8.85E-04

Liver 6.79E-04 7.76E-04 8.73E-04 9.70E-04 1.21E-03 1.46E-03 1.70E-03 1.94E-032.18E-03 2.43E-03

Muscle 2.62E-04 2.99E-04 3.37E-04 3.74E-04 4.68E-04 5.61E-04 6.55E-04 7.48E-04 8.42E-04 9.35E-04

Ovaries 1.09E-03 1.24E-03 1.40E-03 1.55E-03 1.94E-03 2.33E-03 2.72E-03 3.11E-03 3.50E-03 3.89E-03

Pancreas 2.72E-04 3.10E-04 3.49E-04 3.88E-04 4.85E-04 5.82E-04 6.79E-04 7.76E-04 8.73E-04 9.70E-04

Testes 1.14E-04 1.30E-04 1.47E-04 1.63E-04 2.04E-04 2.45E-04 2.85E-04 3.26E-04 3.67E-04 4.08E-04

Thyroid 1.51E-04 1.73E-04 1.94E-04 2.16E-04 2.70E-04 3.24E-04 3.78E-04 4.32E-04 4.86E-04 5.40E-04

R.B.M. 3.54E-04 4.05E-04 4.55E-04 5.06E-04 6.33E-04 7.59E-04 8.86E-04 1.01E-03 1.14E-03 1.27E-03

Bone Surface 2.41E-04 2.75E-04 3.10E-04 3.44E-04 4.30E-04 5.16E-04 6.02E-04 6.88E-04 7.74E-04 8.60E-04

Stomach 9.14E-04 1.04E-03 1.18E-03 1.31E-03 1.63E-031.96E-03 2.29E-03 2.61E-03 2.94E-03 3.27E-03

S.I. 2.14E-03 2.45E-03 2.75E-03 3.06E-03 3.83E-03 4.59E-03 5.36E-03 6.12E-03 6.89E-03 7.65E-03

U.L.I. 7.11E-03 8.13E-03 9.14E-03 1.02E-02 1.27E-02 1.52E-02 1.78E-02 2.03E-02 2.29E-02 2.54E-02

L.L.I. 1.34E-021.53E-02 1.72E-02 1.91E-02 2.39E-02 2.87E-02 3.34E-02 3.82E-02 4.30E-02 4.78E-02

Skin 1.24E-04 1.42E-04 1.59E-04 1.77E-04 2.21E-04 2.66E-04 3.10E-04 3.54E-04 3.98E-04 4.43E-04

Spleen 2.10E-04 2.40E-04 2.70E-04 3.00E-04 3.75E-04 4.50E-04 5.25E-04 6.00E-04 6.75E-04 7.50E-04

Thymus 2.45E-04 2.80E-04 3.15E-04 3.50E-04 4.38E-04 5.25E-04 6.13E-04 7.00E-04 7.88E-04 8.75E-04

Uterus 5.17E-04 5.90E-04 6.64E-04 7.38E-04 9.23E-04 1.11E-03 1.29E-03 1.48E-03 1.66E-03 1.85E-03

ET 2.72E-02 3.10E-02 3.49E-02 3.88E-02 4.85E-02 5.82E-02 6.79E-02 7.76E-02 8.73E-02 9.70E-02

ET1 2.65E+00 3.02E+00 3.40E+00 3.78E+00 4.73E+00 5.67E+00 6.62E+00 7.56E+00 8.51E+00 9.45E+00

ET2 2.45E-02 2.80E-02 3.15E-02 3.50E-02 4.38E-02 5.25E-02 6.13E-02 7.00E-02 7.88E-02 8.75E-02

Lung 5.53E-03 6.32E-03 7.11E-03 7.90E-03 9.88E-03 1.19E-02 1.38E-02 1.58E-02 1.78E-02 1.98E-02

Colon 9.81E-03 1.12E-02 1.26E-02 1.40E-02 1.75E-02 2.10E-02 2.45E-02 2.80E-02 3.15E-02 3.51E-02

LN(ET) 2.30E-02 2.62E-02 2.95E-02 3.28E-02 4.10E-02 4.92E-02 5.74E-02 6.56E-02 7.38E-02 8.20E-02

BBsec 1.22E-02 1.40E-02 1.57E-02 1.75E-02 2.18E-02 2.62E-02 3.06E-02 3.49E-02 3.93E-02 4.37E-02

BBbas 1.10E-02 1.25E-02 1.41E-02 1.57E-02 1.96E-02 2.35E-02 2.74E-02 3.13E-02 3.52E-02 3.92E-02

bb 3.02E-03 3.46E-03 3.89E-03 4.32E-035.40E-03 6.48E-03 7.56E-03 8.64E-03 9.72E-03 1.08E-02

AI 1.99E-03 2.27E-03 2.56E-03 2.84E-03 3.55E-03 4.26E-03 4.97E-03 5.68E-03 6.39E-03 7.10E-03

LN(TH) 5.07E-04 5.79E-04 6.52E-04 7.24E-04 9.05E-04 1.09E-03 1.27E-03 1.45E-03 1.63E-03 1.81E-03

Esophagus 2.45E-04 2.80E-04 3.15E-04 3.50E-04 4.38E-04 5.25E-04 6.13E-04 7.00E-04 7.88E-04 8.75E-04

Gonads 1.09E-03 1.24E-03 1.40E-03 1.55E-03 1.94E-03 2.33E-03 2.72E-03 3.11E-03 3.50E-03 3.89E-03

Intake Rate pCi/d 10331 10331 10331 10331 10331 1033110331 10331 10331 10331

Intake (pCi) 5113 5844 6574 7305 9131 10958 12784 14610 16436 18263

Days 5.28E+07 6.04E+07 6.79E+07 7.55E+07 9.43E+07 1.13E+08 1.32E+08 1.51E+08 1.70E+08 1.89E+08
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 222 of 248

Table D-20. Mn-54, Type M, 50-year equivalent dose for years of chronic intake (Sv) MDA = 1 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 1.42E-06 2.76E-06 4.13E-06 5.51E-06 6.89E-06 8.26E-06 9.65E-06 1.10E-05 1.24E-05 1.38E-05 1.65E-05

Urinary Bladder 4.05E-07 7.86E-07 1.18E-06 1.57E-06 1.96E-06 2.36E-06 2.75E-06 3.14E-06 3.54E-06 3.93E-06 4.71E-06

Brain 3.25E-07 6.30E-07 9.44E-07 1.26E-06 1.57E-06 1.89E-06 2.20E-06 2.52E-06 2.83E-06 3.15E-06 3.78E-06

Breast 1.08E-06 2.10E-06 3.14E-06 4.19E-06 5.24E-06 6.28E-06 7.33E-06 8.38E-06 9.43E-06 1.05E-05 1.26E-05

Gall Bladder 1.08E-06 2.10E-06 3.14E-06 4.19E-06 5.24E-06 6.28E-06 7.33E-06 8.38E-06 9.43E-06 1.05E-05 1.26E-05

Heart Wall 1.82E-06 3.53E-06 5.28E-06 7.05E-06 8.81E-06 1.06E-05 1.23E-05 1.41E-05 1.59E-05 1.76E-05 2.11E-05

Kidneys 7.76E-07 1.51E-06 2.26E-06 3.01E-06 3.76E-06 4.51E-06 5.27E-06 6.02E-06 6.77E-06 7.53E-06 9.03E-06

Liver 2.04E-06 3.97E-06 5.94E-06 7.93E-06 9.91E-06 1.19E-05 1.39E-05 1.59E-05 1.78E-05 1.98E-05 2.38E-05

Muscle 6.89E-07 1.34E-06 2.00E-06 2.67E-06 3.34E-06 4.00E-06 4.67E-06 5.34E-06 6.01E-06 6.68E-06 8.01E-06

Ovaries 8.98E-07 1.74E-06 2.61E-06 3.48E-06 4.35E-06 5.22E-06 6.09E-06 6.97E-06 7.84E-06 8.71E-06 1.04E-05

Pancreas 1.09E-06 2.12E-06 3.18E-06 4.24E-06 5.30E-06 6.36E-06 7.42E-06 8.49E-06 9.55E-06 1.06E-05 1.27E-05

Testes 2.18E-07 4.24E-07 6.34E-07 8.46E-07 1.06E-06 1.27E-06 1.48E-06 1.69E-06 1.90E-06 2.12E-06 2.54E-06

Thyroid 5.93E-07 1.15E-06 1.72E-06 2.30E-06 2.88E-06 3.45E-06 4.03E-06 4.60E-06 5.18E-06 5.75E-06 6.90E-06

R.B.M. 1.17E-06 2.27E-06 3.40E-06 4.54E-06 5.67E-06 6.80E-06 7.94E-06 9.08E-06 1.02E-05 1.13E-05 1.36E-05

Bone Surface 1.17E-06 2.26E-06 3.39E-06 4.52E-06 5.65E-06 6.78E-06 7.91E-06 9.04E-06 1.02E-05 1.13E-05 1.36E-05

Stomach 8.38E-07 1.63E-06 2.44E-06 3.25E-06 4.06E-06 4.87E-06 5.69E-06 6.50E-06 7.31E-06 8.12E-06 9.75E-06

S.I. 8.75E-07 1.70E-06 2.54E-06 3.39E-06 4.24E-06 5.09E-06 5.94E-06 6.79E-06 7.63E-06 8.48E-06 1.02E-05

U.L.I. 1.18E-06 2.30E-06 3.44E-06 4.59E-06 5.74E-06 6.88E-06 8.03E-06 9.18E-06 1.03E-05 1.15E-05 1.38E-05

L.L.I. 1.69E-06 3.28E-06 4.92E-06 6.56E-06 8.20E-06 9.83E-06 1.15E-05 1.31E-05 1.48E-05 1.64E-05 1.97E-05

Skin 3.89E-07 7.54E-07 1.13E-06 1.51E-06 1.88E-06 2.26E-06 2.64E-06 3.01E-06 3.39E-06 3.77E-06 4.52E-06

Spleen 8.89E-07 1.73E-06 2.59E-06 3.45E-06 4.31E-06 5.17E-06 6.04E-06 6.90E-06 7.76E-06 8.62E-06 1.04E-05

Thymus 1.29E-06 2.50E-06 3.74E-06 4.99E-06 6.24E-06 7.49E-06 8.74E-06 9.99E-06 1.12E-05 1.25E-05 1.50E-05

Uterus 5.38E-07 1.04E-06 1.56E-06 2.09E-06 2.61E-06 3.13E-06 3.65E-06 4.17E-06 4.70E-06 5.22E-06 6.26E-06

Colon 1.40E-06 2.72E-06 4.08E-06 5.44E-06 6.80E-06 8.16E-06 9.52E-06 1.09E-05 1.22E-05 1.36E-05 1.63E-05

ET 1.06E-05 2.05E-05 3.08E-05 4.11E-05 5.13E-05 6.15E-05 7.18E-05 8.21E-05 9.24E-05 1.03E-04 1.23E-04

ET1 1.12E-05 2.18E-05 3.26E-05 4.35E-05 5.44E-05 6.53E-05 7.62E-05 8.71E-05 9.79E-05 1.09E-04 1.31E-04

ET2 1.06E-05 2.05E-05 3.08E-05 4.11E-05 5.13E-05 6.15E-05 7.18E-05 8.21E-05 9.24E-05 1.03E-04 1.23E-04

Lung 4.49E-06 8.72E-06 1.31E-05 1.74E-05 2.18E-05 2.61E-05 3.05E-05 3.48E-05 3.92E-05 4.35E-05 5.23E-05

LN(ET) 1.06E-05 2.06E-05 3.08E-05 4.11E-05 5.14E-05 6.16E-05 7.19E-05 8.22E-05 9.25E-05 1.03E-04 1.23E-04

BBsec 4.39E-06 8.51E-06 1.28E-05 1.70E-05 2.13E-05 2.55E-05 2.98E-05 3.40E-05 3.83E-05 4.25E-05 5.10E-05

BBbas 4.39E-06 8.51E-06 1.28E-05 1.70E-05 2.13E-05 2.55E-05 2.98E-05 3.40E-05 3.83E-05 4.25E-05 5.10E-05

bb 4.37E-06 8.49E-06 1.27E-05 1.70E-05 2.12E-05 2.54E-05 2.97E-05 3.39E-05 3.82E-05 4.24E-05 5.09E-05

AI 4.71E-06 9.15E-06 1.37E-05 1.83E-05 2.29E-05 2.74E-05 3.20E-05 3.66E-05 4.11E-05 4.57E-05 5.49E-05

LN(TH) 4.40E-06 8.55E-06 1.28E-05 1.71E-05 2.13E-05 2.56E-05 2.99E-05 3.41E-05 3.84E-05 4.27E-05 5.12E-05

Esophagus 1.29E-06 2.50E-06 3.74E-06 4.99E-06 6.24E-06 7.49E-06 8.74E-06 9.99E-06 1.12E-05 1.25E-05 1.50E-05

Gonads 8.98E-07 1.74E-06 2.61E-06 3.48E-06 4.35E-06 5.22E-06 6.09E-06 6.97E-06 7.84E-06 8.71E-06 1.04E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 223 of 248

Table D-20 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 1.93E-05 2.21E-05 2.48E-05 2.76E-05 3.45E-05 4.14E-05 4.82E-05 5.51E-05 6.20E-05 6.89E-05

Urinary Bladder 5.50E-06 6.29E-06 7.07E-06 7.86E-06 9.82E-06 1.18E-05 1.37E-05 1.57E-05 1.77E-05 1.96E-05

Brain 4.41E-06 5.04E-06 5.67E-06 6.30E-06 7.87E-06 9.44E-06 1.10E-05 1.26E-05 1.42E-05 1.57E-05

Breast 1.47E-05 1.68E-05 1.89E-05 2.10E-05 2.62E-05 3.14E-05 3.67E-05 4.19E-05 4.71E-05 5.24E-05

Gall Bladder 1.47E-05 1.68E-05 1.89E-05 2.10E-05 2.62E-05 3.14E-05 3.67E-05 4.19E-05 4.72E-05 5.24E-05

Heart Wall 2.47E-05 2.82E-05 3.17E-05 3.52E-05 4.41E-05 5.29E-05 6.17E-05 7.05E-05 7.93E-05 8.81E-05

Kidneys 1.05E-05 1.20E-05 1.35E-05 1.51E-05 1.88E-05 2.26E-05 2.63E-05 3.01E-05 3.39E-05 3.76E-05

Liver 2.77E-05 3.17E-05 3.57E-05 3.96E-05 4.95E-05 5.94E-05 6.94E-05 7.93E-05 8.92E-05 9.91E-05

Muscle 9.35E-06 1.07E-05 1.20E-05 1.34E-05 1.67E-05 2.00E-05 2.34E-05 2.67E-05 3.01E-05 3.34E-05

Ovaries 1.22E-05 1.39E-05 1.57E-05 1.74E-05 2.18E-05 2.61E-05 3.05E-05 3.48E-05 3.92E-05 4.35E-05

Pancreas 1.49E-05 1.70E-05 1.91E-05 2.12E-05 2.65E-05 3.18E-05 3.71E-05 4.24E-05 4.77E-05 5.30E-05

Testes 2.96E-06 3.39E-06 3.81E-06 4.23E-06 5.29E-06 6.35E-06 7.41E-06 8.46E-06 9.52E-06 1.06E-05

Thyroid 8.05E-06 9.20E-06 1.04E-05 1.15E-05 1.44E-05 1.73E-05 2.01E-05 2.30E-05 2.59E-05 2.88E-05

R.B.M. 1.59E-05 1.82E-05 2.04E-05 2.27E-05 2.84E-05 3.40E-05 3.97E-05 4.54E-05 5.11E-05 5.67E-05

Bone Surface 1.58E-05 1.81E-05 2.03E-05 2.26E-05 2.82E-05 3.39E-05 3.96E-05 4.52E-05 5.08E-05 5.65E-05

Stomach 1.14E-05 1.30E-05 1.46E-05 1.62E-05 2.03E-05 2.44E-05 2.84E-05 3.25E-05 3.66E-05 4.06E-05

S.I. 1.19E-05 1.36E-05 1.53E-05 1.70E-05 2.12E-05 2.54E-05 2.97E-05 3.39E-05 3.82E-05 4.24E-05

U.L.I. 1.61E-05 1.84E-05 2.07E-05 2.30E-05 2.87E-05 3.44E-05 4.02E-05 4.59E-05 5.16E-05 5.74E-05

L.L.I. 2.30E-05 2.62E-05 2.95E-05 3.28E-05 4.10E-05 4.92E-05 5.74E-05 6.56E-05 7.38E-05 8.20E-05

Skin 5.27E-06 6.03E-06 6.78E-06 7.54E-06 9.42E-06 1.13E-05 1.32E-05 1.51E-05 1.70E-05 1.88E-05

Spleen 1.21E-05 1.38E-05 1.55E-05 1.73E-05 2.16E-05 2.59E-05 3.02E-05 3.45E-05 3.88E-05 4.31E-05

Thymus 1.75E-05 2.00E-05 2.25E-05 2.50E-05 3.12E-05 3.75E-05 4.37E-05 4.99E-05 5.62E-05 6.24E-05

Uterus 7.30E-06 8.35E-06 9.39E-06 1.04E-05 1.30E-05 1.57E-05 1.83E-05 2.09E-05 2.35E-05 2.61E-05

Colon 1.90E-05 2.18E-05 2.45E-05 2.72E-05 3.40E-05 4.08E-05 4.76E-05 5.44E-05 6.12E-05 6.80E-05

ET 1.44E-04 1.64E-04 1.85E-04 2.05E-04 2.57E-04 3.08E-04 3.59E-04 4.11E-04 4.62E-04 5.13E-04

ET1 1.52E-04 1.74E-04 1.96E-04 2.18E-04 2.72E-043.27E-04 3.81E-04 4.35E-04 4.90E-04 5.44E-04

ET2 1.44E-04 1.64E-04 1.85E-04 2.05E-04 2.57E-04 3.08E-04 3.59E-04 4.11E-04 4.62E-04 5.13E-04

Lung 6.10E-05 6.97E-05 7.84E-05 8.71E-05 1.09E-04 1.31E-04 1.52E-04 1.74E-04 1.96E-04 2.18E-04

LN(ET) 1.44E-04 1.64E-04 1.85E-04 2.06E-04 2.57E-04 3.08E-04 3.60E-04 4.11E-04 4.63E-04 5.14E-04

BBsec 5.95E-05 6.81E-05 7.66E-05 8.51E-05 1.06E-04 1.28E-04 1.49E-04 1.70E-04 1.91E-04 2.13E-04

BBbas 5.95E-05 6.80E-05 7.65E-05 8.51E-05 1.06E-04 1.28E-04 1.49E-04 1.70E-04 1.91E-04 2.13E-04

bb 5.94E-05 6.79E-05 7.63E-05 8.48E-05 1.06E-04 1.27E-04 1.48E-04 1.70E-04 1.91E-04 2.12E-04

AI 6.40E-05 7.32E-05 8.23E-05 9.14E-05 1.14E-04 1.37E-04 1.60E-04 1.83E-04 2.06E-04 2.29E-04

LN(TH) 5.98E-05 6.83E-05 7.68E-05 8.54E-05 1.07E-04 1.28E-04 1.49E-04 1.71E-04 1.92E-04 2.13E-04

Esophagus 1.75E-05 2.00E-05 2.25E-05 2.50E-05 3.12E-05 3.75E-05 4.37E-05 4.99E-05 5.62E-05 6.24E-05

Gonads 1.22E-05 1.39E-05 1.57E-05 1.74E-05 2.18E-05 2.61E-05 3.05E-05 3.48E-05 3.92E-05 4.35E-05
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 224 of 248

If either 95Zr or 95Nb is used as the radionuclide to maximize the dose to the organ of interest, then assume an equal intake of the 

other radionuclide to also have occurred.

