Page 1 of 189

DOE Review Release 10/11/2016

Feed Materials Production Center – 

Occupational Internal Dose

ORAUT-TKBS-0017-5 Rev. 02 

Effective Date: 09/30/2016

Supersedes: Revision 01 

Subject Expert(s): Karen S. Kent, Eugene Potter, Bryce Rich, Robert Morris, Billy P. Smith,

Melton Chew, and Matthew G. Arno

Document Owner 

Approval: Signature on File

Karen S. Kent, Document Owner

Approval Date: 09/29/2016

Concurrence: Signature on File

James P. Griffin, Deputy Project Director

Concurrence Date: 09/29/2016

Concurrence: Signature on File

Edward F. Maher, Objective 3 Manager

Concurrence Date: 09/29/2016

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 09/28/2016

Approval: Signature on File

David S. Sundin, Deputy Director, DCAS

Approval Date: 09/30/2016

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0017-5 Revision No. 02 Effective Date: 09/30/2016 Page 2 of 189

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

05/28/2004 00 New Technical Basis Document for the Fernald Environmental 

Management Project (FEMP) – Occupational Internal Dose. Initiated 

by Bryce L. Rich. First approved issue. Incorporates comments from 

OCAS and Task 5 reviews. Initiated by Bryce L. Rich.

07/22/2016 01 Revised in response to Task 5 dose reconstruction needs for 

clarification. Added Purpose, Scope, and Special Exposure Cohort 

sections to Section 1.0. Revised to incorporate responses to 

comments and responses in various white papers that were 

consolidated into ORAUT-RPRT-0052. Incorporated changes in 

response to formal internal and NIOSH review comments and 

Sanford Cohen & Associates matrix issues. Revised to reflect 

changes resulting from Special Exposure Cohort designations. 

Revised recycled uranium mixture and default uranium enrichment 

assumptions. Added 241Am and other radionuclides as recycled 

uranium contaminants. Added instructions for interpretation of in vivo 

data including methods for calculating thorium, thoron, unsupported 

radium, and raffinate exposures, including 226Ra from radon breath 

analysis results. Incorporated supporting documents pertaining to 

rationale for 400 ppb U Pu, uranium (ORAUT-OTIB-0078) and 

thorium coworker intakes, assignment of thorium dose based on 10% 

of the derived air concentrations, thoron exposures, and effective 

derived air concentrations. Updated and added references, as 

appropriate. Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by 

Karen S. Kent. 

09/30/2016 02 Revised to correct Recycled Uranium (RU) contaminant intakes 

(Bq/Bq/ U) in Table 5-34. Corrected Silo activity concentration for 

Ra-228 and Th-228 and modified silo activities from percentage to 

activity fractions in Table 5-13. Corrected “Ratios of silo 1 and silo 2 

nuclides to 226Ra” Table 5-16 and “Ratios of silos 1, 2, and 3 nuclides 

to 230Th” Table 5-17. Corrected typos. Incorporates formal internal 

review comments. No changes occurred as a result of formal NIOSH 

review. Training required: As determined by the Objective Manager. 

Initiated by Karen S. Kent.
Document No. ORAUT-TKBS-0017-5 Revision No. 02 Effective Date: 09/30/2016 Page 3 of 189

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 9

5.1 Introduction ............................................................................................................................ 12

5.1.1 Purpose...................................................................................................................... 13

5.1.2 Scope......................................................................................................................... 13

5.1.3 Special Exposure Cohort............................................................................................ 13

5.2 Background............................................................................................................................ 14

5.3 Radionuclides of Concern ...................................................................................................... 16

5.3.1 Uranium...................................................................................................................... 17

5.3.1.1 Uranium Enrichments................................................................................... 18

5.3.1.2 Chemical Forms and Compounds ................................................................ 18

5.3.1.3 Airborne Dust Potential ................................................................................ 18

5.3.2 Recycled Uranium ...................................................................................................... 21

5.3.2.1 Mass Balance, Material Flow, and Plutonium-239 Levels in Recycled 

Uranium in the Weapons Complex............................................................... 21

5.3.2.2 Major Contaminants in Recycled Uranium ................................................... 22

5.3.2.3 Tower Ash and Plutonium Out of Specification............................................. 22

5.3.2.4 Other Contaminants in Recycled Uranium ................................................... 25

5.3.2.5 Summary Default Values for Recycled Uranium........................................... 25

5.3.3 Thorium ...................................................................................................................... 26

5.3.3.1 Thorium Operations, 1954 to 1979............................................................... 26

5.3.3.2 Thorium Operations, 1979 to 2006............................................................... 26

5.3.3.3 Thoron ......................................................................................................... 28

5.3.3.3.1 Operations and Handling, 1954 to 1971 .................................... 29

5.3.3.3.2 Long-Term DOE Repository Storage, 1972 to 1985 .................. 29

5.3.3.3.3 Passive Storage, 1986 to 1987.................................................. 29

5.3.3.3.4 Remediation, 1988 to 2006 ....................................................... 29

5.3.4 Q-11, K-65, Domestic Uranium, and Raffinates .......................................................... 30

5.3.4.1 Radon Breath Analysis................................................................................. 31

5.3.4.2 Raffinate Exposures..................................................................................... 33

5.3.4.2.1 K-65 Raffinate Drum Transfer Workers, 1952 to 1956............... 33

5.3.4.2.2 Plant 2/3 Workers Handling or Processing Pitchblende 

Ores, 1954 to 1958.................................................................... 34

5.3.4.2.3 Plant 1, 2, 3, and 8 Workers Handling or Processing 

Yellowcake, 1959 to 1961 ......................................................... 34

5.4 Radiological Controls Program............................................................................................... 35

5.4.1 Air Monitoring Program............................................................................................... 36

5.4.2 Bioassay Program ...................................................................................................... 37

5.4.3 Uranium Urinalysis Program....................................................................................... 39

5.4.4 In Vitro Bioassay Procedures for Other Radionuclides ............................................... 40

5.4.5 Gross Beta Urine Sample Results, 1957 .................................................................... 40

5.4.6 In Vivo Bioassay......................................................................................................... 41

5.5 Dose Reconstruction.............................................................................................................. 47

5.5.1 Uranium Dose Assessments....................................................................................... 47

5.5.1.1 Bioassay Data.............................................................................................. 47
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5.5.1.2 Uranium Breathing Zone Results ................................................................. 47

5.5.1.3 Uranium Enrichment .................................................................................... 47

5.5.1.4 Recycled Uranium........................................................................................ 47

5.5.1.5 Chemical Form............................................................................................. 48

5.5.1.6 Unmonitored Uranium Workers.................................................................... 48

5.5.2 Thorium and Thoron Dose Assessments.................................................................... 49

5.5.2.1 Monitored Thorium Assessment Methods, 1979 to 2006.............................. 49

5.5.2.1.1 In Vivo Bioassay, 1979 to 1994 ................................................. 49

5.5.2.1.2 Breathing Zone Air Samples, 1995 to 2006 ............................... 51

5.5.2.1.3 In Vivo Bioassay and BZ Data, 1995 to 2006 ............................ 51

5.5.2.2 Assignment of Unsupported Radium............................................................ 51

5.5.2.3 Assignment of Unmonitored Thorium Dose.................................................. 52

5.5.2.3.1 Thorium Coworker Model, 1979 to 1989.................................... 52

5.5.2.3.2 Airborne Thorium Concentration, 1990 to 1994 ......................... 53

5.5.2.4 Assignment of Thoron Dose, 1954 to 2006 .................................................. 54

5.5.3 Raffinate Dose Assessments...................................................................................... 54

5.5.3.1 1952 to 1954................................................................................................ 54

5.5.3.2 1954 to 1958 (Pitchblende Ores).................................................................. 54

5.5.3.3 1959 to 1961 (Yellowcake)........................................................................... 55

5.5.4 Radon Exposure Assessment..................................................................................... 55

5.5.5 Site Closure................................................................................................................ 55

5.6 Attributions and Annotations .................................................................................................. 55

References ........................................................................................................................................ 58

Glossary ............................................................................................................................................ 71

ATTACHMENT A RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM ................. 75

ATTACHMENT B ANALYSIS OF POTENTIAL THORON EXPOSURES..................................... 84

ATTACHMENT C INTERNAL DOSIMETRY URANIUM COWORKER DATA............................ 114

ATTACHMENT D EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA.......... 131

ATTACHMENT E TECHNICAL BASIS FOR ASSIGNING DOSES FROM EFFECTIVE 

DERIVED AIR CONCENTRATION BREATHING ZONE RESULTS.............. 169

ATTACHMENT F THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 ............................. 175

ATTACHMENT G CLASS W THORIUM-232 INHALATION INTAKE RATE BASED ON 10% 

OF THE DERIVED AIR CONCENTRATION ................................................. 189

LIST OF TABLES 

TABLE TITLE PAGE

5-1 Contractors during the operational history of the Fernald site................................................. 15

5-2 Uranium enrichments and associated isotopes ...................................................................... 19

5-3 Uranium chemical forms......................................................................................................... 19

5-4 Plant processes, materials, and uranium dust release potential ............................................. 20

5-5 Average characteristics of RU and its major contaminants..................................................... 22

5-6 RU contaminant levels by process subgroups........................................................................ 23

5-7 Other radioactive contaminants in RU from Hanford .............................................................. 25
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5-8 RU enrichment.......................................................................................................................25

5-9 Age of plutonium mixture in RU.............................................................................................. 26

5-10 RU contaminant mass concentrations.................................................................................... 26

5-11 Inventory of stored thorium, 1987........................................................................................... 27

5-12 Post-SEC thorium activities and regulatory requirements, 1988 to 2006 ................................ 28

5-13 Isotopic composition of K-65 silos 1, 2, and 3......................................................................... 31

5-14 Radon breath analysis statistical parameters ......................................................................... 32

5-15 Radium intake rate based on radon breath analysis data....................................................... 33

5-16 Ratios of silo 1 and silo 2 nuclides to 226Ra ............................................................................ 33

5-17 Ratios of silos 1, 2, and 3 nuclides to 230Th ............................................................................ 34

5-18 Air monitoring analysis sheet codes ....................................................................................... 37

5-19 Air monitoring methods with applicable MDLs ........................................................................ 37

5-20 Chemical toxicity and associated limits .................................................................................. 38

5-21 Urinary uranium averages summary ...................................................................................... 39

5-22 Routine uranium in vitro bioassay capabilities ........................................................................ 40

5-23 Fernald uranium urinalysis sample type codes....................................................................... 41

5-24 Fernald sample type coding system and personnel assignment codes .................................. 42

5-25 MDAs for nonuranium radionuclides ...................................................................................... 43

5-26 Typical reporting of internal dose determined by direct lung count ......................................... 43

5-27 Typical MIVRML counting schedule at Fernald in the 1970s .................................................. 44

5-28 In vivo bioassay MDAs for the lung ........................................................................................ 45

5-29 Summary of MIVRML chest count MDAs by uranium enrichment, 1968 to 1989.................... 46

5-30 Summary of IVEC chest count MDAs by uranium enrichment, 1989 to 2002 ......................... 46

5-31 Routine uranium urinalysis capabilities................................................................................... 47

5-32 Default uranium enrichment ................................................................................................... 47

5-33 RU contaminant intakes per unit mass of uranium ................................................................. 48

5-34 RU contaminant intakes per unit activity of uranium............................................................... 48

5-35 Methods for thorium dose determination ................................................................................ 49

5-36 Summary of bias corrections for 228Ac and 212Pb in chest counts............................................ 50

5-37 Thorium chest count MDAs (nCi), 1979 to 2001..................................................................... 50

5-38 Custom nuclides from EDAC BZ results................................................................................. 52

5-39 Thorium coworker intake rates, 1979 to 1989......................................................................... 52

5-40 232Th inhalation and ingestion rates based on 10% DAC........................................................ 53

5-41 Thoron worker exposure ........................................................................................................ 54

A-1 Timeline of PGDP shipments to Fernald ................................................................................ 78

A-2 Totals of recycled feeds to Fernald ........................................................................................ 78

B-1 Historical thorium mass balance............................................................................................. 86

B-2 Thorium storage inventory, 1987............................................................................................ 90

B-3 PAEC for thoron and progeny ................................................................................................ 90

B-4 Building dimensions and ventilation systems.......................................................................... 91

B-5 Summary of thoron exposure estimates................................................................................. 96

B-6 Thoron exposure....................................................................................................................97

B-7 Summary of the 1996 through 1998 thoron measurement data............................................ 100

B-8 Summary of WL measurements........................................................................................... 103

B-9 Three high measurements near Plant 8 silo and bins in 1986 .............................................. 104

B-10 Locations and numbers of samples...................................................................................... 109

B-11 Thoron and thoron daughter results ..................................................................................... 110

C-1 Criteria applied to extract for uranium bioassay data............................................................ 116

C-2 Reporting level..................................................................................................................... 117

C-3 Summary of uranium urinary excretion rate analyses, 1952 to 2006 .................................... 118

C-4 Derived uranium intake rates for type F materials, 1952 to 2006.......................................... 121

C-5 Derived uranium intake rates for type M materials, 1952 to 2006......................................... 121

C-6 Derived uranium intake rates for type S materials, 1952 to 2006.......................................... 122
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D-1 Summary of bias and DL for 228Ac and 212Pb for each year .................................................. 134

E-1 Count of EDAC results in HIS-20 ......................................................................................... 170

E-2 Activity fractions for the BL-13 EDAC................................................................................... 171

E-3 Isotopic composition of K-65 silo 1....................................................................................... 172

E-4 Isotopic composition of K-65 silo 2....................................................................................... 172

E-5 Activity fractions for the RT-210 EDAC................................................................................. 174

F-1 Bias for 212Pb in chest .......................................................................................................... 184

F-2 In vivo lung counting statistics and 232Th lung burdens......................................................... 185

F-3 Derived thorium intakes, 1979 through 1989........................................................................ 186

LIST OF FIGURES

FIGURE TITLE PAGE

B-1 232Th decay diagram, including the 220Rn (thoron) decay chain .............................................. 87

B-2 Decay and buildup of 228Th and progeny after processing...................................................... 88

B-3 Thoron comparison, 1996 to 1998........................................................................................ 106

B-4 Measurements and lognormal fits, 1996............................................................................... 107

B-5 WL comparisons, 1986 through 2003................................................................................... 108

B-6 GA center of rod storage area.............................................................................................. 111

B-7 GA north end of rod storage area......................................................................................... 112

B-8 GA TNT and sump cake storage area.................................................................................. 112

B-9 GA wet area, main floor ....................................................................................................... 113

B-10 GA Wet area, top deck......................................................................................................... 113

C-1 Change in lognormal fit for various CLs................................................................................ 118

C-2 Predicted values versus measured results, 01/01/1952 to 12/31/1993, 50th percentile, 

type F................................................................................................................................... 123