Table D-21. Nb-95, Type M, 50-year equivalent dose for years of chronic intake (Sv) MDA = 2.2 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 2.48E-06 4.95E-06 7.43E-06 9.91E-06 1.24E-05 1.49E-05 1.73E-05 1.98E-05 2.23E-05 2.48E-05 2.97E-05

Urinary Bladder 1.19E-06 2.38E-06 3.58E-06 4.77E-06 5.96E-06 7.15E-06 8.35E-06 9.54E-06 1.07E-05 1.19E-05 1.43E-05

Brain 6.25E-07 1.25E-06 1.88E-06 2.50E-06 3.13E-06 3.75E-06 4.38E-06 5.01E-06 5.63E-06 6.26E-06 7.51E-06

Breast 1.93E-06 3.86E-06 5.79E-06 7.72E-06 9.66E-06 1.16E-05 1.35E-05 1.55E-05 1.74E-05 1.93E-05 2.32E-05

Gall Bladder 2.14E-06 4.28E-06 6.42E-06 8.57E-06 1.07E-05 1.28E-05 1.50E-05 1.71E-05 1.93E-05 2.14E-05 2.57E-05

Heart Wall 3.22E-06 6.44E-06 9.67E-06 1.29E-05 1.61E-05 1.93E-05 2.26E-05 2.58E-05 2.90E-05 3.22E-05 3.87E-05

Kidneys 2.41E-06 4.83E-06 7.24E-06 9.66E-06 1.21E-05 1.45E-05 1.69E-05 1.93E-05 2.17E-05 2.41E-05 2.90E-05

Liver 3.68E-06 7.36E-06 1.10E-05 1.47E-05 1.84E-05 2.21E-05 2.58E-05 2.95E-05 3.32E-05 3.68E-05 4.42E-05

Muscle 1.43E-06 2.87E-06 4.30E-06 5.73E-06 7.17E-06 8.60E-06 1.00E-05 1.15E-05 1.29E-05 1.43E-05 1.72E-05

Ovaries 3.05E-06 6.10E-06 9.15E-06 1.22E-05 1.53E-05 1.83E-05 2.14E-05 2.44E-05 2.75E-05 3.05E-05 3.66E-05

Pancreas 2.05E-06 4.09E-06 6.14E-06 8.19E-06 1.02E-05 1.23E-05 1.43E-05 1.64E-05 1.84E-05 2.05E-05 2.46E-05

Testes 5.30E-07 1.06E-06 1.59E-06 2.12E-06 2.65E-06 3.18E-06 3.71E-06 4.24E-06 4.78E-06 5.31E-06 6.37E-06

Thyroid 1.13E-06 2.25E-06 3.38E-06 4.51E-06 5.64E-06 6.76E-06 7.89E-06 9.02E-06 1.01E-05 1.13E-05 1.35E-05

R.B.M. 2.46E-06 4.91E-06 7.37E-06 9.84E-06 1.23E-05 1.47E-05 1.72E-05 1.97E-05 2.21E-05 2.46E-05 2.95E-05

Bone Surface 5.50E-06 1.10E-05 1.65E-05 2.20E-05 2.75E-05 3.30E-05 3.85E-05 4.40E-05 4.95E-05 5.50E-05 6.60E-05

Stomach 2.03E-06 4.05E-06 6.08E-06 8.11E-06 1.01E-05 1.22E-05 1.42E-05 1.62E-05 1.82E-05 2.03E-05 2.43E-05

S.I. 3.16E-06 6.33E-06 9.49E-06 1.27E-05 1.58E-05 1.90E-05 2.22E-05 2.53E-05 2.85E-05 3.17E-05 3.80E-05

U.L.I. 6.32E-06 1.26E-05 1.90E-05 2.53E-05 3.16E-05 3.79E-05 4.43E-05 5.06E-05 5.70E-05 6.33E-05 7.59E-05

L.L.I. 1.29E-05 2.59E-05 3.88E-05 5.18E-05 6.48E-05 7.77E-05 9.06E-05 1.04E-04 1.17E-04 1.30E-04 1.55E-04

Skin 7.65E-07 1.53E-06 2.29E-06 3.06E-06 3.83E-06 4.59E-06 5.36E-06 6.12E-06 6.89E-06 7.65E-06 9.18E-06

Spleen 1.74E-06 3.48E-06 5.22E-06 6.96E-06 8.70E-06 1.04E-05 1.22E-05 1.39E-05 1.57E-05 1.74E-05 2.09E-05

Thymus 2.38E-06 4.76E-06 7.14E-06 9.52E-06 1.19E-05 1.43E-05 1.67E-05 1.90E-05 2.14E-05 2.38E-05 2.86E-05

Uterus 1.59E-06 3.18E-06 4.77E-06 6.36E-06 7.95E-06 9.53E-06 1.11E-05 1.27E-05 1.43E-05 1.59E-05 1.91E-05

Colon 9.19E-06 1.84E-05 2.76E-05 3.68E-05 4.60E-05 5.51E-05 6.43E-05 7.35E-05 8.27E-05 9.19E-05 1.10E-04

ET 5.71E-05 1.14E-04 1.71E-04 2.29E-04 2.86E-04 3.43E-04 4.00E-04 4.57E-04 5.14E-04 5.71E-04 6.86E-04

ET1 3.22E-03 6.44E-03 9.66E-03 1.29E-02 1.61E-02 1.93E-02 2.25E-02 2.58E-02 2.90E-02 3.22E-02 3.87E-02

ET2 5.39E-05 1.08E-04 1.62E-04 2.16E-04 2.70E-04 3.24E-04 3.78E-04 4.32E-04 4.86E-04 5.40E-04 6.48E-04

Lung 4.95E-05 9.89E-05 1.48E-04 1.98E-04 2.47E-04 2.97E-04 3.46E-04 3.96E-04 4.45E-04 4.95E-04 5.94E-04

LN(ET) 4.65E-05 9.29E-05 1.39E-04 1.86E-04 2.32E-04 2.79E-04 3.25E-04 3.72E-04 4.18E-04 4.65E-04 5.58E-04

BBsec 1.26E-04 2.51E-04 3.77E-04 5.03E-04 6.29E-04 7.54E-04 8.80E-04 1.01E-03 1.13E-03 1.26E-03 1.51E-03

BBbas 6.37E-05 1.27E-04 1.91E-04 2.55E-04 3.19E-04 3.82E-04 4.46E-04 5.10E-04 5.74E-04 6.37E-04 7.65E-04

bb 3.91E-05 7.83E-05 1.17E-04 1.57E-04 1.96E-04 2.35E-04 2.74E-04 3.13E-04 3.52E-04 3.92E-04 4.70E-04

AI 1.47E-05 2.94E-05 4.41E-05 5.88E-05 7.35E-05 8.81E-05 1.03E-04 1.18E-04 1.32E-04 1.47E-04 1.76E-04

LN(TH) 8.11E-06 1.62E-05 2.43E-05 3.25E-05 4.06E-05 4.86E-05 5.68E-05 6.49E-05 7.30E-05 8.11E-05 9.74E-05

Esophagus 2.38E-06 4.76E-06 7.14E-06 9.52E-06 1.19E-05 1.43E-05 1.67E-05 1.90E-05 2.14E-05 2.38E-05 2.86E-05

Gonads 3.05E-06 6.10E-06 9.15E-06 1.22E-05 1.53E-05 1.83E-05 2.14E-05 2.44E-05 2.75E-05 3.05E-05 3.66E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 225 of 248

Table D-21 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 3.47E-05 3.97E-05 4.46E-05 4.96E-05 6.20E-05 7.43E-05 8.67E-05 9.91E-05 1.12E-04 1.24E-04

Urinary Bladder 1.67E-05 1.91E-05 2.15E-05 2.39E-05 2.98E-05 3.58E-05 4.17E-05 4.77E-05 5.37E-05 5.96E-05

Brain 8.76E-06 1.00E-05 1.13E-05 1.25E-05 1.56E-05 1.88E-05 2.19E-05 2.50E-05 2.82E-05 3.13E-05

Breast 2.70E-05 3.09E-05 3.48E-05 3.86E-05 4.83E-05 5.79E-05 6.76E-05 7.72E-05 8.69E-05 9.66E-05

Gall Bladder 3.00E-05 3.43E-05 3.86E-05 4.28E-05 5.36E-05 6.43E-05 7.50E-05 8.57E-05 9.64E-05 1.07E-04

Heart Wall 4.51E-05 5.16E-05 5.80E-05 6.45E-05 8.06E-05 9.67E-05 1.13E-04 1.29E-04 1.45E-04 1.61E-04

Kidneys 3.38E-05 3.86E-05 4.35E-05 4.83E-05 6.04E-05 7.25E-05 8.45E-05 9.66E-05 1.09E-04 1.21E-04

Liver 5.16E-05 5.90E-05 6.63E-05 7.37E-05 9.21E-05 1.11E-04 1.29E-04 1.47E-04 1.66E-04 1.84E-04

Muscle 2.01E-05 2.29E-05 2.58E-05 2.87E-05 3.58E-05 4.30E-05 5.02E-05 5.73E-05 6.45E-05 7.17E-05

Ovaries 4.27E-05 4.88E-05 5.49E-05 6.10E-05 7.63E-05 9.16E-05 1.07E-04 1.22E-04 1.37E-04 1.53E-04

Pancreas 2.87E-05 3.28E-05 3.69E-05 4.10E-05 5.12E-05 6.15E-05 7.17E-05 8.19E-05 9.22E-05 1.02E-04

Testes 7.43E-06 8.49E-06 9.55E-06 1.06E-05 1.33E-05 1.59E-05 1.86E-05 2.12E-05 2.39E-05 2.65E-05

Thyroid 1.58E-05 1.80E-05 2.03E-05 2.25E-05 2.82E-05 3.38E-05 3.95E-05 4.51E-05 5.07E-05 5.64E-05

R.B.M. 3.44E-05 3.93E-05 4.43E-05 4.92E-05 6.15E-05 7.38E-05 8.61E-05 9.84E-05 1.11E-04 1.23E-04

Bone Surface 7.70E-05 8.81E-05 9.90E-05 1.10E-04 1.38E-04 1.65E-04 1.93E-04 2.20E-04 2.48E-04 2.75E-04

Stomach 2.84E-05 3.24E-05 3.65E-05 4.06E-05 5.07E-05 6.08E-05 7.10E-05 8.11E-05 9.12E-05 1.01E-04

S.I. 4.43E-05 5.07E-05 5.70E-05 6.33E-05 7.91E-05 9.50E-05 1.11E-04 1.27E-04 1.42E-04 1.58E-04

U.L.I. 8.86E-05 1.01E-04 1.14E-04 1.27E-04 1.58E-04 1.90E-04 2.22E-04 2.53E-042.85E-04 3.16E-04

L.L.I. 1.81E-04 2.07E-04 2.33E-04 2.59E-04 3.24E-04 3.89E-04 4.53E-04 5.18E-04 5.83E-04 6.48E-04

Skin 1.07E-05 1.22E-05 1.38E-05 1.53E-05 1.91E-05 2.30E-05 2.68E-05 3.06E-05 3.44E-05 3.83E-05

Spleen 2.44E-05 2.78E-05 3.13E-05 3.48E-054.35E-05 5.22E-05 6.09E-05 6.96E-05 7.83E-05 8.70E-05

Thymus 3.33E-05 3.81E-05 4.28E-05 4.76E-05 5.95E-05 7.14E-05 8.33E-05 9.52E-05 1.07E-04 1.19E-04

Uterus 2.23E-05 2.54E-05 2.86E-05 3.18E-05 3.97E-05 4.77E-05 5.57E-05 6.36E-05 7.15E-05 7.95E-05

Colon 1.29E-04 1.47E-04 1.65E-04 1.84E-04 2.30E-04 2.76E-04 3.22E-04 3.68E-04 4.14E-04 4.60E-04

ET 8.00E-04 9.14E-04 1.03E-03 1.14E-03 1.43E-03 1.71E-03 2.00E-03 2.29E-03 2.57E-03 2.86E-03

ET1 4.51E-02 5.16E-02 5.80E-02 6.44E-02 8.05E-02 9.67E-02 1.13E-01 1.29E-01 1.45E-01 1.61E-01

ET2 7.56E-04 8.64E-04 9.71E-04 1.08E-03 1.35E-03 1.62E-03 1.89E-03 2.16E-03 2.43E-03 2.70E-03

Lung 6.93E-04 7.92E-04 8.91E-04 9.90E-04 1.24E-03 1.48E-03 1.73E-03 1.98E-03 2.23E-03 2.47E-03

LN(ET) 6.51E-04 7.44E-04 8.37E-04 9.30E-04 1.16E-03 1.40E-03 1.63E-03 1.86E-03 2.09E-03 2.33E-03

BBsec 1.76E-03 2.01E-03 2.26E-03 2.51E-03 3.14E-03 3.77E-03 4.40E-03 5.03E-03 5.66E-03 6.29E-03

BBbas 8.92E-04 1.02E-03 1.15E-03 1.27E-03 1.59E-03 1.91E-03 2.23E-03 2.55E-03 2.87E-03 3.19E-03

bb 5.48E-04 6.27E-04 7.05E-04 7.83E-04 9.79E-04 1.18E-03 1.37E-03 1.57E-03 1.76E-03 1.96E-03

AI 2.06E-04 2.35E-04 2.65E-04 2.94E-04 3.67E-04 4.41E-04 5.14E-04 5.88E-04 6.61E-04 7.35E-04

LN(TH) 1.14E-04 1.30E-04 1.46E-04 1.62E-04 2.03E-04 2.43E-04 2.84E-04 3.25E-04 3.65E-04 4.06E-04

Esophagus 3.33E-05 3.81E-05 4.28E-05 4.76E-05 5.95E-05 7.14E-05 8.33E-05 9.52E-05 1.07E-04 1.19E-04

Gonads 4.27E-05 4.88E-05 5.49E-05 6.10E-05 7.63E-05 9.16E-05 1.07E-04 1.22E-04 1.37E-04 1.53E-04
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 226 of 248

If either 95Zr or 95Nb is used as the radionuclide to maximize the dose to the organ of interest, then assume an equal intake of the 

other radionuclide to also have occurred.

Table D-22. Nb-95, Type W, 50-year equivalent dose for years of chronic intake (Sv) MDA = 2.2 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 2.48E-06 4.95E-06 7.43E-06 9.91E-06 1.24E-05 1.49E-05 1.73E-05 1.98E-05 2.23E-05 2.48E-05 2.97E-05

Urinary Bladder 1.19E-06 2.38E-06 3.58E-06 4.77E-06 5.96E-06 7.15E-06 8.35E-06 9.54E-06 1.07E-05 1.19E-05 1.43E-05

Brain 6.25E-07 1.25E-06 1.88E-06 2.50E-06 3.13E-06 3.75E-06 4.38E-06 5.01E-06 5.63E-06 6.26E-06 7.51E-06

Breast 1.93E-06 3.86E-06 5.79E-06 7.72E-06 9.66E-06 1.16E-05 1.35E-05 1.55E-05 1.74E-05 1.93E-05 2.32E-05

Gall Bladder 2.14E-06 4.28E-06 6.42E-06 8.57E-06 1.07E-05 1.28E-05 1.50E-05 1.71E-05 1.93E-05 2.14E-05 2.57E-05

Heart Wall 3.22E-06 6.44E-06 9.67E-06 1.29E-05 1.61E-05 1.93E-05 2.26E-05 2.58E-05 2.90E-05 3.22E-05 3.87E-05

Kidneys 2.41E-06 4.83E-06 7.24E-06 9.66E-06 1.21E-05 1.45E-05 1.69E-05 1.93E-05 2.17E-05 2.41E-05 2.90E-05

Liver 3.68E-06 7.36E-06 1.10E-05 1.47E-05 1.84E-05 2.21E-05 2.58E-05 2.95E-05 3.32E-05 3.68E-05 4.42E-05

Muscle 1.43E-06 2.87E-06 4.30E-06 5.73E-06 7.17E-06 8.60E-06 1.00E-05 1.15E-05 1.29E-05 1.43E-05 1.72E-05

Ovaries 3.05E-06 6.10E-06 9.15E-06 1.22E-05 1.53E-05 1.83E-05 2.14E-05 2.44E-05 2.75E-05 3.05E-05 3.66E-05

Pancreas 2.05E-06 4.09E-06 6.14E-06 8.19E-06 1.02E-05 1.23E-05 1.43E-05 1.64E-05 1.84E-05 2.05E-05 2.46E-05

Testes 5.30E-07 1.06E-06 1.59E-06 2.12E-06 2.65E-06 3.18E-06 3.71E-06 4.24E-06 4.78E-06 5.31E-06 6.37E-06

Thyroid 1.13E-06 2.25E-06 3.38E-06 4.51E-06 5.64E-06 6.76E-06 7.89E-06 9.02E-06 1.01E-05 1.13E-05 1.35E-05

R.B.M. 2.46E-06 4.91E-06 7.37E-06 9.84E-06 1.23E-05 1.47E-05 1.72E-05 1.97E-05 2.21E-05 2.46E-05 2.95E-05

Bone Surface 5.50E-06 1.10E-05 1.65E-05 2.20E-05 2.75E-05 3.30E-05 3.85E-05 4.40E-05 4.95E-05 5.50E-05 6.60E-05

Stomach 2.03E-06 4.05E-06 6.08E-06 8.11E-06 1.01E-05 1.22E-05 1.42E-05 1.62E-05 1.82E-05 2.03E-05 2.43E-05

S.I. 3.16E-06 6.33E-06 9.49E-06 1.27E-05 1.58E-05 1.90E-05 2.22E-05 2.53E-05 2.85E-05 3.17E-05 3.80E-05

U.L.I. 6.32E-06 1.26E-05 1.90E-05 2.53E-05 3.16E-05 3.79E-05 4.43E-05 5.06E-05 5.70E-05 6.33E-05 7.59E-05

L.L.I. 1.29E-05 2.59E-05 3.88E-05 5.18E-05 6.48E-05 7.77E-05 9.06E-05 1.04E-04 1.17E-04 1.30E-04 1.55E-04

Skin 7.65E-07 1.53E-06 2.29E-06 3.06E-06 3.83E-06 4.59E-06 5.36E-06 6.12E-06 6.89E-06 7.65E-06 9.18E-06

Spleen 1.74E-06 3.48E-06 5.22E-06 6.96E-06 8.70E-06 1.04E-05 1.22E-05 1.39E-05 1.57E-05 1.74E-05 2.09E-05

Thymus 2.38E-06 4.76E-06 7.14E-06 9.52E-06 1.19E-05 1.43E-05 1.67E-05 1.90E-05 2.14E-05 2.38E-05 2.86E-05

Uterus 1.59E-06 3.18E-06 4.77E-06 6.36E-06 7.95E-06 9.53E-06 1.11E-05 1.27E-05 1.43E-05 1.59E-05 1.91E-05

Colon 9.19E-06 1.84E-05 2.76E-05 3.68E-05 4.60E-05 5.51E-05 6.43E-05 7.35E-05 8.27E-05 9.19E-05 1.10E-04

ET 5.71E-05 1.14E-04 1.71E-04 2.29E-04 2.86E-04 3.43E-04 4.00E-04 4.57E-04 5.14E-04 5.71E-04 6.86E-04

ET1 3.22E-03 6.44E-03 9.66E-03 1.29E-02 1.61E-02 1.93E-02 2.25E-02 2.58E-02 2.90E-02 3.22E-02 3.87E-02

ET2 5.39E-05 1.08E-04 1.62E-04 2.16E-04 2.70E-04 3.24E-04 3.78E-04 4.32E-04 4.86E-04 5.40E-04 6.48E-04

Lung 4.95E-05 9.89E-05 1.48E-04 1.98E-04 2.47E-04 2.97E-04 3.46E-04 3.96E-04 4.45E-04 4.95E-04 5.94E-04