C-3 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 50th percentile, 

type F................................................................................................................................... 123

C-4 Predicted values versus measured results, 01/01/1952 to 12/31/1993, 84th percentile, 

type F................................................................................................................................... 123

C-5 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 84th percentile, 

type F................................................................................................................................... 124

C-6 Predicted values versus measured results, 01/01/1952 to 12/31/1993, 50th percentile, 

type M.................................................................................................................................. 124

C-7 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 50th percentile, 

type M.................................................................................................................................. 124

C-8 Predicted values versus measured results, 01/01/1952 to 12/31/1993, 84th percentile, 

type M.................................................................................................................................. 125

C-9 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 84th percentile, 

type M.................................................................................................................................. 125

C-10 Predicted values versus measured results, 01/01/1952 to 12/31/1954, 50th percentile, 

type S................................................................................................................................... 125

C-11 Predicted values versus measured results, 01/01/1955 to 12/31/1955, 50th percentile, 

type S................................................................................................................................... 126

C-12 Predicted values versus measured results, 01/01/1956 to 12/31/1956, 50th percentile, 

type S................................................................................................................................... 126

C-13 Predicted values versus measured results, 01/01/1957 to 12/31/1961, 50th percentile, 

type S................................................................................................................................... 126

C-14 Predicted values versus measured results, 01/01/1962 to 12/31/1966, 50th percentile, 

type S................................................................................................................................... 127
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C-15 Predicted values versus measured results, 01/01/1967 to 12/31/1993, 50th percentile, 

type S................................................................................................................................... 127

C-16 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 50th percentile, 

type S................................................................................................................................... 127

C-17 Summarized results of the predicted values versus the excluded results, 01/01/1952 to 

12/31/2006, 50th percentile, type S...................................................................................... 128

C-18 Predicted values versus measured results, 01/01/1952 to 12/31/1956, 84th percentile, 

type S................................................................................................................................... 128

C-19 Predicted values versus measured results, 01/01/1957 to 12/31/1961, 84th percentile, 

type S................................................................................................................................... 128

C-20 Predicted values versus measured results, 01/01/1962 to 12/31/1966, 84th percentile, 

type S................................................................................................................................... 129

C-21 Predicted values versus measured results, 01/01/1967 to 12/31/1980, 84th percentile, 

type S................................................................................................................................... 129

C-22 Predicted values versus measured results, 01/01/1981 to 12/31/1993, 84th percentile, 

type S................................................................................................................................... 129

C-23 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 84th percentile, 

type S................................................................................................................................... 130

C-24 Summarized results of the predicted values versus the excluded results, 01/01/1952 to 

12/31/2006, 84th percentile, type S...................................................................................... 130

D-1 Normal probability plot of 212Pb chest count results for 1978 ................................................ 133

D-2 Normal probability plot of 212Pb chest count results for 1978 after adjusting for bias............. 133

D-3 Normal probability plot of 212Pb chest count results for 1978 after adjusting for bias............. 134

D-4 Plot of 228Ac versus 212Pb for 1978 where the levels of both nuclides are above the DL ..... 135

D-5 Normal probability plots of 212Pb and 228Ac, 1978 ................................................................. 136

D-6 Normal probability plots of 212Pb and 228Ac, 1979 ................................................................. 137

D-7 Normal probability plots of 212Pb and 228Ac, 1980 ................................................................. 138

D-8 Normal probability plots of 212Pb and 228Ac, 1981 ................................................................. 139

D-9 Normal probability plots of 212Pb and 228Ac, 1982 ................................................................. 140

D-10 Normal probability plots of 212Pb and 228Ac, 1983 ................................................................. 141

D-11 Normal probability plots of 212Pb and 228Ac, 1984 ................................................................. 142

D-12 Normal probability plots of 212Pb and 228Ac, 1985 ................................................................. 143

D-13 Normal probability plots of 212Pb and 228Ac, 1986 ................................................................. 144

D-14 Normal probability plots of 212Pb and 228Ac, 1987 ................................................................. 145

D-15 Normal probability plots of 212Pb and 228Ac, 1988 ................................................................. 146

D-16 Both radionuclides above the DL, 1978................................................................................ 147

D-17 Either radionuclides above the DL, 1978.............................................................................. 148

D-18 Both radionuclides above the DL, 1979................................................................................ 149

D-19 Either radionuclides above the DL, 1979.............................................................................. 150

D-20 Both radionuclides above the DL, 1980................................................................................ 151

D-21 Either radionuclides above the DL, 1980.............................................................................. 152

D-22 Both radionuclides above the DL, 1981................................................................................ 153

D-23 Either radionuclides above the DL, 1981.............................................................................. 154

D-24 Both radionuclides above the DL, 1982................................................................................ 155

D-25 Either radionuclides above the DL, 1982.............................................................................. 156

D-26 Both radionuclides above the DL, 1983................................................................................ 157

D-27 Either radionuclides above the DL, 1983.............................................................................. 158

D-28 Both radionuclides above the DL, 1984................................................................................ 159

D-29 Either radionuclides above the DL, 1984.............................................................................. 160

D-30 Both radionuclides above the DL, 1985................................................................................ 161

D-31 Either radionuclides above the DL, 1985.............................................................................. 162

D-32 Both radionuclides above the DL, 1986................................................................................ 163

D-33 Either radionuclides above the DL, 1986.............................................................................. 164
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D-34 Both radionuclides above the DL, 1987................................................................................ 165

D-35 Either radionuclides above the DL, 1987.............................................................................. 166

D-36 Both radionuclides above the DL, 1988................................................................................ 167

D-37 Either radionuclides above the DL, 1988.............................................................................. 168

E-1 222Rn decay chain showing half-lives.................................................................................... 174

F-1 Natural thorium decay chain showing isotopes and half-lives............................................... 178

F-2 Relative activity of the isotopes in the natural thorium decay chain as a function of time 

after chemical purification..................................................................................................... 179

F-3 228Th activity after a single chemical purification................................................................... 180

F-4 228Ra and 224Ra ingrowth after thorium chemical separation................................................. 181

F-5 228Th activity after a second chemical purification at its lowest equilibration point 

(4.55 years) after the first purification ................................................................................... 181

F-6 224Ra and 228Ra activity after a second chemical purification when 228Th was at its lowest 

equilibration point (4.55 years) after the first purification ...................................................... 182

F-7 228Th activity after a third chemical purification when at its lowest equilibration point 

(7.09 years) after the first purification ................................................................................... 182

F-8 224Ra and 228Ra activity after the third chemical separation .................................................. 183

F-9 50th- and 84th-percentile 232Th chest burdens, 1979 to 1988............................................... 185

F-10 Predicted thorium lung burden calculated using the three-separation model to derive 

thorium intake rates compared with measured chest burden, 1979 to 1989, 

50th percentile, Type M........................................................................................................ 186

F-11 Predicted thorium lung burden calculated using the three-separation model to derive 

thorium intake rates compared with measured chest burden, 1979 to 1989, 

84th percentile, Type M........................................................................................................ 187

F-12 Predicted thorium lung burden calculated using the three separation model to derive 

thorium intake rates compared with measured chest burden, 1979 to 1989, 

50th percentile, Type S ........................................................................................................ 187

F-13 Predicted thorium lung burden calculated using the three separation model to derive 

thorium intake rates compared with measured chest burden, 1979 to 1989, 

84th percentile, Type S ........................................................................................................ 188
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMC African Metal Company

AWE Atomic Weapons Employer

Bq becquerel

BZ breathing zone

CEDE committed effective dose equivalent

CFR Code of Federal Regulations

Ci curie

CL censoring level

cm centimeter

d day

DAC derived air concentration

DCAS Division of Compensation Analysis and Support

DL decision level

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DWE daily weighted (average) exposure

EDAC effective derived air concentration

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ERDA U.S. Energy Research and Development Administration

EU enriched uranium

FEMP Fernald Environmental Management Project

FERMCO Fernald Environmental Restoration Management Corporation

FMPC Feed Materials Production Center

ft foot

g gram

GA general area

gal gallon

GDP gaseous diffusion plant

GM geometric mean

GSD geometric standard deviation

HASL Health and Safety Laboratory

hr hour

ICP-MS inductively coupled plasma mass spectrometry

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

IREP Interactive RadioEpidemiological Program

IVEC In Vivo Examination Center

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

KPA kinetic phosphorescence analysis
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L liter

lb pound

LEU low-enriched uranium

LOOW Lake Ontario Ordnance Works

m meter

MAC maximum allowable concentration

MCW Mallinckrodt Chemical Works

MDA minimum detectable activity

MDL minimum detection level

MeV megaelectron-volt, 1 million electron-volts

mg milligram

MIVRML Mobile In Vivo Radiation Monitoring Laboratory

mL milliliter

mo month

MPLB maximum permissible lung burden

mrem millirem

MT metric ton

MTTh metric ton of thorium

MTU metric ton of uranium

NCG NLO Concentration Guide (same as MAC for NLO only)

nCi nanocurie

ng nanogram

NIOSH National Institute for Occupational Safety and Health

NLO National Lead Company of Ohio

NOCTS NIOSH-DCAS Claims Tracking System

NTS Nevada Test Site

NU natural uranium

OPOS one person–one sample

ORAU Oak Ridge Associated Universities

OSDF Onsite Disposal Facility

PAEC potential alpha energy concentration

PAEE potential alpha energy exposure

pCi picocurie

PGDP Paducah Gaseous Diffusion Plant

PID Principal Internal Dosimetrist

POOS plutonium out of specification

ppb parts per billion

ppm parts per million

PUREX plutonium-uranium extraction

RCRA Resource Conservation and Recovery Act of 1976

RCS Radon Control System

RF release fraction

ROS regression on order statistic

RT radon treatment

RU recycled uranium

s second

SC&A Sanford Cohen & Associates
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SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

TBD technical basis document

TOP Thorium Overpack Project

TRU transuranic

TTA Transfer Tank Area

TWOPOS time-weighted one person–one sample

U.S.C. United States Code

WEMCO Westinghouse Environmental Management Company of Ohio

WL working level

WLM working level month

yr year

μCi microcurie

μg microgram

α alpha particle

§ section or sections
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5.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historical background information and guidance to assist in the preparation of 

dose reconstructions at particular Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

facilities or categories of DOE or AWE facilities. They will be revised in the event additional relevant 

information is obtained about the affected DOE or AWE facility(ies). These documents may be used 

to assist NIOSH staff in the evaluation of Special Exposure Cohort (SEC) petitions and the completion 

of the individual work required for each dose reconstruction.

In this document the word “facility” is used to refer to an area, building, or group of buildings that 

served a specific purpose at a DOE or AWE facility. It does not mean nor should it be equated to an 

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7384l(5) 

and (12) of the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA), respectively. An AWE facility means “a facility, owned by an atomic weapons employer, 

that is or was used to process or produce, for use by the United States, material that emitted radiation 

and was used in the production of an atomic weapon, excluding uranium mining or milling.” 42 U.S.C. 

§ 7384l(5). On the other hand, a DOE facility is defined as “any building, structure, or premise, 

including the grounds upon which such building, structure, or premise is located ... in which 

operations are, or have been, conducted by, or on behalf of, the [DOE] (except for buildings, 

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” 

and with regard to which DOE has or had a proprietary interest, or “entered into a contract with an 

entity to provide management and operation, management and integration, environmental 

remediation services, construction, or maintenance services.” 42 U.S.C. § 7384l(12). The Department 

of Energy (DOE) determines whether a site meets the statutory definition of an AWE facility and the 

Department of Labor (DOL) determines if a site is a DOE facility and, if it is, designates it as such.

Accordingly, a Part B claim for benefits must be based on an energy employee’s eligible employment 

and occupational radiation exposure at a DOE or AWE facility during the facility’s designated time 

period and location (i.e., covered employee). After DOL determines that a claim meets the eligibility 

requirements under EEOICPA, DOL transmits the claim to NIOSH for a dose reconstruction. 

EEOICPA provides, among other things, guidance on eligible employment and the types of radiation 

exposure to be included in an individual dose reconstruction. Under EEOICPA, eligible employment 

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures be incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction.

NIOSH does not consider the following types of exposure as those incurred in the performance of 

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees (NIOSH 2010):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons
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5.1.1 Purpose

This technical basis document (TBD) is part of the overall site profile for the Feed Materials 

Production Center (FMPC) site in Fernald, Ohio, which describes plant facilities and processes, 

historical information, and internal exposure data in relation to dose reconstruction for Fernald site 

workers. This TBD provides a technical basis for evaluating the occupational internal dose for 

EEOICPA claimants who were employed at the site. Since operations ended in 1988, the Fernald site 

has been variously named the Fernald Environmental Management Project (FEMP), the Fernald 

Closure Project, and is now the Fernald Preserve. This document uses FMPC or Fernald for 

convenience.

Occupational internal dose refers to radiation exposures workers received as a result of intakes of 

radionuclides from site operations as well as radon exposures, which might have been enhanced due 

to the nature of those operations. FMPC employees who did not have bioassay data or who were not 

fully monitored for internal exposures may be assigned coworker intakes, in addition to environmental 

doses, as applicable.

5.1.2 Scope

This TBD describes the methods for assessment of internal radiation dose to workers at Fernald. The 

radionuclides of concern are primarily the isotopes of uranium and their progeny. Information is 

included on the various enrichments of uranium that were handled or processed and on the 

contaminants in recycled uranium (RU). The various chemical and physical forms of these materials 

are also covered. Because FMPC also processed some natural thorium, this TBD includes 

information on thorium and its decay products, including thoron and radium, and radionuclides 

associated with raffinate exposure.

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 5.6.

5.1.3 Special Exposure Cohort

The Secretary of the U.S. Department of Health and Human Services has designated three classes of 

employees at Fernald as additions to the SEC:

Thorium, 1968 through 1978

All employees of DOE, its predecessor agencies, and their contractors, or subcontractors 

who worked at the Feed Materials Production Center in Fernald, Ohio, from January 1, 

1968 through December 31, 1978, for a number of work days aggregating at least 250 

work days, occurring either solely under this employment, or in combination with work 

days within the parameters established for one or more other classes of employees 

included in the SEC (Sebelius 2012). 

It was determined that NIOSH lacked sufficient information to allow it to estimate with sufficient 

accuracy the potential internal doses from exposure to thorium, to which employees at this facility may 

have been subjected (Sebelius 2012).

Thorium, 1954 through 1967

All employees of the DOE, its predecessor agencies, and their contractors and 

subcontractors who worked at the Feed Materials Production Center in Fernald, Ohio, 

from January 1, 1954, through December 31, 1967, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment, or in 
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combination with work days within the parameters established for one or more other 

classes of employees included in the Special Exposure Cohort (Sebelius 2013a). 

It was determined that NIOSH lacked sufficient information to allow it to estimate with sufficient 

accuracy the potential internal doses from exposure to thorium, to which employees working at this 

facility may have been subjected.