LN(ET) 4.65E-05 9.29E-05 1.39E-04 1.86E-04 2.32E-04 2.79E-04 3.25E-04 3.72E-04 4.18E-04 4.65E-04 5.58E-04

BBsec 1.26E-04 2.51E-04 3.77E-04 5.03E-04 6.29E-04 7.54E-04 8.80E-04 1.01E-03 1.13E-03 1.26E-03 1.51E-03

BBbas 6.37E-05 1.27E-04 1.91E-04 2.55E-04 3.19E-04 3.82E-04 4.46E-04 5.10E-04 5.74E-04 6.37E-04 7.65E-04

bb 3.91E-05 7.83E-05 1.17E-04 1.57E-04 1.96E-04 2.35E-04 2.74E-04 3.13E-04 3.52E-04 3.92E-04 4.70E-04

AI 1.47E-05 2.94E-05 4.41E-05 5.88E-05 7.35E-05 8.81E-05 1.03E-04 1.18E-04 1.32E-04 1.47E-04 1.76E-04

LN(TH) 8.11E-06 1.62E-05 2.43E-05 3.25E-05 4.06E-05 4.86E-05 5.68E-05 6.49E-05 7.30E-05 8.11E-05 9.74E-05

Esophagus 2.38E-06 4.76E-06 7.14E-06 9.52E-06 1.19E-05 1.43E-05 1.67E-05 1.90E-05 2.14E-05 2.38E-05 2.86E-05

Gonads 3.05E-06 6.10E-06 9.15E-06 1.22E-05 1.53E-05 1.83E-05 2.14E-05 2.44E-05 2.75E-05 3.05E-05 3.66E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 227 of 248

Table D-22 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 3.47E-05 3.97E-05 4.46E-05 4.96E-05 6.20E-05 7.43E-05 8.67E-05 9.91E-05 1.12E-04 1.24E-04

Urinary Bladder 1.67E-05 1.91E-05 2.15E-05 2.39E-05 2.98E-05 3.58E-05 4.17E-05 4.77E-05 5.37E-05 5.96E-05

Brain 8.76E-06 1.00E-05 1.13E-05 1.25E-05 1.56E-05 1.88E-05 2.19E-05 2.50E-05 2.82E-05 3.13E-05

Breast 2.70E-05 3.09E-05 3.48E-05 3.86E-05 4.83E-05 5.79E-05 6.76E-05 7.72E-05 8.69E-05 9.66E-05

Gall Bladder 3.00E-05 3.43E-05 3.86E-05 4.28E-05 5.36E-05 6.43E-05 7.50E-05 8.57E-05 9.64E-05 1.07E-04

Heart Wall 4.51E-05 5.16E-05 5.80E-05 6.45E-05 8.06E-05 9.67E-05 1.13E-04 1.29E-04 1.45E-04 1.61E-04

Kidneys 3.38E-05 3.86E-05 4.35E-05 4.83E-05 6.04E-05 7.25E-05 8.45E-05 9.66E-05 1.09E-04 1.21E-04

Liver 5.16E-05 5.90E-05 6.63E-05 7.37E-05 9.21E-05 1.11E-04 1.29E-04 1.47E-04 1.66E-04 1.84E-04

Muscle 2.01E-05 2.29E-05 2.58E-05 2.87E-05 3.58E-05 4.30E-05 5.02E-05 5.73E-05 6.45E-05 7.17E-05

Ovaries 4.27E-05 4.88E-05 5.49E-05 6.10E-05 7.63E-05 9.16E-05 1.07E-04 1.22E-04 1.37E-04 1.53E-04

Pancreas 2.87E-05 3.28E-05 3.69E-05 4.10E-05 5.12E-05 6.15E-05 7.17E-05 8.19E-05 9.22E-05 1.02E-04

Testes 7.43E-06 8.49E-06 9.55E-06 1.06E-05 1.33E-05 1.59E-05 1.86E-05 2.12E-05 2.39E-05 2.65E-05

Thyroid 1.58E-05 1.80E-05 2.03E-05 2.25E-05 2.82E-05 3.38E-05 3.95E-05 4.51E-05 5.07E-05 5.64E-05

R.B.M. 3.44E-05 3.93E-05 4.43E-05 4.92E-05 6.15E-05 7.38E-05 8.61E-05 9.84E-05 1.11E-04 1.23E-04

Bone Surface 7.70E-05 8.81E-05 9.90E-05 1.10E-04 1.38E-04 1.65E-04 1.93E-04 2.20E-04 2.48E-04 2.75E-04

Stomach 2.84E-05 3.24E-05 3.65E-05 4.06E-05 5.07E-05 6.08E-05 7.10E-05 8.11E-05 9.12E-05 1.01E-04

S.I. 4.43E-05 5.07E-05 5.70E-05 6.33E-05 7.91E-05 9.50E-05 1.11E-04 1.27E-04 1.42E-04 1.58E-04

U.L.I. 8.86E-05 1.01E-04 1.14E-04 1.27E-04 1.58E-04 1.90E-04 2.22E-04 2.53E-04 2.85E-04 3.16E-04

L.L.I. 1.81E-04 2.07E-04 2.33E-04 2.59E-04 3.24E-04 3.89E-04 4.53E-04 5.18E-04 5.83E-04 6.48E-04

Skin 1.07E-05 1.22E-05 1.38E-05 1.53E-05 1.91E-05 2.30E-05 2.68E-05 3.06E-05 3.44E-05 3.83E-05

Spleen 2.44E-05 2.78E-05 3.13E-05 3.48E-05 4.35E-05 5.22E-05 6.09E-05 6.96E-05 7.83E-05 8.70E-05

Thymus 3.33E-05 3.81E-05 4.28E-05 4.76E-05 5.95E-05 7.14E-05 8.33E-05 9.52E-05 1.07E-04 1.19E-04

Uterus 2.23E-05 2.54E-05 2.86E-05 3.18E-05 3.97E-05 4.77E-05 5.57E-05 6.36E-05 7.15E-05 7.95E-05

Colon 1.29E-04 1.47E-04 1.65E-04 1.84E-04 2.30E-04 2.76E-04 3.22E-04 3.68E-04 4.14E-04 4.60E-04

ET 8.00E-04 9.14E-04 1.03E-03 1.14E-03 1.43E-03 1.71E-03 2.00E-03 2.29E-03 2.57E-03 2.86E-03

ET1 4.51E-02 5.16E-02 5.80E-02 6.44E-02 8.05E-02 9.67E-02 1.13E-01 1.29E-01 1.45E-01 1.61E-01

ET2 7.56E-04 8.64E-04 9.71E-04 1.08E-03 1.35E-03 1.62E-03 1.89E-03 2.16E-03 2.43E-03 2.70E-03

Lung 6.93E-04 7.92E-04 8.91E-04 9.90E-04 1.24E-03 1.48E-03 1.73E-03 1.98E-03 2.23E-03 2.47E-03

LN(ET) 6.51E-04 7.44E-04 8.37E-04 9.30E-04 1.16E-03 1.40E-031.63E-03 1.86E-03 2.09E-03 2.33E-03

BBsec 1.76E-03 2.01E-03 2.26E-03 2.51E-03 3.14E-03 3.77E-03 4.40E-03 5.03E-03 5.66E-03 6.29E-03

BBbas 8.92E-04 1.02E-03 1.15E-03 1.27E-03 1.59E-03 1.91E-03 2.23E-03 2.55E-03 2.87E-03 3.19E-03

bb 5.48E-04 6.27E-04 7.05E-04 7.83E-04 9.79E-04 1.18E-03 1.37E-03 1.57E-03 1.76E-03 1.96E-03

AI 2.06E-04 2.35E-04 2.65E-04 2.94E-04 3.67E-04 4.41E-04 5.14E-04 5.88E-04 6.61E-04 7.35E-04

LN(TH) 1.14E-04 1.30E-04 1.46E-04 1.62E-04 2.03E-04 2.43E-04 2.84E-04 3.25E-04 3.65E-04 4.06E-04

Esophagus 3.33E-05 3.81E-05 4.28E-05 4.76E-05 5.95E-05 7.14E-05 8.33E-05 9.52E-05 1.07E-04 1.19E-04

Gonads 4.27E-05 4.88E-05 5.49E-05 6.10E-05 7.63E-05 9.16E-05 1.07E-04 1.22E-04 1.37E-04 1.53E-04
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 228 of 248

If either 95Zr or 95Nb is used as the radionuclide to maximize the dose to the organ of interest, then assume an equal intake of the 

other radionuclide to also have occurred.

Table D-23. Nb-95, Type S, 50-year equivalent dose for years of chronic intake (Sv) (MDA = 2.2 nCi). Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 2.37E-06 4.73E-06 7.10E-06 9.47E-06 1.18E-05 1.42E-05 1.66E-05 1.89E-05 2.13E-05 2.37E-05 2.84E-05

Urinary Bladder 9.59E-07 1.92E-06 2.88E-06 3.84E-06 4.80E-06 5.75E-06 6.71E-06 7.68E-06 8.63E-06 9.59E-06 1.15E-05

Brain 3.90E-07 7.80E-07 1.17E-06 1.56E-06 1.95E-06 2.34E-06 2.73E-06 3.12E-06 3.51E-06 3.90E-06 4.68E-06

Breast 2.36E-06 4.71E-06 7.07E-06 9.43E-06 1.18E-05 1.41E-05 1.65E-05 1.89E-05 2.12E-05 2.36E-05 2.83E-05

Gall Bladder 1.63E-06 3.25E-06 4.88E-06 6.51E-06 8.14E-06 9.76E-06 1.14E-05 1.30E-05 1.46E-05 1.63E-05 1.95E-05

Heart Wall 3.94E-06 7.88E-06 1.18E-05 1.58E-05 1.97E-05 2.36E-05 2.76E-05 3.15E-05 3.55E-05 3.94E-05 4.73E-05

Kidneys 1.30E-06 2.60E-06 3.91E-06 5.21E-06 6.52E-06 7.81E-06 9.12E-06 1.04E-05 1.17E-05 1.30E-05 1.56E-05

Liver 2.30E-06 4.60E-06 6.90E-06 9.21E-06 1.15E-05 1.38E-05 1.61E-05 1.84E-05 2.07E-05 2.30E-05 2.76E-05

Muscle 1.41E-06 2.81E-06 4.21E-06 5.62E-06 7.03E-06 8.43E-06 9.84E-06 1.12E-05 1.26E-05 1.41E-05 1.69E-05

Ovaries 2.99E-06 5.98E-06 8.97E-06 1.20E-05 1.50E-05 1.79E-05 2.09E-05 2.39E-05 2.69E-05 2.99E-05 3.59E-05

Pancreas 1.94E-06 3.87E-06 5.81E-06 7.75E-06 9.69E-06 1.16E-05 1.36E-05 1.55E-05 1.74E-05 1.94E-05 2.33E-05

Testes 3.32E-07 6.64E-07 9.96E-07 1.33E-06 1.66E-06 1.99E-06 2.33E-06 2.66E-06 2.99E-06 3.32E-06 3.99E-06

Thyroid 1.10E-06 2.20E-06 3.29E-06 4.40E-06 5.49E-06 6.59E-06 7.69E-06 8.79E-06 9.89E-06 1.10E-05 1.32E-05

R.B.M. 1.88E-06 3.76E-06 5.64E-06 7.52E-06 9.40E-06 1.13E-05 1.32E-05 1.50E-05 1.69E-05 1.88E-05 2.26E-05

Bone Surface 1.77E-06 3.54E-06 5.31E-06 7.08E-06 8.85E-06 1.06E-05 1.24E-05 1.42E-05 1.59E-05 1.77E-05 2.12E-05

Stomach 2.09E-06 4.17E-06 6.26E-06 8.35E-06 1.04E-05 1.25E-05 1.46E-05 1.67E-05 1.88E-05 2.09E-05 2.51E-05

S.I. 3.12E-06 6.23E-06 9.35E-06 1.25E-05 1.56E-05 1.87E-05 2.18E-05 2.50E-05 2.81E-05 3.12E-05 3.74E-05

U.L.I. 6.70E-06 1.34E-05 2.01E-05 2.68E-05 3.35E-05 4.02E-05 4.69E-05 5.36E-05 6.03E-05 6.70E-05 8.04E-05

L.L.I. 1.42E-05 2.85E-05 4.27E-05 5.70E-05 7.12E-05 8.54E-05 9.97E-05 1.14E-04 1.28E-04 1.42E-04 1.71E-04

Skin 6.89E-07 1.38E-06 2.07E-06 2.76E-06 3.45E-06 4.13E-06 4.83E-06 5.52E-06 6.21E-06 6.89E-06 8.28E-06

Spleen 1.82E-06 3.65E-06 5.47E-06 7.30E-06 9.12E-06 1.09E-05 1.28E-05 1.46E-05 1.64E-05 1.82E-05 2.19E-05

Thymus 2.86E-06 5.72E-06 8.58E-06 1.15E-05 1.43E-05 1.72E-05 2.00E-05 2.29E-05 2.58E-05 2.86E-05 3.44E-05

Uterus 1.39E-06 2.77E-06 4.16E-06 5.54E-06 6.93E-06 8.31E-06 9.70E-06 1.11E-05 1.25E-05 1.39E-05 1.66E-05

Colon 9.96E-06 1.99E-05 2.99E-05 3.98E-05 4.98E-05 5.97E-05 6.97E-05 7.97E-05 8.97E-05 9.96E-05 1.20E-04

ET 6.46E-05 1.29E-04 1.94E-04 2.58E-04 3.23E-04 3.87E-04 4.52E-04 5.17E-04 5.81E-04 6.46E-04 7.75E-04

ET1 3.47E-03 6.94E-03 1.04E-02 1.39E-02 1.74E-02 2.08E-02 2.43E-02 2.78E-02 3.12E-02 3.47E-02 4.17E-02

ET2 6.12E-05 1.22E-04 1.84E-04 2.45E-04 3.06E-04 3.67E-04 4.28E-04 4.90E-04 5.51E-04 6.12E-04 7.34E-04

Lung 6.57E-05 1.31E-04 1.97E-04 2.63E-04 3.29E-04 3.94E-04 4.60E-04 5.26E-04 5.91E-04 6.57E-04 7.89E-04

LN(ET) 5.05E-05 1.01E-04 1.51E-04 2.02E-04 2.52E-04 3.03E-04 3.53E-04 4.04E-04 4.54E-04 5.05E-04 6.06E-04

BBsec 1.65E-04 3.31E-04 4.96E-04 6.62E-04 8.27E-04 9.92E-04 1.16E-03 1.32E-03 1.49E-03 1.65E-03 1.99E-03

BBbas 8.42E-05 1.68E-04 2.52E-04 3.37E-04 4.21E-04 5.05E-04 5.89E-04 6.74E-04 7.58E-04 8.42E-04 1.01E-03

bb 5.19E-05 1.04E-04 1.56E-04 2.08E-04 2.60E-04 3.11E-04 3.63E-04 4.15E-04 4.67E-04 5.19E-04 6.23E-04

AI 2.06E-05 4.11E-05 6.17E-05 8.23E-05 1.03E-04 1.23E-04 1.44E-04 1.65E-04 1.85E-04 2.06E-04 2.47E-04

LN(TH) 1.11E-05 2.22E-05 3.33E-05 4.44E-05 5.55E-05 6.66E-05 7.77E-05 8.88E-05 9.99E-05 1.11E-04 1.33E-04

Esophagus 2.86E-06 5.72E-06 8.58E-06 1.15E-05 1.43E-05 1.72E-05 2.00E-05 2.29E-05 2.58E-05 2.86E-05 3.44E-05

Gonads 2.99E-06 5.98E-06 8.97E-06 1.20E-05 1.50E-05 1.79E-05 2.09E-05 2.39E-05 2.69E-05 2.99E-05 3.59E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 229 of 248

Table D-23 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 3.31E-05 3.79E-05 4.26E-05 4.73E-05 5.92E-05 7.10E-05 8.28E-05 9.47E-05 1.07E-04 1.18E-04

Urinary Bladder 1.34E-05 1.54E-05 1.73E-05 1.92E-05 2.40E-05 2.88E-05 3.36E-05 3.84E-05 4.32E-05 4.80E-05

Brain 5.46E-06 6.24E-06 7.02E-06 7.80E-06 9.75E-06 1.17E-05 1.37E-05 1.56E-05 1.76E-05 1.95E-05

Breast 3.30E-05 3.77E-05 4.24E-05 4.71E-05 5.89E-05 7.07E-05 8.25E-05 9.43E-05 1.06E-04 1.18E-04

Gall Bladder 2.28E-05 2.60E-05 2.93E-05 3.25E-05 4.07E-05 4.88E-05 5.70E-05 6.51E-05 7.32E-05 8.14E-05

Heart Wall 5.52E-05 6.31E-05 7.09E-05 7.88E-05 9.85E-05 1.18E-04 1.38E-04 1.58E-04 1.77E-04 1.97E-04

Kidneys 1.82E-05 2.09E-05 2.35E-05 2.61E-05 3.26E-05 3.91E-05 4.56E-05 5.21E-05 5.87E-05 6.52E-05

Liver 3.22E-05 3.68E-05 4.14E-05 4.60E-05 5.76E-05 6.91E-05 8.06E-05 9.21E-05 1.04E-04 1.15E-04

Muscle 1.97E-05 2.25E-05 2.53E-05 2.81E-05 3.51E-05 4.22E-05 4.92E-05 5.62E-05 6.32E-05 7.03E-05

Ovaries 4.19E-05 4.79E-05 5.39E-05 5.99E-05 7.48E-05 8.98E-05 1.05E-04 1.20E-04 1.35E-04 1.50E-04

Pancreas 2.71E-05 3.10E-05 3.49E-05 3.88E-05 4.85E-05 5.82E-05 6.79E-05 7.75E-05 8.72E-05 9.69E-05

Testes 4.65E-06 5.32E-06 5.98E-06 6.65E-06 8.31E-06 9.97E-06 1.16E-05 1.33E-05 1.50E-05 1.66E-05

Thyroid 1.54E-05 1.76E-05 1.98E-05 2.20E-05 2.75E-05 3.30E-05 3.85E-05 4.40E-05 4.94E-05 5.49E-05

R.B.M. 2.63E-05 3.01E-05 3.39E-05 3.76E-05 4.70E-05 5.64E-05 6.58E-05 7.52E-05 8.46E-05 9.40E-05

Bone Surface 2.48E-05 2.83E-05 3.19E-05 3.54E-05 4.42E-05 5.31E-05 6.19E-05 7.08E-05 7.96E-05 8.85E-05

Stomach 2.92E-05 3.34E-05 3.76E-05 4.18E-05 5.22E-05 6.27E-05 7.31E-05 8.35E-05 9.40E-05 1.04E-04

S.I. 4.37E-05 4.99E-05 5.61E-05 6.24E-05 7.80E-05 9.36E-05 1.09E-04 1.25E-04 1.40E-04 1.56E-04

U.L.I. 9.38E-05 1.07E-04 1.21E-04 1.34E-04 1.68E-04 2.01E-04 2.35E-04 2.68E-04 3.02E-04 3.35E-04

L.L.I. 1.99E-04 2.28E-04 2.56E-04 2.85E-04 3.56E-04 4.27E-04 4.98E-04 5.70E-04 6.41E-04 7.12E-04

Skin 9.65E-06 1.10E-05 1.24E-05 1.38E-05 1.72E-05 2.07E-05 2.41E-05 2.76E-05 3.10E-05 3.45E-05

Spleen 2.56E-05 2.92E-05 3.29E-05 3.65E-05 4.56E-05 5.48E-05 6.39E-05 7.30E-05 8.21E-05 9.13E-05

Thymus 4.01E-05 4.58E-05 5.15E-05 5.73E-05 7.16E-05 8.59E-05 1.00E-04 1.15E-04 1.29E-04 1.43E-04

Uterus 1.94E-05 2.22E-05 2.50E-05 2.77E-05 3.47E-05 4.16E-05 4.85E-05 5.54E-05 6.24E-05 6.93E-05

Colon 1.39E-04 1.59E-04 1.79E-04 1.99E-04 2.49E-04 2.99E-04 3.49E-04 3.98E-04 4.48E-04 4.98E-04