Uranium, 1951 through 1983 [applies only to subcontractors]

All employees of the Feed Materials Production Center in Fernald, Ohio, who were not 

employed by National Lead of Ohio, NLO, or the Department of Energy or its 

predecessor agencies, who worked at FMPC from January 1, 1951, through December 

31, 1983, for a number of work days aggregating at least 250 work days, occurring either 

solely under this employment, or in combination with work days within the parameters 

established for one or more other classes of employees included in the Special Exposure 

Cohort (Sebelius 2013b). 

It was determined that NIOSH lacked sufficient information to allow it to estimate with sufficient 

accuracy the potential internal doses from exposure to uranium, to which employees of 

subcontractors who worked at the FMPC may have been subjected.

This dose reconstruction infeasibility for the period from 1951 through 1983 applies only to 

subcontractors. NIOSH has access to an electronic dataset that contains the results of the uranium 

urinalysis bioassay program for all years of FMPC operations, and the overwhelming majority of 

employees of the prime contractor National Lead Company of Ohio (later named NLO) have results in 

the bioassay dataset. However, the dataset does not contain bioassay results for employees of 

companies other than the prime contractor (i.e., non-prime contractor employees). NIOSH has 

obtained a limited number of bioassay samples from non-prime contractor employees through data 

captures, but cannot be certain that all non-prime contractor employee bioassay data were retained 

by the site or captured by data capture efforts. In addition, there are some reasons to conclude that 

the prime contractor did not consistently evaluate whether non-prime contractor employees should be 

monitored for radiation exposure because of the transitory nature of their work (Sebelius 2013b).

Dose reconstruction guidance in this document for the period from January 1, 1954, to December 31, 

1983, is presented to provide a technical basis for partial dose reconstructions for claims not 

compensated under the SEC (i.e., nonpresumptive cancers and SEC employment less than 

250 days). Although it is not possible to completely reconstruct internal radiation doses for all workers 

for this period, NIOSH has determined, and the U.S. Department of Health and Human Services has 

concurred, that it is feasible to reconstruct external radiation doses for all FMPC workers for this 

period (Sebelius 2013b).

5.2 BACKGROUND

Construction at FMPC began in 1951 on a 1,050-acre site near the small rural community of Fernald, 

Ohio. Table 5-1 lists the five contractors to DOE and its predecessor agencies [i.e., the U.S. Atomic 

Energy Commission (AEC) and the U.S. Energy and Research Development Administration (ERDA)] 

that operated Fernald during its operational history.

The primary missions at FMPC were the processing of uranium ores to high-quality finished metal 

products and the production of thorium metal parts and thorium feedstock for processes at other DOE 

sites. The 10 plants that constituted the Fernald facility processed a variety of natural uranium (NU) 

and low-enriched uranium (LEU) ores and scrap materials for the creation of high-purity uranium 

metal parts. The thorium production included metal parts and feedstock for the weapons production 

programs. Operations involved thousands of metric tons of ores, dry powder products, and corrosive
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Table 5-1. Contractors during the operational history of the Fernald site.

Contractors Dates Time

NLO October 1951 to January 1986 34+ yr

Westinghouse Materials Company of Ohio (WEMCO) January 1986 to December 1992 7 yr

Fernald Environmental Restoration Management Corporation 

(FERMCO),a Fluor Daniel Fernald, and Fluor Fernald

December 1992 to 2006 

(transferred to DOE Office of 

Legacy Management)

14+ yr

a. FERMCO received the first DOE Environmental Restoration Management contract for FEMP.

chemicals in processes that were inherently dusty, which produced an environment with internal 

intake potential.

Due to the inherent nature of the processes, the limitations of the ventilation and material confinement 

systems, and the volume (and mass) of the materials, significant environmental and in-plant releases 

of radioactive materials occurred during FMPC operations. The work environment included a chronic 

potential for internal exposure, as demonstrated by the comprehensive air monitoring program and 

the urine sampling program for uranium (documented by air samples and uranium urinalysis). During 

the early years, plant workers were routinely required to wear respiratory protection because of 

significant radioactive dust levels approaching or exceeding FMPC airborne alpha activity guidelines 

referred to as maximum allowable concentrations (MACs) [or NLO Concentration Guide (NCG)] 

depending on the period. In addition, over the operational history of Fernald, metric tons of uranium 

and thorium products were released from the ventilation stacks to the environment.

The number of personnel at Fernald peaked at 2,891 in 1956 and slowly declined to 538 in 1979; an 

estimated total of 7,300 personnel worked at Fernald from 1953 through 1990. Given the site 

conditions of routine measurable air activity in the workplace and significant routine releases to the 

surrounding areas, all of these personnel had some potential for internal exposure from either being in 

process or other areas associated with loose contamination, or as a result of environmental releases. 

Additional radionuclides were introduced from other feedstock sources: radium and thorium from 

pitchblende ores from the Belgian Congo, transuranic (TRU) elements and fission products from RU, 

and thorium.

Pitchblende ores from the Belgian Congo (known as Q-11 material), which had unusually high radium 

and thorium activities, were processed from 1953 to 1955 as part of the uranium ores processing 

program. The liquid extraction column raffinates (known as K-65 material) from uranium extraction 

processes were stored in two dedicated concrete silos (by agreement, they belonged to the Belgian 

government). Additional pitchblende chemical extraction residues from another AEC site, Mallinckrodt 

Chemical Works (MCW), were added to the silos starting from the time of completion of the silos in 

1953 and continuing until 1958. The materials in these silos became the property and responsibility of 

the United States when the lease agreement expired in 1983. This material remained on the Fernald 

site in the K-65 silos until March 2005 (Fluor Fernald 2006, p. 104) and represented a continuing 

internal exposure potential (for any operations in which direct contact with these residues was 

required) from the unusually high concentrations of 226Ra and its progeny, 210Pb and 210Po. 

These three radionuclides comprise approximately 90% of the total activity in the K-65 materials. The 

uranium progeny 230Th was also present in significant quantities (7% to 10%) in this location (see 

Table 5-13 later in this document). Historical environmental releases from this location, with the 

resultant internal dose potential, occurred from the ingrown radon gases and the progeny.

Fernald processed thorium from 1954 to 1979 and was the national thorium materials repository for 

DOE starting in 1972. About two-thirds of the material in the repository was processed and stored at 

Fernald, while the remainder originated at other DOE facilities. The thorium was stored as metal and 

stable chemical compounds [e.g., thorium gel, Th(OH)4] containing primarily 232Th and its long-lived 

progeny 228Th with all of the progeny present, which included the short-lived 220Rn (thoron gas). The 
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original thorium inventory was contained in approximately 15,000 storage containers of various sizes. 

The January 2001 inventory was verified to be approximately 500,000 lb stored in approximately 200 

metal boxes (Tomes 2001). Internal thorium exposure potential is associated with all of the thorium 

processes and during handling and repackaging of the thorium in storage drums.

The initial uranium processing started in 1951 with virgin stock from uranium mines and mills. RU 

was received and processed at Fernald as early as 1961. RU is that which has been used in a 

reactor environment (such as fuel elements or TRU production targets) and then processed to 

separate the unspent uranium from fission products, activation products, and TRU elements. RU 

contains trace quantities of plutonium isotopes, 237Np, and 99Tc as the primary contaminants. Some 

waste materials from the uranium enrichment facilities in the form of tower ashes were sent to Fernald 

for uranium recovery and had higher than average TRU contaminant concentrations as a result of 

enriching RU material. The levels of contaminants in these materials containing RU were known 

before receipt and blended with other uranium feedstock for processing. Personnel who were 

exposed to uranium contamination could also be exposed to the RU contaminants, which could have 

contributed to unmonitored internal exposure.