ET 9.04E-04 1.03E-03 1.16E-03 1.29E-03 1.62E-03 1.94E-03 2.26E-03 2.58E-03 2.91E-03 3.23E-03

ET1 4.86E-02 5.56E-02 6.25E-02 6.94E-02 8.68E-02 1.04E-01 1.22E-01 1.39E-01 1.56E-01 1.74E-01

ET2 8.57E-04 9.79E-04 1.10E-03 1.22E-03 1.53E-03 1.84E-03 2.14E-03 2.45E-03 2.75E-03 3.06E-03

Lung 9.20E-04 1.05E-03 1.18E-03 1.31E-03 1.64E-03 1.97E-03 2.30E-03 2.63E-03 2.96E-03 3.29E-03

LN(ET) 7.07E-04 8.08E-04 9.09E-04 1.01E-03 1.26E-03 1.51E-03 1.77E-03 2.02E-032.27E-03 2.52E-03

BBsec 2.32E-03 2.65E-03 2.98E-03 3.31E-03 4.14E-03 4.96E-03 5.79E-03 6.62E-03 7.44E-03 8.27E-03

BBbas 1.18E-03 1.35E-03 1.52E-03 1.68E-03 2.11E-03 2.53E-03 2.95E-03 3.37E-03 3.79E-03 4.21E-03

bb 7.27E-04 8.31E-04 9.35E-04 1.04E-03 1.30E-03 1.56E-03 1.82E-03 2.08E-03 2.34E-03 2.60E-03

AI 2.88E-04 3.29E-04 3.70E-04 4.11E-04 5.14E-04 6.17E-04 7.20E-04 8.23E-04 9.25E-04 1.03E-03

LN(TH) 1.55E-04 1.78E-04 2.00E-04 2.22E-04 2.78E-04 3.33E-04 3.89E-04 4.44E-04 5.00E-04 5.55E-04

Esophagus 4.01E-05 4.58E-05 5.15E-05 5.73E-05 7.16E-05 8.59E-05 1.00E-04 1.15E-04 1.29E-04 1.43E-04

Gonads 4.19E-05 4.79E-05 5.39E-05 5.99E-05 7.48E-05 8.98E-05 1.05E-04 1.20E-04 1.35E-04 1.50E-04
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 230 of 248

Table D-24. Ru-106, Type F, f1=0.05, 50-year equivalent dose for years of chronic intake (Sv) MDA = 1 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 5.76E-06 9.06E-06 1.25E-05 1.62E-05 2.01E-05 2.40E-05 2.79E-05 3.19E-05 3.59E-05 3.59E-05 4.78E-05

Urinary Bladder 6.63E-06 1.04E-05 1.44E-05 1.87E-05 2.31E-05 2.76E-05 3.22E-05 3.68E-05 4.13E-05 4.13E-05 5.51E-05

Brain 5.65E-06 8.89E-06 1.23E-05 1.59E-05 1.97E-05 2.35E-05 2.74E-05 3.13E-05 3.52E-05 3.52E-05 4.69E-05

Breast 5.62E-06 8.84E-06 1.22E-05 1.59E-05 1.96E-05 2.34E-05 2.73E-05 3.11E-05 3.50E-05 3.50E-05 4.67E-05

Gall Bladder 5.76E-06 9.06E-06 1.25E-05 1.62E-05 2.01E-05 2.40E-05 2.79E-05 3.19E-05 3.59E-05 3.59E-05 4.78E-05

Heart Wall 5.75E-06 9.05E-06 1.25E-05 1.62E-05 2.00E-05 2.39E-05 2.79E-05 3.19E-05 3.58E-05 3.58E-05 4.78E-05

Kidneys 5.74E-06 9.02E-06 1.25E-05 1.62E-05 2.00E-05 2.39E-05 2.78E-05 3.18E-05 3.57E-05 3.57E-05 4.76E-05

Liver 5.74E-06 9.03E-06 1.25E-05 1.62E-05 2.00E-05 2.39E-05 2.78E-05 3.18E-05 3.58E-05 3.58E-05 4.77E-05

Muscle 5.69E-06 8.95E-06 1.24E-05 1.61E-05 1.98E-05 2.37E-05 2.76E-05 3.15E-05 3.55E-05 3.55E-05 4.73E-05

Ovaries 5.80E-06 9.13E-06 1.26E-05 1.64E-05 2.02E-05 2.41E-05 2.81E-05 3.21E-05 3.61E-05 3.61E-05 4.82E-05

Pancreas 5.78E-06 9.09E-06 1.26E-05 1.63E-05 2.01E-05 2.41E-05 2.80E-05 3.20E-05 3.60E-05 3.60E-05 4.80E-05

Testes 5.69E-06 8.95E-06 1.24E-05 1.60E-05 1.98E-05 2.37E-05 2.76E-05 3.15E-05 3.54E-05 3.54E-05 4.73E-05

Thyroid 5.72E-06 8.99E-06 1.24E-05 1.61E-05 1.99E-05 2.38E-05 2.77E-05 3.17E-05 3.56E-05 3.56E-05 4.75E-05

R.B.M. 5.72E-06 9.00E-06 1.25E-05 1.61E-05 1.99E-05 2.38E-05 2.78E-05 3.17E-05 3.57E-05 3.57E-05 4.75E-05

Bone Surface 5.75E-06 9.04E-06 1.25E-05 1.62E-05 2.00E-05 2.39E-05 2.79E-05 3.18E-05 3.58E-05 3.58E-05 4.77E-05

Stomach 5.94E-06 9.35E-06 1.29E-05 1.68E-05 2.07E-05 2.47E-05 2.88E-05 3.29E-05 3.70E-05 3.70E-05 4.94E-05

S.I. 6.31E-06 9.93E-06 1.37E-05 1.78E-05 2.20E-05 2.63E-05 3.06E-05 3.50E-05 3.93E-05 3.93E-05 5.24E-05

U.L.I. 9.79E-06 1.54E-05 2.13E-05 2.76E-05 3.41E-05 4.07E-05 4.74E-05 5.42E-05 6.10E-05 6.10E-05 8.13E-05

L.L.I. 1.76E-05 2.76E-05 3.83E-05 4.96E-05 6.13E-05 7.32E-05 8.52E-05 9.74E-05 1.10E-04 1.10E-04 1.46E-04

Skin 5.60E-06 8.81E-06 1.22E-05 1.58E-05 1.95E-05 2.33E-05 2.72E-05 3.10E-05 3.49E-05 3.49E-05 4.65E-05

Spleen 5.74E-06 9.02E-06 1.25E-05 1.62E-05 2.00E-05 2.39E-05 2.78E-05 3.18E-05 3.57E-05 3.57E-05 4.76E-05

Thymus 5.72E-06 8.99E-06 1.25E-05 1.61E-05 1.99E-05 2.38E-05 2.77E-05 3.17E-05 3.56E-05 3.56E-05 4.75E-05

Uterus 5.79E-06 9.11E-06 1.26E-05 1.63E-05 2.02E-05 2.41E-05 2.81E-05 3.21E-05 3.61E-05 3.61E-05 4.81E-05

ET 8.13E-06 1.28E-05 1.77E-05 2.29E-05 2.83E-05 3.38E-05 3.94E-05 4.50E-05 5.06E-05 5.06E-05 6.75E-05

ET1 1.12E-03 1.76E-03 2.44E-03 3.16E-03 3.90E-03 4.66E-03 5.43E-03 6.20E-03 6.97E-03 6.97E-03 9.30E-03

ET2 7.02E-06 1.10E-05 1.53E-05 1.98E-05 2.45E-05 2.92E-05 3.40E-05 3.89E-05 4.37E-05 4.37E-05 5.83E-05

Lung 5.71E-06 8.99E-06 1.24E-05 1.61E-05 1.99E-05 2.38E-05 2.77E-05 3.17E-05 3.56E-05 3.56E-05 4.75E-05

Colon 1.32E-05 2.07E-05 2.86E-05 3.71E-05 4.58E-05 5.47E-05 6.38E-05 7.29E-05 8.20E-05 8.20E-05 1.09E-04

LN(ET) 6.81E-06 1.07E-05 1.48E-05 1.92E-05 2.37E-05 2.83E-05 3.30E-05 3.77E-05 4.24E-05 4.24E-05 5.65E-05

BBsec 5.74E-06 9.02E-06 1.25E-05 1.62E-05 2.00E-05 2.39E-05 2.78E-05 3.18E-05 3.57E-05 3.57E-05 4.77E-05

BBbas 5.74E-06 9.02E-06 1.25E-05 1.62E-05 2.00E-05 2.39E-05 2.78E-05 3.18E-05 3.57E-05 3.57E-05 4.76E-05

bb 5.71E-06 8.97E-06 1.24E-05 1.61E-05 1.99E-05 2.37E-05 2.77E-05 3.16E-05 3.55E-05 3.55E-05 4.74E-05

AI 5.70E-06 8.97E-06 1.24E-05 1.61E-05 1.99E-05 2.37E-05 2.76E-05 3.16E-05 3.55E-05 3.55E-05 4.74E-05

LN(TH) 5.70E-06 8.96E-06 1.24E-05 1.61E-05 1.98E-05 2.37E-05 2.76E-05 3.15E-05 3.55E-05 3.55E-05 4.73E-05

Esophagus 5.72E-06 8.99E-06 1.25E-05 1.61E-05 1.99E-05 2.38E-05 2.77E-05 3.17E-05 3.56E-05 3.56E-05 4.75E-05

Gonads 5.80E-06 9.13E-06 1.26E-05 1.64E-05 2.02E-05 2.41E-05 2.81E-05 3.21E-05 3.61E-05 3.61E-05 4.82E-05

Intake rate pCi/d 50.469 39.686 36.629 35.558 35.158 35.004 34.944 34.921 34.911 34.909 34.906

Days 365 730 1095 1461 1826 2190 2556 2922 3287 3652 4383

Intake (pCi) 1.84E+04 2.90E+04 4.01E+04 5.20E+04 6.42E+04 7.67E+04 8.93E+04 1.02E+05 1.15E+05 1.27E+05 1.53E+05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 231 of 248

Table D-24 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 5.58E-05 6.37E-05 7.16E-05 7.96E-05 9.95E-05 1.19E-04 1.39E-04 1.59E-04 1.79E-04 1.99E-04

Urinary Bladder 6.43E-05 7.36E-05 8.28E-05 9.20E-05 1.15E-04 1.38E-04 1.61E-04 1.84E-04 2.07E-04 2.30E-04

Brain 5.48E-05 6.27E-05 7.06E-05 7.84E-05 9.80E-05 1.18E-04 1.37E-04 1.57E-04 1.76E-04 1.96E-04

Breast 5.45E-05 6.24E-05 7.02E-05 7.80E-05 9.75E-05 1.17E-04 1.37E-04 1.56E-04 1.76E-04 1.95E-04

Gall Bladder 5.58E-05 6.37E-05 7.16E-05 7.96E-05 9.95E-05 1.19E-04 1.39E-04 1.59E-04 1.79E-04 1.99E-04

Heart Wall 5.57E-05 6.37E-05 7.16E-05 7.96E-05 9.95E-05 1.19E-04 1.39E-04 1.59E-04 1.79E-04 1.99E-04

Kidneys 5.56E-05 6.37E-05 7.16E-05 7.96E-05 9.95E-05 1.19E-04 1.39E-04 1.59E-041.79E-04 1.99E-04

Liver 5.56E-05 6.37E-05 7.16E-05 7.96E-05 9.95E-05 1.19E-04 1.39E-04 1.59E-04 1.79E-04 1.99E-04

Muscle 5.52E-05 6.30E-05 7.09E-05 7.88E-05 9.85E-05 1.18E-04 1.38E-04 1.58E-04 1.77E-04 1.97E-04

Ovaries 5.62E-05 6.43E-05 7.24E-05 8.04E-05 1.01E-04 1.21E-04 1.41E-04 1.61E-04 1.81E-04 2.01E-04

Pancreas 5.60E-05 6.40E-05 7.20E-05 8.00E-05 1.00E-04 1.20E-04 1.40E-04 1.60E-04 1.80E-04 2.00E-04

Testes 5.51E-05 6.30E-05 7.09E-05 7.88E-05 9.85E-05 1.18E-04 1.38E-04 1.58E-04 1.77E-04 1.97E-04

Thyroid 5.54E-05 6.34E-05 7.13E-05 7.92E-05 9.90E-05 1.19E-04 1.39E-04 1.58E-04 1.78E-04 1.98E-04

R.B.M. 5.55E-05 6.34E-05 7.13E-05 7.92E-05 9.90E-05 1.19E-04 1.39E-04 1.58E-04 1.78E-04 1.98E-04

Bone Surface 5.57E-05 6.37E-05 7.16E-05 7.96E-05 9.95E-05 1.19E-04 1.39E-04 1.59E-04 1.79E-04 1.99E-04

Stomach 5.76E-05 6.59E-05 7.42E-05 8.24E-05 1.03E-04 1.24E-04 1.44E-04 1.65E-04 1.85E-04 2.06E-04

S.I. 6.12E-05 6.98E-05 7.85E-05 8.72E-05 1.09E-04 1.31E-04 1.53E-04 1.74E-04 1.96E-04 2.18E-04

U.L.I. 9.48E-05 1.08E-04 1.22E-04 1.36E-04 1.70E-04 2.03E-04 2.37E-04 2.71E-04 3.05E-04 3.39E-04

L.L.I. 1.70E-04 1.95E-04 2.19E-04 2.43E-04 3.04E-04 3.65E-04 4.26E-04 4.86E-04 5.47E-04 6.08E-04

Skin 5.43E-05 6.21E-05 6.98E-05 7.76E-05 9.70E-05 1.16E-04 1.36E-04 1.55E-041.75E-04 1.94E-04

Spleen 5.56E-05 6.34E-05 7.13E-05 7.92E-05 9.90E-05 1.19E-04 1.39E-04 1.58E-04 1.78E-04 1.98E-04

Thymus 5.54E-05 6.34E-05 7.13E-05 7.92E-05 9.90E-05 1.19E-04 1.39E-04 1.58E-04 1.78E-04 1.98E-04

Uterus 5.61E-05 6.40E-05 7.20E-05 8.00E-05 1.00E-04 1.20E-04 1.40E-04 1.60E-04 1.80E-04 2.00E-04

ET 7.88E-05 8.99E-05 1.01E-04 1.12E-04 1.41E-04 1.69E-04 1.97E-04 2.25E-04 2.53E-04 2.81E-04

ET1 1.08E-02 1.24E-02 1.39E-02 1.55E-02 1.94E-02 2.32E-02 2.71E-02 3.10E-02 3.48E-02 3.87E-02

ET2 6.80E-05 7.78E-05 8.75E-05 9.72E-05 1.22E-04 1.46E-04 1.70E-04 1.94E-04 2.19E-04 2.43E-04

Lung 5.54E-05 6.34E-05 7.13E-05 7.92E-05 9.90E-05 1.19E-04 1.39E-04 1.58E-04 1.78E-04 1.98E-04

Colon 1.27E-04 1.46E-04 1.64E-04 1.82E-04 2.28E-04 2.73E-04 3.19E-04 3.64E-04 4.10E-04 4.55E-04

LN(ET) 6.60E-05 7.55E-05 8.50E-05 9.44E-05 1.18E-04 1.42E-04 1.65E-04 1.89E-04 2.12E-04 2.36E-04

BBsec 5.56E-05 6.37E-05 7.16E-05 7.96E-05 9.95E-05 1.19E-04 1.39E-04 1.59E-04 1.79E-04 1.99E-04

BBbas 5.56E-05 6.37E-05 7.16E-05 7.96E-05 9.95E-05 1.19E-04 1.39E-04 1.59E-04 1.79E-04 1.99E-04

bb 5.53E-05 6.30E-05 7.09E-05 7.88E-05 9.85E-05 1.18E-04 1.38E-04 1.58E-04 1.77E-04 1.97E-04

AI 5.52E-05 6.30E-05 7.09E-05 7.88E-05 9.85E-05 1.18E-04 1.38E-04 1.58E-04 1.77E-04 1.97E-04

LN(TH) 5.52E-05 6.30E-05 7.09E-05 7.88E-05 9.85E-05 1.18E-04 1.38E-04 1.58E-04 1.77E-04 1.97E-04

Esophagus 5.54E-05 6.34E-05 7.13E-05 7.92E-05 9.90E-05 1.19E-04 1.39E-04 1.58E-04 1.78E-04 1.98E-04

Gonads 5.62E-05 6.43E-05 7.24E-05 8.04E-05 1.01E-04 1.21E-04 1.41E-04 1.61E-04 1.81E-04 2.01E-04

Intake rate pCi/d 34.906 34.906 34.906 34.906 34.906 34.906 34.906 34.906 34.906 34.906

Days 5113 5844 6574 7305 9131 10958 12784 14610 16436 18263

Intake (pCi) 1.78E+05 2.04E+05 2.29E+05 2.55E+05 3.19E+05 3.82E+05 4.46E+05 5.10E+05 5.74E+05 6.37E+05
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 232 of 248

Table D-25. Ru-106, Type S, f1=0.05, 50-year equivalent dose for years of chronic intake (Sv) MDA = 1 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 1.25E-06 1.93E-06 2.67E-06 3.47E-06 4.28E-06 5.12E-06 5.96E-06 6.81E-06 7.66E-06 8.51E-06 1.02E-05

Urinary Bladder 9.38E-07 1.45E-06 2.01E-06 2.61E-06 3.22E-06 3.85E-06 4.49E-06 5.12E-06 5.76E-06 6.40E-06 7.68E-06

Brain 8.03E-07 1.24E-06 1.72E-06 2.23E-06 2.76E-06 3.29E-06 3.84E-06 4.39E-06 4.93E-06 5.48E-06 6.57E-06

Breast 1.25E-06 1.94E-06 2.68E-06 3.47E-06 4.29E-06 5.13E-06 5.97E-06 6.82E-06 7.67E-06 8.53E-06 1.02E-05

Gall Bladder 9.64E-07 1.49E-06 2.07E-06 2.68E-06 3.31E-06 3.96E-06 4.61E-06 5.27E-06 5.92E-06 6.58E-06 7.90E-06

Heart Wall 1.61E-06 2.50E-06 3.46E-06 4.48E-06 5.54E-06 6.61E-06 7.70E-06 8.80E-06 9.90E-06 1.10E-05 1.32E-05

Kidneys 9.38E-07 1.45E-06 2.01E-06 2.61E-06 3.23E-06 3.85E-06 4.49E-06 5.13E-06 5.76E-06 6.40E-06 7.68E-06

Liver 1.17E-06 1.81E-06 2.51E-06 3.25E-06 4.01E-06 4.79E-06 5.58E-06 6.38E-06 7.17E-06 7.97E-06 9.56E-06

Muscle 9.80E-07 1.52E-06 2.10E-06 2.72E-06 3.37E-06 4.02E-06 4.68E-06 5.35E-06 6.02E-06 6.68E-06 8.02E-06

Ovaries 9.10E-07 1.41E-06 1.95E-06 2.53E-06 3.13E-06 3.73E-06 4.35E-06 4.97E-06 5.59E-06 6.21E-06 7.45E-06

Pancreas 1.13E-06 1.75E-06 2.43E-06 3.14E-06 3.89E-06 4.64E-06 5.41E-06 6.18E-06 6.95E-06 7.72E-06 9.26E-06

Testes 7.84E-07 1.21E-06 1.68E-06 2.18E-06 2.69E-06 3.22E-06 3.75E-06 4.28E-06 4.81E-06 5.35E-06 6.42E-06

Thyroid 9.68E-07 1.50E-06 2.08E-06 2.69E-06 3.33E-06 3.97E-06 4.63E-06 5.29E-06 5.95E-06 6.61E-06 7.93E-06

R.B.M. 1.04E-06 1.62E-06 2.24E-06 2.90E-06 3.59E-06 4.28E-06 4.99E-06 5.70E-06 6.41E-06 7.12E-06 8.55E-06

Bone Surface 9.75E-07 1.51E-06 2.09E-06 2.71E-06 3.35E-06 4.00E-06 4.66E-06 5.33E-06 5.99E-06 6.65E-06 7.99E-06

Stomach 1.85E-06 2.87E-06 3.97E-06 5.14E-06 6.36E-06 7.59E-06 8.84E-06 1.01E-05 1.14E-05 1.26E-05 1.51E-05

S.I. 2.85E-06 4.42E-06 6.12E-06 7.93E-06 9.80E-06 1.17E-05 1.36E-05 1.56E-05 1.75E-05 1.95E-05 2.34E-05

U.L.I. 1.28E-05 1.98E-05 2.75E-05 3.56E-05 4.40E-05 5.26E-05 6.12E-05 6.99E-05 7.87E-05 8.74E-05 1.05E-04

L.L.I. 3.57E-05 5.54E-05 7.66E-05 9.93E-05 1.23E-04 1.47E-04 1.71E-04 1.95E-04 2.19E-04 2.44E-04 2.92E-04

Skin 8.63E-07 1.34E-06 1.85E-06 2.40E-06 2.97E-06 3.54E-06 4.13E-06 4.71E-06 5.30E-06 5.89E-06 7.07E-06

Spleen 1.11E-06 1.72E-06 2.38E-06 3.08E-06 3.81E-06 4.55E-06 5.30E-06 6.05E-06 6.80E-06 7.56E-06 9.07E-06

Thymus 1.35E-06 2.09E-06 2.89E-06 3.75E-06 4.64E-06 5.54E-06 6.45E-06 7.37E-06 8.29E-06 9.20E-06 1.10E-05

Uterus 8.47E-07 1.31E-06 1.82E-06 2.35E-06 2.91E-06 3.48E-06 4.05E-06 4.63E-06 5.20E-06 5.78E-06 6.94E-06

ET 2.56E-05 3.96E-05 5.48E-05 7.11E-05 8.79E-05 1.05E-04 1.22E-04 1.40E-04 1.57E-04 1.74E-04 2.09E-04

ET1 1.80E-03 2.80E-03 3.87E-03 5.02E-03 6.20E-03 7.41E-03 8.63E-03 9.86E-03 1.11E-02 1.23E-02 1.48E-02

ET2 2.38E-05 3.68E-05 5.10E-05 6.61E-05 8.17E-05 9.75E-05 1.14E-04 1.30E-04 1.46E-04 1.62E-04 1.95E-04

Lung 2.87E-04 4.45E-04 6.16E-04 7.99E-04 9.87E-04 1.18E-03 1.37E-03 1.57E-03 1.76E-03 1.96E-03 2.35E-03

Colon 2.27E-05 3.52E-05 4.87E-05 6.31E-05 7.81E-05 9.32E-05 1.09E-04 1.24E-04 1.40E-04 1.55E-04 1.86E-04

LN(ET) 5.23E-05 8.10E-05 1.12E-04 1.45E-04 1.80E-04 2.15E-04 2.50E-04 2.86E-04 3.21E-04 3.57E-04 4.28E-04

BBsec 3.00E-04 4.65E-04 6.43E-04 8.34E-04 1.03E-03 1.23E-03 1.43E-03 1.64E-03 1.84E-03 2.05E-03 2.46E-03

BBbas 2.97E-04 4.61E-04 6.38E-04 8.26E-04 1.02E-03 1.22E-03 1.42E-03 1.62E-03 1.83E-03 2.03E-03 2.43E-03

bb 2.83E-04 4.39E-04 6.08E-04 7.88E-04 9.74E-04 1.16E-03 1.36E-03 1.55E-03 1.74E-03 1.93E-03 2.32E-03

AI 2.81E-04 4.35E-04 6.02E-04 7.80E-04 9.64E-04 1.15E-03 1.34E-03 1.53E-03 1.72E-03 1.91E-03 2.30E-03

LN(TH) 8.32E-05 1.29E-04 1.78E-04 2.31E-04 2.86E-04 3.41E-04 3.98E-04 4.54E-04 5.11E-04 5.68E-04 6.81E-04

Esophagus 1.35E-06 2.09E-06 2.89E-06 3.75E-06 4.64E-06 5.54E-06 6.45E-06 7.37E-06 8.29E-06 9.20E-06 1.10E-05

Gonads 9.10E-07 1.41E-06 1.95E-06 2.53E-06 3.13E-06 3.73E-06 4.35E-06 4.97E-06 5.59E-06 6.21E-06 7.45E-06

Intake rate pCi/d 81.693 63.311 58.415 56.744 56.123 55.882 55.782 55.741 55.726 55.717 55.713

Days 365 730 1095 1461 1826 2190 2556 2922 3287 3652 4383

Intake (pCi) 2.98E+04 4.62E+04 6.40E+04 8.29E+04 1.02E+05 1.22E+05 1.43E+05 1.63E+05 1.83E+05 2.03E+05 2.44E+05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 233 of 248

Table D-25 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 1.19E-05 1.36E-05 1.53E-05 1.70E-05 2.13E-05 2.55E-05 2.98E-05 3.40E-05 3.83E-05 4.25E-05

Urinary Bladder 8.96E-06 1.02E-05 1.15E-05 1.28E-05 1.60E-05 1.92E-05 2.24E-05 2.56E-05 2.88E-05 3.20E-05

Brain 7.67E-06 8.77E-06 9.86E-06 1.10E-05 1.37E-05 1.64E-05 1.92E-05 2.19E-05 2.47E-05 2.74E-05

Breast 1.19E-05 1.36E-05 1.53E-05 1.70E-05 2.13E-05 2.56E-05 2.98E-05 3.41E-05 3.83E-05 4.26E-05

Gall Bladder 9.21E-06 1.05E-05 1.18E-05 1.32E-05 1.65E-05 1.97E-05 2.30E-05 2.63E-05 2.96E-05 3.29E-05

Heart Wall 1.54E-05 1.76E-05 1.98E-05 2.20E-05 2.75E-05 3.30E-05 3.85E-05 4.40E-05 4.95E-05 5.50E-05

Kidneys 8.96E-06 1.02E-05 1.15E-05 1.28E-05 1.60E-05 1.92E-05 2.24E-05 2.56E-05 2.88E-05 3.20E-05

Liver 1.12E-05 1.28E-05 1.44E-05 1.60E-05 2.00E-05 2.39E-05 2.79E-05 3.19E-05 3.59E-05 3.99E-05

Muscle 9.36E-06 1.07E-05 1.20E-05 1.34E-05 1.67E-05 2.00E-05 2.34E-05 2.67E-05 3.01E-05 3.34E-05

Ovaries 8.69E-06 9.94E-06 1.12E-05 1.24E-05 1.56E-05 1.87E-05 2.18E-05 2.49E-05 2.80E-05 3.11E-05

Pancreas 1.08E-05 1.23E-05 1.39E-05 1.54E-05 1.93E-05 2.32E-05 2.70E-05 3.09E-05 3.47E-05 3.86E-05

Testes 7.49E-06 8.56E-06 9.61E-06 1.07E-05 1.34E-05 1.60E-05 1.87E-05 2.14E-05 2.40E-05 2.67E-05

Thyroid 9.25E-06 1.06E-05 1.19E-05 1.32E-05 1.65E-05 1.98E-05 2.31E-05 2.64E-05 2.97E-05 3.30E-05

R.B.M. 9.97E-06 1.14E-05 1.28E-05 1.42E-05 1.78E-05 2.14E-05 2.49E-05 2.85E-05 3.20E-05 3.56E-05

Bone Surface 9.32E-06 1.06E-05 1.20E-05 1.33E-05 1.67E-05 2.00E-05 2.33E-05 2.66E-05 3.00E-05 3.33E-05

Stomach 1.77E-05 2.02E-05 2.27E-05 2.52E-05 3.16E-05 3.79E-05 4.42E-05 5.05E-05 5.68E-05 6.31E-05

S.I. 2.72E-05 3.11E-05 3.50E-05 3.89E-05 4.87E-05 5.84E-05 6.81E-05 7.78E-05 8.76E-05 9.73E-05

U.L.I. 1.22E-04 1.40E-04 1.57E-04 1.75E-04 2.19E-04 2.62E-04 3.06E-04 3.50E-04 3.93E-04 4.37E-04

L.L.I. 3.41E-04 3.90E-04 4.39E-04 4.88E-04 6.10E-04 7.32E-04 8.54E-04 9.76E-04 1.10E-03 1.22E-03

Skin 8.25E-06 9.42E-06 1.06E-05 1.18E-05 1.48E-05 1.77E-05 2.07E-05 2.36E-05 2.66E-05 2.95E-05

Spleen 1.06E-05 1.21E-05 1.36E-05 1.51E-05 1.89E-05 2.27E-05 2.65E-05 3.02E-05 3.40E-05 3.78E-05

Thymus 1.29E-05 1.47E-05 1.66E-05 1.84E-05 2.30E-05 2.76E-05 3.22E-05 3.68E-05 4.14E-05 4.60E-05

Uterus 8.09E-06 9.25E-06 1.04E-05 1.16E-05 1.45E-05 1.73E-05 2.02E-05 2.31E-05 2.60E-05 2.89E-05

ET 2.44E-04 2.79E-04 3.14E-04 3.49E-04 4.36E-04 5.23E-04 6.10E-04 6.98E-04 7.85E-04 8.72E-04

ET1 1.72E-02 1.97E-02 2.22E-02 2.46E-02 3.08E-02 3.70E-02 4.31E-02 4.93E-02 5.54E-02 6.16E-02

ET2 2.27E-04 2.59E-04 2.92E-04 3.24E-04 4.06E-04 4.87E-04 5.68E-04 6.49E-04 7.30E-04 8.11E-04

Lung 2.74E-03 3.14E-03 3.53E-03 3.92E-03 4.90E-03 5.88E-03 6.86E-03 7.84E-03 8.82E-03 9.80E-03

Colon 2.17E-04 2.48E-04 2.79E-04 3.10E-04 3.88E-04 4.65E-04 5.43E-04 6.20E-04 6.98E-04 7.75E-04

LN(ET) 4.99E-04 5.71E-04 6.41E-04 7.12E-04 8.90E-04 1.07E-03 1.25E-03 1.42E-03 1.60E-03 1.78E-03

BBsec 2.86E-03 3.27E-03 3.67E-03 4.08E-03 5.10E-03 6.12E-03 7.14E-03 8.16E-03 9.18E-03 1.02E-02

BBbas 2.84E-03 3.25E-03 3.64E-03 4.04E-03 5.05E-03 6.06E-03 7.07E-03 8.08E-03 9.09E-03 1.01E-02

bb 2.71E-03 3.10E-03 3.48E-03 3.87E-03 4.84E-03 5.80E-03 6.77E-03 7.74E-03 8.70E-03 9.67E-03

AI 2.68E-03 3.06E-03 3.45E-03 3.83E-03 4.79E-03 5.74E-03 6.70E-03 7.66E-03 8.61E-03 9.57E-03

LN(TH) 7.94E-04 9.08E-04 1.02E-03 1.14E-03 1.42E-03 1.70E-03 1.99E-03 2.27E-03 2.56E-03 2.84E-03

Esophagus 1.29E-05 1.47E-05 1.66E-05 1.84E-05 2.30E-05 2.76E-05 3.22E-05 3.68E-05 4.14E-05 4.60E-05

Gonads 8.69E-06 9.94E-06 1.12E-05 1.24E-05 1.56E-05 1.87E-05 2.18E-05 2.49E-05 2.80E-05 3.11E-05

Intake rate pCi/d 55.71 55.71 55.71 55.71 55.71 55.71 55.71 55.71 55.71 55.71

Days 5113 5844 6574 7305 9131 10958 12784 14610 16436 18263

Intake (pCi) 2.85E+05 3.26E+05 3.66E+05 4.07E+05 5.09E+05 6.10E+05 7.12E+05 8.14E+05 9.16E+05 1.02E+06
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 234 of 248

Table D-26. Zn-65, Type S, f1=0.49999, 50-year equivalent dose for years of chronic intake (Sv) MDA = 5.1 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 4.66E-06 7.81E-06 1.13E-05 1.50E-05 1.87E-05 2.24E-05 2.62E-05 2.99E-05 3.36E-05 3.74E-05 4.49E-05

Urinary Bladder 2.76E-06 4.62E-06 6.70E-06 8.87E-06 1.11E-05 1.33E-05 1.55E-05 1.77E-05 1.99E-05 2.21E-05 2.66E-05

Brain 2.54E-06 4.25E-06 6.16E-06 8.16E-06 1.02E-05 1.22E-05 1.42E-05 1.63E-05 1.83E-05 2.04E-05 2.45E-05

Breast 3.38E-06 5.67E-06 8.21E-06 1.09E-05 1.36E-05 1.63E-05 1.90E-05 2.17E-05 2.44E-05 2.71E-05 3.26E-05

Gall Bladder 3.40E-06 5.69E-06 8.25E-06 1.09E-05 1.36E-05 1.64E-05 1.91E-05 2.18E-05 2.45E-05 2.73E-05 3.27E-05

Heart Wall 5.03E-06 8.43E-06 1.22E-05 1.62E-05 2.02E-05 2.42E-05 2.83E-05 3.23E-05 3.63E-05 4.04E-05 4.85E-05

Kidneys 3.41E-06 5.71E-06 8.28E-06 1.10E-05 1.37E-05 1.64E-05 1.92E-05 2.19E-05 2.46E-05 2.74E-05 3.29E-05

Liver 3.95E-06 6.61E-06 9.59E-06 1.27E-05 1.58E-05 1.90E-05 2.22E-05 2.53E-05 2.85E-05 3.16E-05 3.80E-05

Muscle 3.17E-06 5.30E-06 7.69E-06 1.02E-05 1.27E-05 1.52E-05 1.78E-05 2.03E-05 2.28E-05 2.54E-05 3.05E-05

Ovaries 3.49E-06 5.84E-06 8.47E-06 1.12E-05 1.40E-05 1.68E-05 1.96E-05 2.24E-05 2.52E-05 2.80E-05 3.36E-05

Pancreas 4.14E-06 6.93E-06 1.00E-05 1.33E-05 1.66E-05 1.99E-05 2.32E-05 2.66E-05 2.99E-05 3.33E-05 3.99E-05

Testes 2.39E-06 4.01E-06 5.81E-06 7.70E-06 9.60E-06 1.15E-05 1.34E-05 1.54E-05 1.73E-05 1.93E-05 2.31E-05

Thyroid 3.30E-06 5.53E-06 8.01E-06 1.06E-05 1.32E-05 1.59E-05 1.85E-05 2.12E-05 2.39E-05 2.65E-05 3.18E-05

R.B.M. 4.38E-06 7.34E-06 1.06E-05 1.41E-05 1.76E-05 2.11E-05 2.46E-05 2.81E-05 3.16E-05 3.51E-05 4.22E-05

Bone Surface 4.70E-06 7.87E-06 1.14E-05 1.51E-05 1.89E-05 2.26E-05 2.64E-05 3.02E-05 3.40E-05 3.78E-05 4.53E-05

Stomach 3.35E-06 5.61E-06 8.13E-06 1.08E-05 1.34E-05 1.61E-05 1.88E-05 2.15E-05 2.42E-05 2.69E-05 3.23E-05

S.I. 3.40E-06 5.70E-06 8.27E-06 1.09E-05 1.37E-05 1.64E-05 1.91E-05 2.18E-05 2.45E-05 2.73E-05 3.27E-05

U.L.I. 3.40E-06 5.70E-06 8.27E-06 1.10E-05 1.37E-05 1.64E-05 1.91E-05 2.19E-05 2.46E-05 2.74E-05 3.29E-05

L.L.I. 4.04E-06 6.76E-06 9.81E-06 1.30E-05 1.62E-05 1.94E-05 2.27E-05 2.59E-05 2.91E-05 3.24E-05 3.89E-05

Skin 2.11E-06 3.54E-06 5.14E-06 6.80E-06 8.49E-06 1.02E-05 1.19E-05 1.36E-05 1.53E-05 1.70E-05 2.04E-05

Spleen 3.78E-06 6.33E-06 9.17E-06 1.21E-05 1.52E-05 1.82E-05 2.12E-05 2.42E-05 2.72E-05 3.03E-05 3.63E-05

Thymus 4.28E-06 7.16E-06 1.04E-05 1.38E-05 1.72E-05 2.06E-05 2.40E-05 2.75E-05 3.09E-05 3.44E-05 4.13E-05

Uterus 3.25E-06 5.44E-06 7.89E-06 1.05E-05 1.30E-05 1.56E-05 1.82E-05 2.09E-05 2.35E-05 2.61E-05 3.14E-05

ET 1.24E-05 2.08E-05 3.02E-05 4.00E-05 4.99E-05 5.98E-05 6.98E-05 7.98E-05 8.98E-05 9.98E-05 1.20E-04

ET1 6.11E-05 1.02E-04 1.48E-04 1.96E-04 2.45E-04 2.94E-04 3.43E-04 3.92E-04 4.41E-04 4.90E-04 5.88E-04

ET2 1.24E-05 2.07E-05 3.01E-05 3.98E-05 4.97E-05 5.95E-05 6.95E-05 7.94E-05 8.93E-05 9.93E-05 1.19E-04

Lung 1.09E-05 1.83E-05 2.65E-05 3.52E-05 4.39E-05 5.26E-05 6.14E-05 7.01E-05 7.89E-05 8.76E-05 1.05E-04

Colon 3.68E-06 6.16E-06 8.93E-06 1.18E-05 1.48E-05 1.77E-05 2.07E-05 2.36E-05 2.66E-05 2.95E-05 3.54E-05

LN(ET) 1.23E-05 2.06E-05 2.98E-05 3.95E-05 4.93E-05 5.91E-05 6.90E-05 7.89E-05 8.88E-05 9.86E-05 1.18E-04

BBsec 1.16E-05 1.95E-05 2.82E-05 3.74E-05 4.66E-05 5.59E-05 6.52E-05 7.46E-05 8.39E-05 9.33E-05 1.12E-04

BBbas 1.11E-05 1.86E-05 2.70E-05 3.57E-05 4.45E-05 5.34E-05 6.23E-05 7.12E-05 8.01E-05 8.90E-05 1.07E-04

bb 9.95E-06 1.67E-05 2.42E-05 3.20E-05 3.99E-05 4.79E-05 5.59E-05 6.39E-05 7.19E-05 7.99E-05 9.59E-05

AI 1.15E-05 1.92E-05 2.79E-05 3.69E-05 4.61E-05 5.52E-05 6.45E-05 7.37E-05 8.29E-05 9.21E-05 1.11E-04

LN(TH) 9.81E-06 1.64E-05 2.38E-05 3.16E-05 3.94E-05 4.72E-05 5.51E-05 6.30E-05 7.09E-05 7.88E-05 9.45E-05

Esophagus 4.28E-06 7.16E-06 1.04E-05 1.38E-05 1.72E-05 2.06E-05 2.40E-05 2.75E-05 3.09E-05 3.44E-05 4.13E-05

Gonads 3.49E-06 5.84E-06 8.47E-06 1.12E-05 1.40E-05 1.68E-05 1.96E-05 2.24E-05 2.52E-05 2.80E-05 3.36E-05

Intake rate pCi/d 124.78 104.52 101.01 100.27 100.1 100.06 100.05 100.05 100.05 100.05 100.05

Days 365 730 1095 1461 1826 2190 2556 2922 3287 3652 4383

Intake (pCi) 4.55E+04 7.63E+04 1.11E+05 1.46E+05 1.83E+05 2.19E+05 2.56E+05 2.92E+05 3.29E+05 3.65E+05 4.39E+05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 235 of 248

Table D-26 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 5.23E-05 5.98E-05 6.73E-05 7.48E-05 9.34E-05 1.12E-04 1.31E-04 1.50E-04 1.68E-04 1.87E-04

Urinary Bladder 3.10E-05 3.54E-05 3.98E-05 4.43E-05 5.53E-05 6.64E-05 7.74E-05 8.85E-05 9.96E-05 1.11E-04

Brain2.85E-05 3.26E-05 3.67E-05 4.08E-05 5.09E-05 6.11E-05 7.13E-05 8.15E-05 9.17E-05 1.02E-04