The original health and safety program was conducted with an industrial hygiene emphasis, based on 

uranium heavy-metal toxicology, with awareness and utilization of technology from other AEC site 

programs (e.g., Y-12 programs). Basic changes in the radiological protection program occurred with 

the contractor change in 1986 with the addition of radiation safety staff and increased analyses of all 

radiological hazards.

5.3 RADIONUCLIDES OF CONCERN

The following are the primary radionuclides that could have led to internal doses during the production 

history of Fernald:

• Uranium isotopes 234U, 235U, 236U, and 238U (although 236U is listed as an expected uranium 

component, is found primarily in RU and represents <1% of the dose from exposure to RU);

• Uranium progeny 230Th, 226Ra, 222Rn, 231Pa, and 227Ac (230Th was present primarily as a 

uranium decay product);

• RU contaminants 238Pu, 239Pu, 240Pu, 241Pu, 241Am, 237Np, 99Tc, 232Th, 228Th, 90Sr, 103/106Ru, and 

95Zr/95Nb;

• Thorium isotopes 232Th and 228Th; and 230Th

• Thorium progeny 226Ra, 228Ac, 224Ra, 220Rn, and 212Pb (228Ac and 212Pb are listed because they 

were used in in vivo counting to quantify the parents 232Th and 228Th).

Other infrequent, and perhaps one-of-a-kind, operations might have exposed workers to additional 

radionuclides. One example was the changing of the ion-getter vacuum pump blades in a CockcroftWalton neutron generator, which was performed on June 24, 1966. An approved procedure was 

developed and followed. Supplied air and protective clothing were worn, and the pump was 

disassembled in a hood. The workers might have been exposed to tritium combined with titanium 

inside the pump. A series of tritium urine bioassay samples were collected and evaluated at Mound 

Laboratory and indicated no positive results (Ross and Starkey 1966). If unusual bioassay results are 

found in claimant files, the Site Lead and Principal Internal Dosimetrist (PID) should be consulted.
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5.3.1 Uranium

The FMPC production facilities consisted of 10 production plants, each of which produced a product 

for the next process in a series of steps, terminating in high-purity uranium metal stock or parts. 

Operations in each major facility are briefly described below.

The Pilot Plant, as an operational prototype of the entire production process, began operations in 

October 1951. The purpose was to develop performance data for designing large-scale equipment for 

the rest of the plant. At the same time, limited quantities of uranium metal were produced during the 

initial Pilot Plant operation. In addition, UF6 was converted to UF4. On February 14, 1966, an 

accidental release from the Pilot Plant of approximately 1,200 kg of uranium occurred during 

conversion operations that resulted in elevated personnel exposures (Warner 1966). The enrichment 

is unknown, so 2% should be assumed for any worker identified as a subject of this incident. In 

addition, more than 70% of the thorium at FMPC was handled and processed from 1964 to 1979 in 

the Pilot Plant.

Plant 1, the Sampling Plant, started operations in December 1953 and was later designated the 

official AEC sampling station for uranium along with isotopic assays of uranium ores and concentrates 

(usually U3O8). The sampling process frequently involved the dusty operations of mixing, blending,

etc. of large quantities of uranium and thorium materials.

Plants 2 and 3, the Ore Refinery Plants, began operations in December 1953 and achieved full-scale 

production early in 1954. The chemical processing in these facilities involved a three-step operation 

that began with nitric acid leaching of uranium from dry solid feed materials. This was followed by 

solvent extraction processing using a modified plutonium-uranium extraction (PUREX)1 process to 

produce a high-purity solution of uranyl nitrate (UNH). The final step was a high-temperature thermal 

decomposition of the UNH solution to produce UO3 product in dry, solid form and nitric oxides as 

effluent. After startup, these two plants generally operated as a single plant due to the integral 

process of the operations. During operations, the Plant 2/3 complex was a single large plant (2A), 

which was surrounded by several small buildings and tank farms that performed various support 

functions. For example, the Hot Raffinate Building (3E) was used to filter insoluble material from UNH 

and to process raffinates.

Plant 4, the Green Salt Plant, which converted UO3 to UF4 (green salt), began operating in October 

1953 and achieved full-scale operation in 1954. The two-step process involved hydrogen reduction of 

UO3 powder to UO2 solids and then to UF4 by the addition of anhydrous hydrofluoric acid. These 

processes mixed, agitated, and transferred metric tons of solid uranium and produced some of the 

higher levels of airborne radioactivity at the site. Some air sample data sheet information indicates 

that at least a limited amount of thorium might have been processed in Plant 4.

Plant 5, the Metals Production Plant, provided a chemical conversion of UF4 powder to a uranium 

metal “derby” by a thermite furnace reduction process with magnesium metal. This facility began 

operations late in 1953. This process concentrated the very small concentrations of remaining TRU 

elements and uranium progeny in the MgF2 slag. The transport and mixing of feed materials and 

MgF2 slag led to inhalation exposure potentials that were also in the higher level categories at the site.

Plant 6, the Metals Fabrication Plant, produced metal parts in rolling mills and machining lathes (as 

did Plant 9). Plant 6 began operations in 1953. Uranium metal fires were common, resulting in 

elevated airborne uranium concentrations.

 

1 The PUREX process is a chemical extraction technology for the separation of radionuclides from uranium and 

was used at several DOE sites in a variety of applications.
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Plant 7, the Hexafluoride Reduction Plant, was a chemical process facility that operated from 1954 to 

1957 and provided an additional source of UF4 by hydrogen reduction of UF6. UF6 was converted by 

temperature control to the gaseous phase, which increased the potential for uranium inhalation 

exposures.

Plant 8, the Scrap Recovery Plant, provided a scrap uranium recovery function. Plant 8 began 

operations in 1954 and was also ranked in the higher air activity level areas at the site. Thorium 

scraps and residues were processed in 1966, and from 1969 to 1971.

Plant 9, the Special Products Plant, included casting of uranium metal and high-purity recycled metal 

scraps into ingots. Operations began late in 1953. Thorium was processed as metal and briquettes 

in 1954 and 1955.

5.3.1.1 Uranium Enrichments

Late in 1964, FMPC provided the first production of 1.95% 235U billets for the Hanford Site. During the 

following production years, uranium was processed in a variety of enrichments ranging from depleted 

to as high as 20% (DOE 1996). The quantities of enriched material above 2% were not documented, 

but were qualitatively reported to be small and/or insignificant in total mass. The reported highest 

enrichment level processed in quantity was 2%. As another point of reference, a 1961 to 1984 history 

(Dolan and Hill 1988) of the average uranium enrichments in dust collector stack discharges 

demonstrates that 235U enrichments ranged from 0.20% to 1.68% with an average of 0.70%. Of the 

total quantity of uranium that was received and processed at FMPC, less than 25% was enriched 

above normal (60,181 MTU of the total 246,683 MTU). Approximately 95% (208 g of the total 218 g) 

of the 239Pu that was received at the site came in the enriched uranium (EU) receipts (DOE 2000a).

Table 5-2 lists the primary assumptions for FMPC uranium enrichments and the isotopes associated 

with these enrichments.

In the absence of specific enrichment information, and considering the above available data on 

processing experience of uranium enrichments at FMPC, the default assumption for periods after 

1964 is 2% enrichment for bioassay data in milligram quantities of uranium. A 1% uranium 

enrichment can be assumed through 1964 to ensure results are favorable to claimants (NIOSH 2008, 

NIOSH 2014a).