Breast 3.80E-05 4.34E-05 4.88E-05 5.43E-05 6.78E-05 8.14E-05 9.49E-05 1.09E-04 1.22E-04 1.36E-04

Gall Bladder 3.82E-05 4.36E-05 4.91E-05 5.45E-05 6.81E-05 8.18E-059.54E-05 1.09E-04 1.23E-04 1.36E-04

Heart Wall 5.65E-05 6.46E-05 7.27E-05 8.08E-05 1.01E-04 1.21E-04 1.41E-04 1.62E-04 1.82E-04 2.02E-04

Kidneys 3.83E-05 4.38E-05 4.93E-05 5.48E-05 6.84E-05 8.21E-05 9.58E-05 1.10E-04 1.23E-04 1.37E-04

Liver 4.43E-05 5.06E-05 5.69E-05 6.33E-05 7.91E-05 9.49E-05 1.11E-04 1.27E-04 1.42E-04 1.58E-04

Muscle 3.55E-05 4.06E-05 4.57E-05 5.08E-05 6.34E-05 7.61E-05 8.88E-05 1.02E-04 1.14E-04 1.27E-04

Ovaries 3.92E-05 4.48E-05 5.04E-05 5.60E-05 7.00E-05 8.40E-05 9.80E-05 1.12E-04 1.26E-04 1.40E-04

Pancreas 4.66E-05 5.32E-05 5.99E-05 6.65E-05 8.31E-05 9.98E-05 1.16E-04 1.33E-04 1.50E-04 1.66E-04

Testes 2.70E-05 3.08E-05 3.47E-05 3.85E-05 4.81E-05 5.78E-05 6.74E-05 7.70E-05 8.66E-05 9.63E-05

Thyroid 3.71E-05 4.24E-05 4.77E-05 5.30E-05 6.63E-05 7.95E-05 9.28E-05 1.06E-04 1.19E-04 1.33E-04

R.B.M. 4.92E-05 5.62E-05 6.32E-05 7.03E-05 8.78E-05 1.05E-04 1.23E-04 1.41E-04 1.58E-04 1.76E-04

Bone Surface 5.29E-05 6.04E-05 6.80E-05 7.55E-05 9.44E-05 1.13E-04 1.32E-04 1.51E-04 1.70E-04 1.89E-04

Stomach 3.76E-05 4.30E-05 4.84E-05 5.38E-05 6.72E-05 8.06E-05 9.41E-05 1.08E-04 1.21E-04 1.34E-04

S.I. 3.82E-05 4.36E-05 4.91E-05 5.45E-05 6.81E-05 8.18E-05 9.54E-05 1.09E-04 1.23E-04 1.36E-04

U.L.I. 3.83E-05 4.38E-05 4.93E-05 5.48E-05 6.84E-05 8.21E-05 9.58E-05 1.10E-04 1.23E-04 1.37E-04

L.L.I. 4.53E-05 5.18E-05 5.83E-05 6.48E-05 8.09E-05 9.71E-05 1.13E-04 1.30E-04 1.46E-04 1.62E-04

Skin 2.38E-05 2.72E-05 3.06E-05 3.40E-05 4.25E-05 5.10E-05 5.95E-05 6.80E-05 7.65E-05 8.50E-05

Spleen 4.24E-05 4.84E-05 5.45E-05 6.05E-05 7.56E-05 9.08E-05 1.06E-04 1.21E-04 1.36E-04 1.51E-04

Thymus 4.81E-05 5.50E-05 6.19E-05 6.88E-05 8.59E-05 1.03E-04 1.20E-04 1.38E-04 1.55E-04 1.72E-04

Uterus 3.66E-05 4.18E-05 4.70E-05 5.23E-05 6.53E-05 7.84E-05 9.14E-05 1.05E-04 1.18E-04 1.31E-04

ET 1.40E-04 1.60E-04 1.80E-04 2.00E-04 2.49E-04 2.99E-04 3.49E-04 3.99E-04 4.49E-04 4.99E-04

ET1 6.86E-04 7.84E-04 8.82E-04 9.80E-04 1.23E-03 1.47E-03 1.72E-03 1.96E-03 2.21E-03 2.45E-03

ET2 1.39E-04 1.59E-04 1.79E-04 1.99E-04 2.48E-04 2.98E-04 3.47E-04 3.97E-04 4.47E-04 4.96E-04

Lung 1.23E-04 1.40E-04 1.58E-04 1.75E-04 2.19E-04 2.63E-04 3.07E-04 3.51E-04 3.94E-04 4.38E-04

Colon 4.13E-05 4.72E-05 5.31E-05 5.90E-05 7.38E-05 8.85E-05 1.03E-04 1.18E-04 1.33E-04 1.48E-04

LN(ET) 1.38E-04 1.58E-04 1.78E-04 1.97E-04 2.47E-04 2.96E-04 3.45E-04 3.95E-04 4.44E-04 4.93E-04

BBsec 1.31E-04 1.49E-04 1.68E-04 1.87E-04 2.33E-04 2.80E-04 3.26E-04 3.73E-04 4.20E-04 4.66E-04

BBbas 1.25E-04 1.42E-04 1.60E-04 1.78E-04 2.23E-04 2.67E-04 3.12E-04 3.56E-04 4.01E-04 4.45E-04

bb 1.12E-04 1.28E-04 1.44E-04 1.60E-04 2.00E-04 2.40E-04 2.80E-04 3.20E-04 3.59E-04 3.99E-04

AI 1.29E-04 1.47E-04 1.66E-04 1.84E-04 2.30E-04 2.76E-04 3.22E-04 3.69E-04 4.15E-04 4.61E-04

LN(TH) 1.10E-04 1.26E-04 1.42E-04 1.58E-04 1.97E-04 2.36E-04 2.76E-04 3.15E-04 3.54E-04 3.94E-04

Esophagus 4.81E-05 5.50E-05 6.19E-05 6.88E-05 8.59E-05 1.03E-04 1.20E-04 1.38E-04 1.55E-04 1.72E-04

Gonads 3.92E-05 4.48E-05 5.04E-05 5.60E-05 7.00E-05 8.40E-05 9.80E-05 1.12E-04 1.26E-04 1.40E-04

Intake rate pCi/d 100.05 100.05 100.05 100.05 100.05 100.05 100.05 100.05 100.05 100.05

Days 5113 5844 6574 7305 9131 10958 12784 14610 16436 18263

Intake (pCi) 5.12E+05 5.85E+05 6.58E+05 7.31E+05 9.14E+05 1.10E+06 1.28E+06 1.46E+06 1.64E+06 1.83E+06
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 236 of 248

If either 95Zr or 95Nb is used as the radionuclide to maximize the dose to the organ of interest, then assume an equal intake of the 

other radionuclide to also have occurred.

Table D-27. Zr-95, Type M, 50-year equivalent dose for years of chronic intake (Sv) MDA = 2.2 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 4.41E-06 8.74E-06 1.31E-05 1.75E-05 2.19E-05 2.62E-05 3.06E-05 3.50E-05 3.93E-05 4.37E-05 5.25E-05

Urinary Bladder 1.40E-06 2.77E-06 4.15E-06 5.54E-06 6.92E-06 8.30E-06 9.69E-06 1.11E-05 1.25E-05 1.38E-05 1.66E-05

Brain 1.51E-06 2.99E-06 4.48E-06 5.97E-06 7.46E-06 8.95E-06 1.04E-05 1.19E-05 1.34E-05 1.49E-05 1.79E-05

Breast 3.20E-06 6.34E-06 9.50E-06 1.27E-05 1.58E-05 1.90E-05 2.22E-05 2.53E-05 2.85E-05 3.17E-05 3.80E-05

Gall Bladder 2.18E-06 4.32E-06 6.47E-06 8.64E-06 1.08E-05 1.29E-05 1.51E-05 1.73E-05 1.94E-05 2.16E-05 2.59E-05

Heart Wall 5.25E-06 1.04E-05 1.56E-05 2.08E-05 2.60E-05 3.12E-05 3.64E-05 4.16E-05 4.68E-05 5.20E-05 6.24E-05

Kidneys 2.38E-06 4.71E-06 7.06E-06 9.42E-06 1.18E-05 1.41E-05 1.65E-05 1.88E-05 2.12E-05 2.35E-05 2.82E-05

Liver 2.97E-06 5.88E-06 8.82E-06 1.18E-05 1.47E-05 1.76E-05 2.06E-05 2.35E-05 2.65E-05 2.94E-05 3.53E-05

Muscle 2.27E-06 4.50E-06 6.76E-06 9.01E-06 1.13E-05 1.35E-05 1.58E-05 1.80E-05 2.03E-05 2.25E-05 2.70E-05

Ovaries 3.31E-06 6.56E-06 9.84E-06 1.31E-05 1.64E-05 1.97E-05 2.29E-05 2.62E-05 2.95E-05 3.28E-05 3.94E-05

Pancreas 3.10E-06 6.14E-06 9.20E-06 1.23E-05 1.53E-05 1.84E-05 2.15E-05 2.45E-05 2.76E-05 3.07E-05 3.68E-05

Testes 7.21E-07 1.43E-06 2.14E-06 2.86E-06 3.57E-06 4.28E-06 5.00E-06 5.71E-06 6.43E-06 7.14E-06 8.57E-06

Thyroid 2.02E-06 4.01E-06 6.01E-06 8.02E-06 1.00E-05 1.20E-05 1.40E-05 1.60E-05 1.80E-05 2.00E-05 2.40E-05

R.B.M. 9.06E-06 1.79E-05 2.69E-05 3.59E-05 4.49E-05 5.38E-05 6.28E-05 7.18E-05 8.07E-05 8.97E-05 1.08E-04

Bone Surface 5.01E-05 9.93E-05 1.49E-04 1.99E-04 2.48E-04 2.98E-04 3.47E-04 3.97E-04 4.47E-04 4.96E-04 5.96E-04

Stomach 2.70E-06 5.35E-06 8.03E-06 1.07E-05 1.34E-05 1.61E-05 1.87E-05 2.14E-05 2.41E-05 2.68E-05 3.21E-05

S.I. 3.71E-06 7.35E-06 1.10E-05 1.47E-05 1.84E-05 2.20E-05 2.57E-05 2.94E-05 3.31E-05 3.67E-05 4.41E-05

U.L.I. 7.99E-06 1.58E-05 2.37E-05 3.17E-05 3.96E-05 4.75E-05 5.54E-05 6.33E-05 7.12E-05 7.91E-05 9.50E-05

L.L.I. 1.83E-05 3.62E-05 5.43E-05 7.25E-05 9.05E-05 1.09E-04 1.27E-04 1.45E-04 1.63E-04 1.81E-04 2.17E-04

Skin 1.31E-06 2.59E-06 3.89E-06 5.19E-06 6.49E-06 7.78E-06 9.08E-06 1.04E-05 1.17E-05 1.30E-05 1.56E-05

Spleen 2.71E-06 5.37E-06 8.05E-06 1.07E-05 1.34E-05 1.61E-05 1.88E-05 2.15E-05 2.41E-05 2.68E-05 3.22E-05

Thymus 3.89E-06 7.70E-06 1.15E-05 1.54E-05 1.92E-05 2.31E-05 2.69E-05 3.08E-05 3.46E-05 3.85E-05 4.62E-05

Uterus 1.90E-06 3.77E-06 5.66E-06 7.55E-06 9.43E-06 1.13E-05 1.32E-05 1.51E-05 1.70E-05 1.89E-05 2.26E-05

Colon 1.24E-05 2.46E-05 3.70E-05 4.93E-05 6.16E-05 7.39E-05 8.62E-05 9.86E-05 1.11E-04 1.23E-04 1.48E-04

ET 6.10E-05 1.21E-04 1.81E-04 2.42E-04 3.02E-04 3.62E-04 4.23E-04 4.83E-04 5.44E-04 6.04E-04 7.25E-04

ET1 7.48E-03 1.48E-02 2.22E-02 2.96E-02 3.70E-02 4.44E-02 5.18E-02 5.92E-02 6.66E-02 7.40E-02 8.89E-02

ET2 5.36E-05 1.06E-04 1.59E-04 2.12E-04 2.65E-04 3.18E-04 3.71E-04 4.25E-04 4.78E-04 5.31E-04 6.37E-04

Lung 1.14E-04 2.26E-04 3.38E-04 4.51E-04 5.64E-04 6.76E-04 7.89E-04 9.03E-04 1.02E-03 1.13E-03 1.35E-03

LN(ET) 3.42E-05 6.78E-05 1.02E-04 1.36E-04 1.69E-04 2.03E-04 2.37E-04 2.71E-04 3.05E-04 3.39E-04 4.07E-04

BBsec 2.79E-04 5.52E-04 8.28E-04 1.11E-03 1.38E-03 1.66E-03 1.93E-03 2.21E-03 2.49E-03 2.76E-03 3.31E-03

BBbas 2.00E-04 3.97E-04 5.95E-04 7.93E-04 9.91E-04 1.19E-03 1.39E-03 1.59E-03 1.78E-03 1.98E-03 2.38E-03

bb 6.64E-05 1.32E-04 1.97E-04 2.63E-04 3.29E-04 3.95E-04 4.61E-04 5.27E-04 5.92E-04 6.58E-04 7.90E-04

AI 3.61E-05 7.14E-05 1.07E-04 1.43E-04 1.79E-04 2.14E-04 2.50E-04 2.86E-04 3.21E-04 3.57E-04 4.29E-04

LN(TH) 1.47E-05 2.90E-05 4.36E-05 5.81E-05 7.26E-05 8.71E-05 1.02E-04 1.16E-04 1.31E-04 1.45E-04 1.74E-04

Esophagus 3.89E-06 7.70E-06 1.15E-05 1.54E-05 1.92E-05 2.31E-05 2.69E-05 3.08E-05 3.46E-05 3.85E-05 4.62E-05

Gonads 3.31E-06 6.56E-06 9.84E-06 1.31E-05 1.64E-05 1.97E-05 2.29E-05 2.62E-05 2.95E-05 3.28E-05 3.94E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 237 of 248

Table D-27 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 6.12E-05 7.00E-05 7.87E-05 8.74E-05 1.09E-04 1.31E-04 1.53E-04 1.75E-04 1.97E-04 2.19E-04

Urinary Bladder 1.94E-05 2.22E-05 2.49E-05 2.77E-05 3.46E-05 4.16E-05 4.85E-05 5.54E-05 6.23E-05 6.93E-05

Brain 2.09E-05 2.39E-05 2.69E-05 2.99E-05 3.73E-05 4.48E-05 5.22E-05 5.97E-05 6.72E-05 7.46E-05

Breast 4.44E-05 5.07E-05 5.70E-05 6.34E-05 7.92E-05 9.51E-05 1.11E-04 1.27E-04 1.43E-04 1.58E-04

Gall Bladder 3.02E-05 3.45E-05 3.88E-05 4.32E-05 5.40E-05 6.47E-05 7.55E-05 8.63E-05 9.71E-05 1.08E-04

Heart Wall 7.28E-05 8.32E-05 9.36E-05 1.04E-04 1.30E-04 1.56E-04 1.82E-04 2.08E-04 2.34E-04 2.60E-04

Kidneys 3.29E-05 3.76E-05 4.23E-05 4.71E-05 5.88E-05 7.06E-05 8.23E-05 9.41E-05 1.06E-04 1.18E-04

Liver 4.12E-05 4.71E-05 5.30E-05 5.88E-05 7.36E-05 8.83E-05 1.03E-04 1.18E-04 1.32E-04 1.47E-04

Muscle 3.15E-05 3.60E-05 4.05E-05 4.50E-05 5.63E-05 6.76E-05 7.88E-05 9.01E-05 1.01E-04 1.13E-04

Ovaries 4.59E-05 5.25E-05 5.90E-05 6.56E-05 8.20E-05 9.84E-05 1.15E-04 1.31E-04 1.48E-04 1.64E-04

Pancreas 4.30E-05 4.91E-05 5.52E-05 6.14E-05 7.67E-05 9.21E-05 1.07E-04 1.23E-04 1.38E-04 1.53E-04

Testes 1.00E-05 1.14E-05 1.29E-05 1.43E-05 1.79E-05 2.14E-05 2.50E-05 2.86E-05 3.21E-05 3.57E-05

Thyroid 2.80E-05 3.21E-05 3.61E-05 4.01E-05 5.01E-05 6.01E-05 7.01E-05 8.01E-05 9.02E-05 1.00E-04

R.B.M. 1.26E-04 1.44E-04 1.61E-04 1.79E-04 2.24E-04 2.69E-04 3.14E-04 3.59E-04 4.04E-04 4.49E-04

Bone Surface 6.95E-04 7.94E-04 8.94E-04 9.93E-04 1.24E-03 1.49E-03 1.74E-03 1.99E-03 2.23E-03 2.48E-03

Stomach 3.75E-05 4.28E-05 4.82E-05 5.35E-05 6.69E-05 8.03E-05 9.37E-05 1.07E-04 1.20E-04 1.34E-04

S.I. 5.14E-05 5.88E-05 6.61E-05 7.35E-05 9.18E-05 1.10E-04 1.29E-04 1.47E-04 1.65E-04 1.84E-04

U.L.I. 1.11E-04 1.27E-04 1.42E-04 1.58E-04 1.98E-04 2.37E-04 2.77E-04 3.17E-04 3.56E-04 3.96E-04

L.L.I. 2.53E-04 2.90E-04 3.26E-04 3.62E-04 4.53E-04 5.43E-04 6.34E-04 7.24E-04 8.15E-04 9.05E-04

Skin 1.82E-05 2.08E-05 2.33E-05 2.59E-05 3.24E-05 3.89E-05 4.54E-05 5.19E-05 5.84E-05 6.49E-05

Spleen 3.76E-05 4.29E-05 4.83E-05 5.37E-05 6.71E-05 8.05E-05 9.39E-05 1.07E-04 1.21E-04 1.34E-04

Thymus 5.39E-05 6.16E-05 6.93E-05 7.70E-05 9.62E-05 1.15E-04 1.35E-04 1.54E-04 1.73E-04 1.92E-04

Uterus 2.64E-05 3.02E-05 3.40E-05 3.77E-05 4.72E-05 5.66E-05 6.60E-05 7.55E-05 8.49E-05 9.43E-05

Colon 1.73E-04 1.97E-04 2.22E-04 2.46E-04 3.08E-04 3.70E-04 4.31E-04 4.93E-04 5.55E-04 6.16E-04

ET 8.46E-04 9.67E-04 1.09E-03 1.21E-03 1.51E-03 1.81E-03 2.11E-03 2.42E-03 2.72E-03 3.02E-03

ET1 1.04E-01 1.18E-01 1.33E-01 1.48E-01 1.85E-01 2.22E-01 2.59E-01 2.96E-01 3.33E-01 3.70E-01

ET2 7.43E-04 8.49E-04 9.55E-04 1.06E-03 1.33E-03 1.59E-03 1.86E-03 2.12E-03 2.39E-03 2.65E-03

Lung 1.58E-03 1.81E-03 2.03E-03 2.26E-03 2.82E-03 3.38E-03 3.95E-03 4.51E-03 5.08E-03 5.64E-03

LN(ET) 4.74E-04 5.42E-04 6.10E-04 6.78E-04 8.47E-04 1.02E-03 1.19E-03 1.36E-03 1.52E-03 1.69E-03

BBsec 3.87E-03 4.42E-03 4.97E-03 5.52E-03 6.90E-03 8.29E-03 9.67E-03 1.10E-02 1.24E-02 1.38E-02

BBbas 2.78E-03 3.17E-03 3.57E-03 3.97E-03 4.96E-03 5.95E-03 6.94E-03 7.93E-03 8.92E-039.91E-03

bb 9.21E-04 1.05E-03 1.18E-03 1.32E-03 1.65E-03 1.97E-03 2.30E-03 2.63E-03 2.96E-03 3.29E-03

AI 5.00E-04 5.71E-04 6.43E-04 7.14E-04 8.93E-04 1.07E-03 1.25E-03 1.43E-03 1.61E-03 1.79E-03

LN(TH) 2.03E-04 2.32E-04 2.61E-04 2.90E-04 3.63E-04 4.36E-04 5.08E-04 5.81E-04 6.53E-04 7.26E-04

Esophagus 5.39E-05 6.16E-05 6.93E-05 7.70E-05 9.62E-05 1.15E-04 1.35E-04 1.54E-04 1.73E-04 1.92E-04

Gonads 4.59E-05 5.25E-05 5.90E-05 6.56E-05 8.20E-05 9.84E-05 1.15E-04 1.31E-04 1.48E-04 1.64E-04
Effective Date: 0

4/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 238 of 248

If either 95Zr or 95Nb is used as the radionuclide to maximize the dose to the organ of interest, then assume an equal intake of the 

other radionuclide to also have occurred.

Table D-28. Zr-95, Type S, 50-year equivalent dose for years of chronic intake (Sv) MDA = 2.2 nCi. Years of Exposure

Organ 1 2 3 4 5 6 7 8 9 10 12

Adrenals 4.59E-06 9.11E-06 1.37E-05 1.82E-05 2.28E-05 2.73E-05 3.19E-05 3.65E-05 4.10E-05 4.56E-05 5.47E-05

Urinary Bladder 7.76E-07 1.54E-06 2.31E-06 3.08E-06 3.85E-06 4.61E-06 5.39E-06 6.16E-06 6.93E-06 7.69E-06 9.23E-06

Brain 5.10E-07 1.01E-06 1.52E-06 2.03E-06 2.53E-06 3.04E-06 3.54E-06 4.05E-06 4.56E-06 5.06E-06 6.07E-06

Breast 4.80E-06 9.51E-06 1.43E-05 1.90E-05 2.38E-05 2.85E-05 3.33E-05 3.81E-05 4.28E-05 4.76E-05 5.71E-05

Gall Bladder 2.13E-06 4.22E-06 6.32E-06 8.44E-06 1.05E-05 1.26E-05 1.48E-05 1.69E-05 1.90E-05 2.11E-05 2.53E-05

Heart Wall 8.04E-06 1.59E-05 2.39E-05 3.19E-05 3.99E-05 4.78E-05 5.58E-05 6.38E-05 7.17E-05 7.97E-05 9.57E-05

Kidneys 1.85E-06 3.67E-06 5.50E-06 7.34E-06 9.18E-06 1.10E-05 1.28E-05 1.47E-05 1.65E-05 1.84E-05 2.20E-05

Liver 3.92E-06 7.78E-06 1.17E-05 1.56E-05 1.95E-05 2.33E-05 2.72E-05 3.11E-05 3.50E-05 3.89E-05 4.67E-05

Muscle 2.25E-06 4.45E-06 6.68E-06 8.91E-06 1.11E-05 1.34E-05 1.56E-05 1.78E-05 2.01E-05 2.23E-05 2.67E-05

Ovaries 2.40E-06 4.75E-06 7.12E-06 9.50E-06 1.19E-05 1.42E-05 1.66E-05 1.90E-05 2.14E-05 2.38E-05 2.85E-05

Pancreas 3.47E-06 6.88E-06 1.03E-05 1.38E-05 1.72E-05 2.07E-05 2.41E-05 2.76E-05 3.10E-05 3.44E-05 4.13E-05

Testes 2.71E-07 5.38E-07 8.07E-07 1.08E-06 1.35E-06 1.61E-06 1.88E-06 2.15E-06 2.42E-06 2.69E-06 3.23E-06

Thyroid 1.99E-06 3.94E-06 5.91E-06 7.88E-06 9.85E-06 1.18E-05 1.38E-05 1.58E-05 1.77E-05 1.97E-05 2.37E-05

R.B.M. 3.12E-06 6.19E-06 9.29E-06 1.24E-05 1.55E-05 1.86E-05 2.17E-05 2.48E-05 2.79E-05 3.10E-05 3.72E-05

Bone Surface 4.07E-06 8.07E-06 1.21E-05 1.61E-05 2.02E-05 2.42E-05 2.83E-05 3.23E-05 3.63E-05 4.04E-05 4.84E-05

Stomach 3.21E-06 6.36E-06 9.53E-06 1.27E-05 1.59E-05 1.91E-05 2.23E-05 2.54E-05 2.86E-05 3.18E-05 3.82E-05

S.I. 3.14E-06 6.22E-06 9.33E-06 1.25E-05 1.56E-05 1.87E-05 2.18E-05 2.49E-05 2.80E-05 3.11E-05 3.74E-05

U.L.I. 8.48E-06 1.68E-05 2.52E-05 3.37E-05 4.21E-05 5.04E-05 5.89E-05 6.73E-05 7.57E-05 8.41E-05 1.01E-04

L.L.I. 2.03E-05 4.02E-05 6.03E-05 8.04E-05 1.00E-04 1.21E-04 1.41E-04 1.61E-04 1.81E-04 2.01E-04 2.41E-04

Skin 1.14E-06 2.27E-06 3.40E-06 4.54E-06 5.67E-06 6.80E-06 7.94E-06 9.08E-06 1.02E-05 1.13E-05 1.36E-05

Spleen 3.38E-06 6.71E-06 1.01E-05 1.34E-05 1.68E-05 2.01E-05 2.35E-05 2.68E-05 3.02E-05 3.36E-05 4.03E-05

Thymus 5.65E-06 1.12E-05 1.68E-05 2.24E-05 2.80E-05 3.36E-05 3.92E-05 4.48E-05 5.04E-05 5.60E-05 6.72E-05

Uterus 1.15E-06 2.29E-06 3.43E-06 4.58E-06 5.72E-06 6.86E-06 8.01E-06 9.15E-06 1.03E-05 1.14E-05 1.37E-05

Colon 1.36E-05 2.69E-05 4.04E-05 5.39E-05 6.73E-05 8.07E-05 9.42E-05 1.08E-04 1.21E-04 1.35E-04 1.62E-04

ET 8.38E-05 1.66E-04 2.49E-04 3.32E-04 4.15E-04 4.98E-04 5.81E-04 6.65E-04 7.48E-04 8.31E-04 9.97E-04

ET1 8.47E-03 1.68E-02 2.52E-02 3.36E-02 4.20E-02 5.04E-02 5.88E-02 6.72E-02 7.56E-02 8.40E-02 1.01E-01

ET2 7.54E-05 1.49E-04 2.24E-04 2.99E-04 3.74E-04 4.48E-04 5.23E-04 5.98E-04 6.73E-04 7.48E-04 8.97E-04

Lung 1.71E-04 3.39E-04 5.08E-04 6.78E-04 8.48E-04 1.02E-03 1.19E-03 1.36E-03 1.53E-03 1.70E-03 2.04E-03

LN(ET) 4.09E-05 8.12E-05 1.22E-04 1.62E-04 2.03E-04 2.43E-04 2.84E-04 3.25E-04 3.65E-04 4.06E-04 4.87E-04

BBsec 4.02E-04 7.97E-04 1.20E-03 1.60E-03 1.99E-03 2.39E-03 2.79E-03 3.19E-03 3.59E-03 3.99E-03 4.79E-03

BBbas 2.89E-04 5.74E-04 8.60E-04 1.15E-03 1.43E-03 1.72E-03 2.01E-03 2.30E-03 2.58E-03 2.87E-03 3.44E-03

bb 1.04E-04 2.05E-04 3.08E-04 4.11E-04 5.13E-04 6.16E-04 7.18E-04 8.21E-04 9.24E-04 1.03E-03 1.23E-03

AI 6.42E-05 1.27E-04 1.91E-04 2.55E-04 3.18E-04 3.82E-04 4.46E-04 5.10E-04 5.73E-04 6.37E-04 7.64E-04

LN(TH) 2.65E-05 5.26E-05 7.88E-05 1.05E-04 1.31E-04 1.58E-04 1.84E-04 2.10E-04 2.37E-04 2.63E-04 3.16E-04

Esophagus 5.65E-06 1.12E-05 1.68E-05 2.24E-05 2.80E-05 3.36E-05 3.92E-05 4.48E-05 5.04E-05 5.60E-05 6.72E-05

Gonads 2.40E-06 4.75E-06 7.12E-06 9.50E-06 1.19E-05 1.42E-05 1.66E-05 1.90E-05 2.14E-05 2.38E-05 2.85E-05
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 239 of 248

Table D-28 (Continued)

Years of Exposure

Organ 14 16 18 20 25 30 35 40 45 50

Adrenals 6.38E-05 7.29E-05 8.20E-05 9.11E-05 1.14E-04 1.37E-04 1.59E-04 1.82E-04 2.05E-04 2.28E-04

Urinary Bladder 1.08E-05 1.23E-05 1.39E-05 1.54E-05 1.92E-05 2.31E-05 2.69E-05 3.08E-05 3.46E-05 3.85E-05

Brain 7.09E-06 8.10E-06 9.11E-06 1.01E-05 1.27E-05 1.52E-05 1.77E-05 2.03E-05 2.28E-05 2.53E-05

Breast 6.66E-05 7.61E-05 8.56E-05 9.51E-05 1.19E-04 1.43E-04 1.66E-04 1.90E-04 2.14E-04 2.38E-04

Gall Bladder 2.95E-05 3.37E-05 3.80E-05 4.22E-05 5.27E-05 6.33E-05 7.38E-05 8.44E-05 9.49E-05 1.05E-04

Heart Wall 1.12E-04 1.28E-04 1.43E-04 1.59E-04 1.99E-04 2.39E-04 2.79E-04 3.19E-04 3.59E-04 3.99E-04

Kidneys 2.57E-05 2.94E-05 3.30E-05 3.67E-05 4.59E-05 5.51E-05 6.42E-05 7.34E-05 8.26E-05 9.18E-05

Liver 5.45E-05 6.23E-05 7.01E-05 7.78E-05 9.73E-05 1.17E-04 1.36E-04 1.56E-04 1.75E-04 1.95E-04

Muscle 3.12E-05 3.56E-05 4.01E-05 4.46E-05 5.57E-05 6.68E-05 7.80E-05 8.91E-05 1.00E-04 1.11E-04

Ovaries 3.33E-05 3.80E-05 4.28E-05 4.75E-05 5.94E-05 7.13E-05 8.32E-05 9.50E-05 1.07E-04 1.19E-04

Pancreas 4.82E-05 5.51E-05 6.20E-05 6.89E-05 8.61E-05 1.03E-04 1.21E-04 1.38E-04 1.55E-04 1.72E-04

Testes 3.77E-06 4.31E-06 4.85E-06 5.38E-06 6.73E-06 8.08E-06 9.42E-06 1.08E-05 1.21E-05 1.35E-05

Thyroid 2.76E-05 3.15E-05 3.55E-05 3.94E-05 4.93E-05 5.91E-05 6.90E-05 7.88E-05 8.87E-05 9.86E-05

R.B.M. 4.34E-05 4.96E-05 5.58E-05 6.20E-05 7.75E-05 9.30E-05 1.08E-04 1.24E-04 1.39E-04 1.55E-04

Bone Surface 5.65E-05 6.46E-05 7.27E-05 8.07E-05 1.01E-04 1.21E-04 1.41E-04 1.61E-04 1.82E-04 2.02E-04

Stomach 4.45E-05 5.09E-05 5.72E-05 6.36E-05 7.95E-05 9.54E-05 1.11E-04 1.27E-04 1.43E-04 1.59E-04

S.I. 4.36E-05 4.98E-05 5.60E-05 6.23E-05 7.78E-05 9.34E-05 1.09E-04 1.25E-04 1.40E-04 1.56E-04

U.L.I. 1.18E-04 1.35E-04 1.51E-04 1.68E-04 2.10E-04 2.52E-04 2.94E-04 3.37E-04 3.79E-04 4.21E-04

L.L.I. 2.81E-04 3.22E-04 3.62E-04 4.02E-04 5.02E-04 6.03E-04 7.03E-04 8.04E-04 9.04E-04 1.01E-03

Skin 1.59E-05 1.82E-05 2.04E-05 2.27E-05 2.84E-05 3.40E-05 3.97E-05 4.54E-05 5.11E-05 5.67E-05

Spleen 4.70E-05 5.37E-05 6.04E-05 6.71E-05 8.39E-05 1.01E-04 1.17E-04 1.34E-04 1.51E-04 1.68E-04

Thymus 7.84E-05 8.96E-05 1.01E-04 1.12E-04 1.40E-04 1.68E-04 1.96E-04 2.24E-04 2.52E-04 2.80E-04

Uterus 1.60E-05 1.83E-05 2.06E-05 2.29E-05 2.86E-05 3.43E-05 4.00E-05 4.58E-05 5.15E-05 5.72E-05

Colon 1.89E-04 2.15E-04 2.42E-04 2.69E-04 3.37E-04 4.04E-04 4.71E-04 5.39E-04 6.06E-04 6.73E-04

ET 1.16E-03 1.33E-03 1.50E-03 1.66E-03 2.08E-03 2.49E-03 2.91E-03 3.32E-03 3.74E-03 4.15E-03

ET1 1.18E-01 1.34E-01 1.51E-01 1.68E-01 2.10E-01 2.52E-01 2.94E-01 3.36E-01 3.78E-01 4.20E-01

ET2 1.05E-03 1.20E-03 1.35E-03 1.50E-03 1.87E-03 2.24E-03 2.62E-03 2.99E-03 3.37E-03 3.74E-03

Lung 2.37E-03 2.71E-03 3.05E-03 3.39E-03 4.24E-03 5.09E-03 5.94E-03 6.78E-03 7.63E-03 8.48E-03

LN(ET) 5.68E-04 6.50E-04 7.31E-04 8.12E-04 1.02E-03 1.22E-03 1.42E-03 1.62E-03 1.83E-03 2.03E-03

BBsec 5.58E-03 6.38E-03 7.18E-03 7.98E-03 9.97E-03 1.20E-02 1.40E-02 1.60E-02 1.79E-02 1.99E-02

BBbas 4.02E-03 4.59E-03 5.17E-03 5.74E-03 7.17E-03 8.61E-03 1.00E-02 1.15E-02 1.29E-02 1.43E-02

bb 1.44E-03 1.64E-03 1.85E-03 2.05E-03 2.57E-03 3.08E-03 3.59E-03 4.11E-03 4.62E-03 5.13E-03

AI 8.92E-04 1.02E-03 1.15E-03 1.27E-03 1.59E-03 1.91E-03 2.23E-03 2.55E-03 2.87E-03 3.18E-03

LN(TH) 3.68E-04 4.21E-04 4.73E-04 5.26E-04 6.57E-04 7.89E-04 9.20E-04 1.05E-031.18E-03 1.31E-03

Esophagus 7.84E-05 8.96E-05 1.01E-04 1.12E-04 1.40E-04 1.68E-04 1.96E-04 2.24E-04 2.52E-04 2.80E-04

Gonads 3.33E-05 3.80E-05 4.28E-05 4.75E-05 5.94E-05 7.13E-05 8.32E-05 9.50E-05 1.07E-04 1.19E-04
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ATTACHMENT E

OCCUPATIONAL EXTERNAL DOSE FOR MONITORED WORKERS

TABLE OF CONTENTS

Section

E.0 Occupational External Dose .....................................................................................................

Page

241

E.1 Years of Exposure.......................................................................................................... 241

E.2 Rate ............................................................................................................................... 241

E.3 Radiation Type ............................................................................................................... 241

E.3.1 Beta/Photon.............................................................................................................. 241

E.3.2 Neutron .................................................................................................................... 241

E.4 Distribution Parameters.................................................................................................. 242

E.4.1 Parameter #1............................................................................................................ 242

E.4.1.1 Resolution of Photon, Neutron and Tritium Dose ............................................. 243

E.4.1.2 Adjustments to Recorded Photon Dose ........................................................... 243

E.4.1.3 Adjustments to Recorded Neutron Dose.......................................................... 244

E.4.1.4 Unmonitored Photon Dose............................................................................... 244

E.4.1.5 Missed Photon Dose........................................................................................ 244

E.4.1.5.1 Facility/Location ...................................................................................... 244

E.4.1.5.2 Dosimeter Type....................................................................................... 245

E.4.1.5.3 Year ........................................................................................................ 246

E.4.1.5.4 Energy Range ......................................................................................... 246

E.4.1.6 Missed Neutron Dose Prior to 1971 ................................................................. 246

E.4.1.7 Missed Neutron Dose After 1970 ..................................................................... 247

E.4.1.8 Organ Dose Equivalent.................................................................................... 247

E.4.2 Parameter #2............................................................................................................ 248

E.4.3 Organ Dose Conversion Factors .............................................................................. 248

LIST OF TABLES

Table

E-1 IREP dose parameter input screen...........................................................................

Page

241

E-2 Selection of beta and photon radiation energies and percentages for SRS 

facilities and processes............................................................................................. 242

E-3 Selection of neutron radiation energies and percentages for SRS facilities 

and processes .......................................................................................................... 243

E-4 Adjustments to reported SRS deep photon dose ...................................................... 244

E-5 Summary of neutron dose fractions and associated ICRP 60 correction 

factors ...................................................................................................................... 245

E-6 Missed photon dose according to SRS facility .......................................................... 245

E-7 Missed photon dose according to dosimeter type ..................................................... 246

E-8 Missed photon dose according to year ..................................................................... 246

E-9 Neutron-to-photon ratios by SRS process................................................................. 247

E-10 SRS TLND Limits of Detection.................................................................................. 248

E-11 Default exposure geometries to calculate organ dose .............................................. 248
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E.0 OCCUPATIONAL EXTERNAL DOSE

The information needed to evaluate claims is directed to the technical parameters of the annual 

estimates of the primary organ dose that is calculated from the dosimeter interpreted personal dose 

equivalent, Hp(10), and, in the case of skin, testicular and breast cancer, Hp(0.07), used as a 

consistent basis of comparison for all years of SRS occupational radiation exposure. 

The primary IREP screen used to input dose parameters is illustrated in Table E-1. Input to these 

fields is obtained from the SRS dose of record. The claim provides the primary organ of interest and 

other worker information needed to run IREP. Guidance to the dose reconstruction analyst to select 

technical external dose parameters to complete the respective fields in Table E-1 is presented in the 

following sections. 

Table E-1. IREP dose parameter input screen.

Exposure Radiation type Distribution parameters

# Year Rate Type 1 2 3

1 1960 Acute Photon, 30-250 keV Normal 2 2 0

2 1961 Acute

E.1 Years of Exposure

The years of exposure should be identified from the SRS radiation dose reports. Missed dose is 

calculated for all zero or missing records within a reporting year as noted by non-zero or blank entries. 

SRS policies require monitoring of all workers occupationally exposed. Missed dose should be 

calculated for a claimant for each year of record as an employee unless there are valid reasons for 

years in which there is no SRS recorded dose. 

E.2 Rate 

As identified in OCAS-IG-001, acute is selected for all types of external beta and photon dose. 

Chronic is selected for tritium and neutron dose.

E.3 Radiation Type

E.3.1 Beta/Photon

Claimant-favorable assumptions should be made using guidance in Table E-2 for beta and photon 

(i.e., x-ray and gamma rays) radiation types to assure that dose is not underestimated. The values 

presented in this table are intended to provide a reasonable estimate of parameters used to calculate 

the organ dose without significant numerical error for long-term SRS workers in the respective 

facilities. There is no direct evidence to select the specific values shown other than considerations of 

the radiation sources and usual work tasks. In those cases where there is some doubt in the values, a 

range of realistic values should be selected for comparison and the most claimant-favorable option 

selected.