5.3.1.2 Chemical Forms and Compounds

There are approximately seven steps in the process of conversion of uranium ore or other scrap 

recovery materials to metallic uranium. Those steps produce a number of compounds, each of which 

has specific chemical characteristics that are associated with different internal exposure parameters. 

Each of the compounds in Table 5-3 was handled in metric ton quantities. Most of the compounds 

were dry powder or granular in form and represented a dust hazard potential as the material was 

processed, transferred, and otherwise handled. Because all solubilities of uranium were handled in 

large quantities, all must be considered when assessing an intake unless specific information is 

available.

5.3.1.3 Airborne Dust Potential

Production operations that involved handling dry uranium materials were generally equipped with 

engineered ventilation systems for controlling dusts. Standard operating procedures required the use 

of respiratory equipment when dusty conditions were anticipated. Good housekeeping involving the 

immediate clean-up of spilled uranium products was also a standing policy and practice. In spite of 

this emphasis on engineered and administrative contamination controls and policy to reduce the 
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Table 5-2. Uranium enrichments and associated isotopes.

a 

DU

Isotopes Mass %

Relative activity 

(pCi/μg) Activity %

U-234 0.001 0.062 15.46

U-235 0.20 0.004 1.07

U-236 0.00031 0.0002 0.05

U-238 99.80 0.3354 83.42

Total Not applicable 0.402 Not applicable

NU

Isotopes Mass %

Relative activity 

(pCi/μg) Activity %

U-234 0.0054 0.334 48.86

U-235 0.7204 0.016 2.28

U-236 0 0 0

U-238 99.274 0.334 48.86

Total Not applicable 0.683 Not applicable

1% EU

Isotopes Mass %

Relative activity 

(pCi/μg) Activity %

U-234 0.01 0.62 63.7

U-235 1.0 0.02 2.1

U-236 0 0 0

U-238 98.99 0.333 34.2

Total Not applicable 0.973 Not applicable

2% EU

Isotopes Mass %

Relative activity 

(pCi/μg) Activity %

U-234 0.02 1.243 76.9

U-235 2.0 0.043 2.68

U-236 0 0 0

U-238 97.98 0.329 20.4

Total Not applicable 1.616 Not applicable

a. U-236 is less than 1% mass activity in DU, NU, 1% EU, and 2% EU.

Table 5-3. Uranium chemical forms. 

Uranium compound

Chemical 

formula

Lung 

absorption 

type

Uranium hexafluoride (gas) UF6 F

Uranyl fluoride UO2F2 F

Uranyl nitrate UO2(NO3)2 F

Uranium trioxide (orange oxide) UO3 M

Uranium tetrafluoride (green salt) UF4 M

Uranium dioxide (brown oxide) UO2 S

Uranium tetraoxide UO4 M

Uranium oxide (yellowcake) U3O8 S

High fired uranium oxides UO2 S

release of radioactive materials, spills and routine releases occurred. In addition to the routine 

releases at FMPC, there were frequent “upset” conditions (i.e., spills, effluent filter ruptures, etc.) that 

produced episodic airborne radioactivity in the work areas and plant effluents, and were of a 

magnitude such that the ventilation systems were unable to contain all of the releases.
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In 2000, a Fernald team working on the DOE Ohio Field Office Recycled Uranium Project Report

(DOE 2000a) qualitatively rated various plant processes in relation to the potential for producing 

airborne dusts in high, medium, or low categories, based on the criteria below. Though qualitative, 

this evaluation was consistent with historical site air activity measurements and recorded internal 

exposures. The evaluation involved the following ratings and the qualitative descriptions of “very 

large quantities,” “small quantity, intermittent,” and “highly reliable” are used as quoted in the 

reference:

• High ratings were assigned for operations that processed very large quantities of materials 

using processes or equipment susceptible to occasional events. Examples of such events 

include digestion tank fuming and denitration pot eruptions in Plant 2/3, failure of 

hydrofluorination banks in Plant 4, temperature excursions in Plant 8 furnace operations, and 

failure of ventilation system dust bags during both operation and maintenance. The ventilation 

systems were inadequate to contain the dust emissions from these events, and all plant 

workers were susceptible to exposure from these releases.

• Low ratings were assigned for small quantity, intermittent operations, those that did not involve 

appreciable amounts of dry materials, and those considered highly reliable.

• Medium ratings were assigned for operations which were neither of clearly high nor clearly low 

radioactive material release potential.

Table 5-4 lists the summary results of this evaluation including related information on the typical 

materials.

Table 5-4. Plant processes, materials, and uranium dust release potential. 

Plant Process Compounda 

Airborne dust potential 

(qualitative)

5 Metal reduction and casting UF4, U, U3O8, residues High

8 Feed preparation furnaces U3O8, U, residues High

4 Banks 7–9, packing stations UO3, UO2, UF4 High

2/3 Digestion, extraction denitration Prepared feed, U3O8, UO2, UO3 Medium

9 Reduction, casting UF4, U, U3O8, residues Medium

Pilot Hex reduction, metal operations UF4, U, U3O8, residues Medium

1 Milling U, U3O8, residues Low

a. MgF2 (one of the residue products) in Plants 1, 2/3, 5, 9, and the Pilot Plant, contributed to the dust potential rating, but

did not carry more than trace quantities of radionuclides. The hazard potential for MgF2 typically was associated with 

nonuranium radionuclides (i.e., uranium progeny and RU contaminants).

Accordingly, jobs at Fernald which contributed to higher airborne uranium exposures in routine 

operations included (Heatherton 1975, p. 159):

• Changing hoppers;

• Lidding and delidding drums;

• Handling contaminated drums;

• Adjusting weights at filling stations;

• Dumping drums of concentrate;

• Operating pot-filling machine in the Metals Plant;

• Breakout of furnace pots and molds;

• Cleaning uranium surfaces, graphite crucibles and molds;

• Reassembly of crucible and mold parts;

• Operating crushing and grinding equipment;

• Changing receiving drums at dust collectors;
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• Cleaning out dust collector housings and furnace enclosures; and

• Breaking up clogged material in containers, conveyors, downcomers, and other equipment.

5.3.2 Recycled Uranium

This section extracts some of the information on RU in ORAUT-RPRT-0052, Feed Materials 

Production Center Internal Dose Topics (ORAUT 2011a), which can be reviewed for additional details.

RU is uranium that has been recovered from irradiated production reactor fuel and plutonium 

production fuels. The RU was separated in the chemical processing plants at Hanford and the 

Savannah River Site (SRS), but was known to contain traces of TRU elements and fission product 

impurities. Most of the RU delivered to Fernald came from the Hanford Site, the Paducah Gaseous 

Diffusion Plant (PGDP), and SRS. The primary contaminants were 238Pu, 239

Pu, 241Am, 237Np, and a 

fission product, 99Tc. The 239Pu, 237Np, and 99Tc were the radionuclides of greatest concentrations 

and were tracked and documented for control purposes.

The first uranium feedstock introduced to the Fernald plant in 1953 was high-grade ores (50% to 80% 

uranium) (i.e., pitchblende ores from the Belgian Congo and other locations, which could be 

processed in a modified PUREX process directly to high-purity uranium). Other uranium feedstock 

came from low-grade ores, which was processed near the mines by mills that produced NU in the 

form of yellowcake (U3O8) in the range of 60% uranium. The yellowcake required further processing 

by PUREX to a high-purity state. When RU from the chemical processing plants was introduced to 

the process feed streams at Fernald it was a high-purity oxide and was normally blended with NU or 

other existing feedstocks.