E.3.2 Neutron

Table E-3 summarizes the default neutron dose distributions by energy for each neutron exposure 

area. 
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Table E-2. Selection of beta and photon radiation energies and percentages for SRS facilities and 

processes. Process/

buildings Description

Operations Radiation 

type

Energy 

Begin End selection Percent

Fuel fabrication Produced reactor fuel and target assemblies. 1953 1989 Beta > 15 keV 100%

300M Uranium fuel fabrication 1953 1989 Photon 30 – 250 keV 100%

Reactors During Operation: Highly dispersed fields of higher-energy photon radiation 

fields from fission process, activation/ fission product nuclides.. Potential for 

significant airborne nuclides and there may be significant higher-energy beta 

radiation. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50%

While not in Operation: Highly dispersed fields of higher-energy photon 

radiation fields from activation/fission product nuclides.. There may be 

significant higher-energy beta radiation during maintenance work resulting from 

fission products.

105C Heavy water moderated reactors 3/1955 6/1985

105K 10/1954 1988

1992 7/1992

105L 7/1954 1968

1985 6/1988

105P 2/1954 8/1988

105R 12/1953 6/1964

Processing plants Radiochemical Operations: Highly dispersed fields of higher-energy photon 

radiation fields from activation/fission product nuclides dominant to most 

exposure profiles. Potential for higher-energy beta radiation during sampling 

and, maintenance work resulting from fission products. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50% 200F Radiochemical processing plant 1954

200H Radiochemical processing plant 1955

utonium 

production

Plutonium Component Production: Plutonium is machined into weapon 

components using a glove box assembly process with predominant close 

anterior exposure to workers. Radiation characteristics in this area involve 

significant lower-energy photons and lower-energy neutron radiation.

Photon < 30 keV

30 – 250 keV

25%

75%

Plutonium Storage: Radiation characteristics in this area generally involve 

dispersed lower-energy neutron radiation and scattered photons to include 60 

keV Am-241 gamma ray.

221F, B-line Plutonium Finishing Process 1954

221H, B-line Plutonium Finishing Process 1955

772F Production Control Laboratory 1955

235F Plutonium Fuel Facility 1955

736A Calibration Facility 1953

adionuclide 

production

Radiation characteristics are highly dependent upon the radionuclide being 

produced. SRS has produced several radionuclides including tritium, 238Pu, 239Pu, americium, curium, californium, etc. 

(Specific to 

radionuclide)

Photon, 

default

30 – 250 keV

> 250 keV

50%

50%

773A Savannah River Laboratory 1953

Waste handling Radiation characteristics are highly dependent upon the source of waste but 

typically fission product nuclides (90Sr/90Y, 137Cs) are dominant. Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50% 200F General waste handling, storage tanks, etc. 1953 Ongoing

200H General waste handling, storage tanks, etc 1953 Ongoing

E.4 Distribution Parameters

Selection of the distribution parameters in Table E-1 involves adjustments to the dose of record for 

missed dose prior to entry into the IREP input screen. The selection of a normal distribution for the 

type determines the definition of Parameters #1 and #2. For a normal distribution, Parameter #3 is 

not used.

E.4.1 Parameter #1

For a normal distribution, parameter #1 is the mean of the distribution of recorded dose for each year 

of monitoring. Prior to calculating this, it is necessary to adjust the recorded dose for exposure profiles 

described in Section E.4.1.1 where the recorded dose of record will under-estimate Hp(10). 
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Table E-3. Selection of neutron radiation energies and percentages for SRS facilities and processes.

Process Description/buildings

Operations Neutron 

energy

Default dose fraction 

Begin End (%)

Fuel fabrication Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and plutonium – aluminum 

(Pu-Al) alloy targets.

(321-M) 0.1-2 MeV

2-20 MeV

83%

17%

Reactors During Reactor Operation: Low-level neutron exposure through shielding. In addition, some intermittent neutron 

exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987

10-100 keV

0.1 – 2 MeV

15%

85%

(105K) 10/1954 1/1988

1992 7/1992

(105L) 7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium 

production

Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into production pucks. 

Work is primarily conducted in glove boxes with predominant close anterior exposure to workers. Radiation 

levels at the beginning of the process are fairly constant while levels at end of process are closely related to 

production levels. 

Plutonium Finishing Process

(221-H, B-Line)

11/1954 1988 0.1-2 MeV

2-20 MeV

60%

40%

Plutonium Finishing Process

(221-F, B-Line)

10/1955 1988 0.1-2 MeV 100%

Plutonium Storage 0.1-2 MeV

2-20 MeV

86%

14%

Production Control Laboratory

(772-F)

1955 0.1-2 meV 100%

Plutonium Fuel Facility

(235-F)

1955 0.1-2 meV 100%

Radionuclide 

production and 

calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS has produced 

several radionuclides including 238Pu, 239Pu, americium, curium, californium, etc. At the calibration facility 

radiological instruments are calibrated using a variety of sources with different neutron energy spectra. 

Californium Production Facility (773-A) 0.1-2 MeV

2-20 MeV

50%

50%

Calibration Facility (736-A) 0.1-2 MeV

2-20 MeV

83%

17%

E.4.1.1 Resolution of Photon, Neutron and Tritium Dose

For workers who terminated SRS employment prior to 1979, SRS hard-copy radiation records must

be consulted to separate the photon, neutron and tritium dose for workers. Hard copy records do 

separate the recorded whole body dose (i.e., photon + neutron + tritium) with the exception of the 

period from the second quarter of 1963 through December 31, 1972. The missing dose components 

of the total whole body dose could be estimated from the average of the prior year (i.e., before second 

quarter of 1963) and subsequent year (after 1972) dose results according to the proportion of dose 

from photon, neutron and tritium dose based on examination of continuity in the worker’s job and work 

activities. A claimant-favorable option is simply to assign the whole body dose (i.e., photon + neutron 

+ tritium) as a photon dose, and using the methods in the following sections a claimant-favorable 

neutron dose will also be calculated assuming the worker was working in a facility with neutron 

exposure.

E.4.1.2 Adjustments to Recorded Photon Dose

Adjustments to the SRS reported photon dose are necessary to arrive at a claimant-favorable dose 

considering the uncertainty associated primarily with the complex workplace radiation fields and 

exposure orientations. Taylor et al (1995) identified calibration corrections of 1.119 and 1.039 to the 

reported SRS TLD-Deep photon dose for the periods prior to 1986 and 1987, respectively, to estimate 

the Hp(10) dose. This is summarized in Table E-4.
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Table E-4. Adjustments to reported SRS deep photon dose.

Time period Dosimeter Facility Step Adjustment to reported dose

Prior to 1987 All beta/photon 

dosimeters

All facilities A Multiply reported TLD-Deep photon dose by 

a factor of 1.119 to estimate Hp(10).

For the year of 

1987

TLD beta/photon 

dosimeter

All facilities B Multiply reported TLD-Deep photon dose by 

a factor of 1.039 to estimate Hp(10).

Post 1987 TLD beta/photon 

dosimeter

All facilities No adjustments made. 

E.4.1.3 Adjustments to Recorded Neutron Dose

The Savannah River Site (SRS) incorporated the energy variation of the dose equivalent into their 

calibration methodology. As a result, the recorded neutron dose is a combination of all neutron 

energies. In order to calculate the dose input to IREP (Table E-1), the recorded neutron dose must be 

separated into neutron energy groups as shown in Table E-3 and subsequently converted to ICRP 60 

(1990) methodology. 

The neutron dose equivalent correction factor can be calculated by dividing the dose fractions from 

Table E-4 for each facility and neutron energy category group shown in Table E-2 by multiplying the 

recorded neutron dose by the area specific correction factors the neutron dose equivalent is 

calculated. For example consider a 1000 mrem recorded neutron dose by an energy employee 

working at the HB-Line, the corrected neutron dose is 1144 mrem from neutrons between 0.1-2.0 

MeV and

E.4.1.4 Unmonitored Photon Dose

529 mrem from neutrons with energy between 2 and 20 MeV. Thus the corrected neutron 

dose is a total of 1673 mrem. These corrections should be applied to measured dose, missed dose 

and dose determined based on a neutron to photon ratio. Table E-5 summarizes the dose fractions 

and associated correction factors for each neutron exposure area. 

Adjustments to the recorded annual dose can be made using dose results for co-workers or using the 

recorded dose prior to and after the period of missed dose. These situations do require careful 

examination. 

E.4.1.5 Missed Photon Dose

The missed photon dose for dosimeter results less-than the MDL is particularly important for earlier

years when MDLs were higher and dosimeter exchange was more frequent. OCAS-IG-001 guidance 

should be followed to calculate the missed dose by using a claimant favorable maximum potential 

missed dose. This is calculated using the Minimum Detection Level (MDL) multiplied by the number 

of zero dose results to provide an estimate of the maximum potential missed dose. The following 

sections describe the potential missed photon dose corrections according to facility/location, 

dosimeter type, year and energy range.

E.4.1.5.1 Facility/Location

The potential missed photon dose for the laboratory determined MDL and exchange frequency is 

presented in Table E-6. 
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Table E-5. Summary of neutron dose fractions and associated ICRP 60 correction factors.

Process Description/buildings

Operations Neutron 

energy

Default dose 

fraction (%)

ICRP 60 

Begin End correction factor

Fuel fabrication Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and plutonium – aluminum 

(Pu-Al) alloy targets.

(321-M) 0.1-2 MeV

2-20 MeV

83%

17%

1.58

0.23

Reactors During Reactor Operation: Low-level neutron exposure through shielding. In addition, some intermittent neutron 

exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987

10-100 keV

0.1 – 2 MeV

15%

85%

0.28

1.62

(105K) 10/1954 1/1988

1992 7/1992

(105L) 7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium 

production

Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into production pucks. Work 

is primarily conducted in glove boxes with predominant close anterior exposure to workers. Radiation levels at the 

beginning of the process are fairly constant while levels at end of process are closely related to production levels. 

Plutonium Finishing Process

(221-H, B-Line)

11/1954 1988 0.1-2 MeV

2-20 MeV

60%

40%

1.14

0.53

Plutonium Finishing Process

(221-F, B-Line)

10/1955 1988 0.1-2 MeV 100% 1.91

Plutonium Storage 0.1-2 MeV

2-20 MeV

86%

14%

1.64

0.19

Production Control Laboratory

(772-F)

1955

Plutonium Fuel Facility

(235-F)

1955

Radionuclide

production and 

calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS has produced several 

radionuclides including 238Pu, 239Pu, americium, curium, californium, etc. At the calibration facility radiological 

instruments are calibrated using a variety of sources with different neutron energy spectra.

Californium Production Facility 

(773-A)

0.1-2 MeV

2-20 MeV

50%

50%

Calibration Facility (736-A) 0.1-2 MeV

2-20 MeV

83%

17%

Table E-6. Missed photon dose according to SRS facility.

SRS facility type Dosimeter Period of use

Exchange 

frequency

MDL 

(mSv)a

Max. annual 

missed dose

(mSv)b

Fuel fabrication, SRS 

reactors, processing 

plants, waste handling

ORNL two-element 

film dosimeter

Prior to 3/52 Weekly 0.4 20.8 

SRS two element 

film dosimeter

After 3/52 through 9/30/1957 Weekly 0.4 20.8 

10/1/1957 through 11/8/1959 Biweekly 0.4 10.4 

Above plus plutonium 

and nuclide production 

SRS multi-element 

film dosimeter

11/9/1959 through 12/31/1964 Biweekly 0.4 10.4

1/1/1965 through 12/31/1965 4-week 0.4 5.2

1/1/1966 through 3/31/1970 Monthly 0.4 4.8

All SRS SRS TLD 4/1/1970 through 6/30/1983 Monthly 0.15 1.8

Panasonic TLD 7/1/1983 to 2003 (ongoing) Monthly 0.05 0.6

Panasonic TLND 1/1/1995 to 2003 (ongoing) Monthly (n=12) 0.15 0.9

a. Estimated film dosimeter detection levels based on NIOSH (1993), NRC (1989), and Wilson et al (1990). TLD detection levels from 

Taylor et al (1995) and SRS (1993). 

b. Maximum annual missed dose calculated from OCAS-IG-001 (NIOSH 2000). 

E.4.1.5.2 Dosimeter Type

Table E-7 presents the maximum potential missed dose according to SRS facility, time period, 

exchange frequency and MDL.
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Table E-7. Missed photon dose according to dosimeter type.

Dosimeter

type Period of use

MDL

(mSv)

Exchange 

frequency

Max. annual 

missed dose

(mSv)

Two-element film Prior to 3/52 through 9/30/1957 0.4 Weekly (n = 52) 20.8

10/1/1957 through 11/8/1959 0.4 Biweekly (n=26) 10.4

Multi-element film 11/9/1959 through 12/31/1964 0.4 Biweekly (n=26) 10.4

1/1/1965 through 12/31/1965 0.4 4-week (n=13) 5.2

1/1/1966 through 3/31/1970 0.4 Monthly (n=12) 4.8

TLD 4/1/1970 through 6/30/1983 0.15 Monthly (n=12) 1.8

7/1/1983 to 2003 (ongoing) 0.05 Monthly (n=12) 0.6

0.05 Quarterly (n = 4) 0.2

E.4.1.5.3 Year

Table E-8 presents the potential maximum missed photon dose according to year (actually by period 

according to dosimeter type and exchange). 

Table E-8. Missed photon dose according to year.

Period of usea

MDL 

(mSv)

Exchange 

frequency

Max. annual 

missed dose

(mSv)

Prior to 1958 0.4 Weekly (n = 52) 20.8

1959 0.4 Biweekly (n=26) 10.4

1960 through 1964 0.4 Biweekly (n=26) 10.4

1965 0.4 4-week (n=13) 5.2

1966 through 1969 0.4 Monthly (n=12) 4.8

1970 through 1983 0.15 Monthly (n=12) 1.8

1984 to 2003 (ongoing) 0.05 Monthly (n=12) 0.6

a. Actual periods by mm/dd/yyyy categorized according to year.

E.4.1.5.4 Energy Range

An estimate of the missed photon dose by energy range is possible based on the type of facility and 

predominant radionuclides such as intermediate (>100 keV) energies for all facilities handling 

activation and fission product nuclides, primarily lower (<100 keV) energy photons for plutonium 

facilities and lower energy photons for uranium fuel fabrication facilities. The recorded dose from the 

dosimeter response does not typically provide sufficient information to estimate discrete energy 

ranges. It is possible to examine the energy response characteristics of the respective multi-element 

dosimeters but this analysis does not recognize the substantial uncertainties present in the workplace 

associated with shielding, radiation scattering and mixed radiation fields.

E.4.1.6 Missed Neutron Dose Prior to 1971

Prior to 1971, the NTA film used at SRS did not provide a claimant-favorable estimate of the neutron 

dose. Essentially all SRS workers who may have received significant neutron dose did receive a 

measured photon dose. As such, using a ratio of the potential neutron dose to the measured photon 

dose is done as a claimant-favorable option to reconstruct an individual worker neutron dose. Table 

E-8 provides a summary of the recommended neutron to photon ratios for the different process areas. 

As can be determined from E-9, the recommended method to apply the ratio is as a log normal 

distribution using the geometric mean and geometric standard deviation. Caution, however, should 

be employed when using Table E-9 and consideration should be given to an individual energy 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 247 of 248

Table E-9. Neutron-to-photon ratios by SRS process.

Process Description/buildings

Operations Neutron to photon ratio

Begin End

Geometric 

mean (GM)

Geometric standard 

deviation (GSD)

Upper 

95th% 

Fuel fabrication Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and plutonium – aluminum alloy 

targets.

(321-M) (a) (a) (a)

Reactors During Reactor Operation: Low-level neutron exposure through shielding. In addition, some intermittent neutron 

exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987 0.18 2.52 0.82

(105K) 10/1954 1/1988

1992 7/1992

(105L) 7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium 

production

Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into production pucks. Work is 

primarily conducted in glove boxes with predominant close anterior exposure to workers. Radiation levels at the 

beginning of the process are fairly constant while levels at end of process are closely related to production levels.

Plutonium Finishing Process (221-H, B-Line) 11/1954 1988 0.91 2.84 5.05

Plutonium Finishing Process (221-F, B-Line) 10/1955 1988 0.36 2.52 1.65

Plutonium Storage

Production Control Laboratory (772-F) 1955

Plutonium Fuel Facility (235-F) 1955

Radionuclide 

production and 

calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS has produced several 

radionuclides including 238Pu, 239Pu, americium, curium, californium, etc. At the calibration facility radiological 

instruments are calibrated using a variety of sources with different neutron energy spectra. 

Californium Production Facility (773-A) 0.62 2.29 2.41

Calibration Facility (736-A) (b) (b) (b)

a. During Pu-Al target production, workers with potential for neutron exposure wore NTA film, thus LOD values listed in table 5.5.2-1 should be used. 

b. Neutron to photon ratio for the calibration facility is not expected to be greater than the ratio observed in the californium production facility. As a claimant 

favorable approximation the Californium Production facility ratio can be used. 

employee’s neutron to photon ratio if there is sufficient individual data. For example if an individual 

has recorded neutron dose either prior to or after the estimation period that has a higher neutron to 

photon ratio than what is provided in the table, then the higher neutron to photon ratio should be used 

provided that the work assignments were similar for the two time periods. The values in Table E-9 

should be used when there is known technical limitations in the neutron dosimetry technology or no 

neutron monitoring data. 

Due to potential for photon missed dose effecting the neutron to photon ratio, when applying the ratio 

in Table E-9, the photon dose should be the combination of both the measured photon dose 

(Dosimeter Dose) and the photon missed dose. When overestimating dose reconstructions, the 

upper 95% ratio can be used as a claimant-favorable worst case ratio and the associated neutron 

uncertainty can be omitted. In underestimating dose reconstructions, the geometric mean can be 

used as a best case ratio. 

E.4.1.7 Missed Neutron Dose After 1970 

Following SRS implementation of the TLND on January 1, 1971 recorded neutron dose has been 

reasonably accurate. As such, OCA-IG-001 guidance on missed dose should be followed in 

accordance with the LOD values presented in Table E-10. 

E.4.1.8 Organ Dose Equivalent

Once the adjusted photon and neutron doses have been calculated for each year, the values are used 

to calculate the organ dose distribution for the primary organ of interest identified in the claim. Table 

E-5 summarizes default workplace geometries. These can be used in case more applicable values 
Effective Date: 04/05/2005 Revision No. 03 Document No. ORAUT-TKBS-0003 Page 248 of 248

Table E-10. SRS TLND Limits of Detection.

Dosimeter type

Dates of use Exchange 

frequency

Reported 

LOD (mrem)

Estimated 

LOD (mrem)

Max. annual missed 

Begin End dose (mrem)

SRS Hoy TLND 1/1/1971 12/31/1994 Monthly (n=12) 10c 20 240

Panasonic TLND 1/1/1995 Present Monthly (n=12) 15 15 180

a. Based on Wilson et al (1990).

b. Correction for estimated 25% under response of NTA film for some facilities.

c. Minimum Recorded Dose not necessarily LOD.

Table E-11. Default exposure geometries to calculate organ dose.

Claim statusa Job categoryb Exposure geometry Percentagec

Non-compensable All AP 100%

Compensable - workers All AP 50%

ROT 50%

Compensable - supervisors All AP 50%

ISO 50%

a. Specific time spans for the various SRS facility operation can be found in the section 5.

b. More than one job category may be needed for longer-term employed workers. 

c. Apply this percentage to the dose conversion factor (NIOSH 2002a, Appendix B) to arrive at the 

total organ dose equivalent from the adjusted recorded dose. 

(NIOSH 2002) cannot be determined. A range of reasonable estimates may be evaluated to arrive at 

a claimant-favorable selection.

E.4.2 Parameter #2 

Parameter #2 is the standard deviation of the normal distribution for the organ dose. The individual 

dose result for each dosimeter exchange period will be available to calculate the mean and standard 

deviation for each year. If not available, the adjusted organ dose can be used for each year and a 

default standard deviation value used for parameter #2. 

E.4.3 Organ Dose Conversion Factors 

OCAS-IG-001 Appendix A (NIOSH 2002) contains a detailed discussion of the conversion of 

measured doses to organ dose equivalent, and Appendix B contains the appropriate dose conversion 

factors (DCFs) for each organ, radiation type, and energy range based on the type of monitoring 

performed. In some cases, simplifying assumptions are appropriate.