Relatively small quantities of RU (a recorded mass of 45 MT RU) from Hanford were received at 

FMPC as early as 1955 (DOE 2000a, 2000b) but did not constitute a feed stream component of 

significance. The first significant quantity of RU was received for processing on February 13, 1961, in 

a shipment of UF6 that was to be converted to UF4 (DOE 2000b) and introduced to the feed streams 

of the plant.

5.3.2.1 Mass Balance, Material Flow, and Plutonium-239 Levels in Recycled Uranium in the 

Weapons Complex

From the beginning of operations, urinalyses for uranium were conducted extensively for purposes of 

controlling biological heavy-metal concerns. The presence of RU contaminants was recognized but 

was not specifically analyzed due to the anticipated low levels and related low-hazard contribution. 

The concentration of contaminants in RU changed with time, so default values for dose reconstruction 

are defined for discrete time intervals. Because this represents a bounding approach, it is not 

necessary to precisely quantify the mass balance and material flow of RU in the weapons complex to 

establish a bounding estimate of the contaminant concentration in RU at Fernald.

While DOE documents about mass flows of RU in the weapons complex include some discrepancies, 

these discrepancies in mass flow do not preclude use of the data to define bounding default 

assumptions. The defaults were chosen such that they bound the highest values of all the processes, 

with the exception of the concentrations in the gaseous diffusion plant (GDP) tails and wastes.

The concentration of the contaminants varies slightly with the source of the data. For example, the 

mass concentration of plutonium in DOE (2000b) is 0.9 ppb; for the complex-wide rollup, the mass 

concentration is 3.7 ppb; in DOE (2003), the mass concentration from the primary sites is 4.2 ppb. 

These values are essentially the same and all well below the 10-ppb control guideline at Hanford.
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5.3.2.2 Major Contaminants in Recycled Uranium

Table 5-5 details DOE’s estimate of the quantity and characteristics of RU and its major contaminants 

at Fernald. Approximately 218 g of 239Pu, 25.7 kg of 237Np, and 332 kg of 99Tc were received in 

246,683 MT of RU during the history of the plant (DOE 2000b). The historical average of 

contaminants in RU was approximately 1 ppb 239Pu, 104 ppb 237Np, and 1,346 ppb 99Tc.

Table 5-5. Average characteristics of RU and its major contaminants. 

Uranium 

enrichment

MTU 

receipts

Percent 

uranium 

receipts

Total Pu239 (g)Pu-239 

(ppb)

Total 

Np-237 

(g)

Np-237 

(ppb)

Total 

Tc-99

(g)

Tc-99

(ppb)

EU 60,181 24.4 207.9 3.5 19,048 316 328,740 5,382

NUa 89,649 36.3 4.1 <0.1 3,026 34 1,197 13

DU 96,853 39.3 5.7 <0.1 3,669 38 2,061 21

Total or 

historical 

average

246,683

total

100.0

total

217.7

total

0.9

historical

average

25,743

total

104

historical 

average

331,998

total

1,346

historical 

average

a. Referenced as “normal uranium,” but is natural uranium for the purposes of discussion.

A limit of 10 ppb total plutonium in RU was set by the nuclear weapons complex (primarily Oak Ridge) 

for use at Hanford (ORAUT 2015a; DOE 2000a). Initially no numeric limits were specified for 237Np or 

99Tc. The plutonium limit of 10 ppb was set such that the radiological impurities did not exceed 0.1% 

of the alpha activity from NU.

5.3.2.3 Tower Ash and Plutonium Out of Specification

The gaseous diffusion process concentrated the RU contaminants, primarily due to lack of volatility of 

the fluoride compounds in comparison with those of uranium, in tower ash and other residues in the 

UF6 conversion and handling processes. These ash and residues also contained significant amounts 

of uranium, which required that they be recovered according to economic discard criteria. In 1972, 

the decision was made by the AEC to process these materials from PGDP and the other GDP sites. 

Fernald was directed to process most of these materials. This decision introduced RU contaminants 

to Fernald with plutonium out of specification (POOS), meaning that it exceeded the 10 ppbU limit.

POOS feedstock from PGDP was sent to FMPC for reprocessing with the prior knowledge that the 

material was known to be above the established contamination limit. The 1970s and increased 

amounts in the 1980s GDP shipments accounted for a major portion of the total TRU inventory at 

FMPC, essentially doubling the plant inventory over this period (Bassett et al. 1989; DOE 2000b). 

The intent was to blend these materials with the inventories at the plant and still maintain the 

plantwide limits that had been determined to provide an adequate level of safety (i.e., staying below 

the average sitewide plutonium specification of 10 ppb). Workers handling POOS materials directly at 

Plant 1 and other locations were protected with airline respiratory equipment, particularly for the 1976 

shipment of tower ashes from PGDP. Consequently, exposures to RU contaminants greater than the 

10 ppbU could have increased significantly starting in 1973.

An examination of the statistical data sheets in Appendices C and F of DOE (2000b) indicates that 

maximum values in some subgroups (as determined by the 95th percentile of a lognormal distribution) 

ranged up to approximately 400 ppb Pu. These values are summarized in Table 5-6. A notable

exception is the tower ash from PGDP, which had a mean value of 1,732 ppb Pu (range of 0.6 to 

3,505). This maximum-level POOS material was handled as a recognized short-term risk during feed 

preparations and blending operations. Additional protective precautions were applied during these 

short-term tasks, and radiological procedures were implemented (WMCO 1986a; Jelinek and Thomas 

1985). These precautions included airline respiratory protection and protective clothing over normal 
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plant clothing that consisted of head covering, gloves, and shoe covers. Tape was applied at ankles 

and sleeves. Weekly urine bioassay was required of those involved (Author unknown undated a).

Table 5-6. RU contaminant levels by process subgroups. 

Subgroup Description

Pu ppb U at 

95%

Np ppb U at 

95%

Tc ppb U at 

95%

1A Miscellaneous 133 3,692 3,060

1B Miscellaneous – minor offsite 1.6 179 1

2 UF6 source UF4 (GDP tails) 1.2 84. 171

3 UF6 source metal and scrap 0.01 4.2 15

4 Normal U product, residues, and scrap 0.12 139 33

5 Enriched UF6 source products and residues 3 109 15,410

6A UO3 PUREX source (A508)(unblended) 4.6 655 14,119

6B LEU products A508 UO3/UF4 (low cross) 4.1 643 19,732

6C LEU products A508 UO3/UF4 (high cross) 34 1,198 4,823

6D A500 coded enriched residues 7.6 246 2,701

6E SRS UNH 37.8 Not measured Not measured

6F SRS UO3 – not shipped to FMPC 4.6 Not measured Not measured

7A A508 based derbies 17 513 2,870

7B A508 based ingots and metal 1.7 433 133

8a Enriched MgF2 342 4,417 4,619

9 GDP incinerator ash and scrap residues 283 5,116 474

10A Tower ash and decontamination residues 1,732 17,277 4,445

10B UO3 from tower ash 34.9 842 4,154

11a Waste residues 152 10,666 6,991

a. Although the results in the table are reported in ppb U, this measure might appear less meaningful in subgroups in 

which there is a reduced amount of uranium (in one-digit percentage levels), such as subgroups 8 and 11, in which the 

MgF2 and chemical plant raffinates accumulated some isotopes but were relatively low in uranium by design, yet still 

continued to be a process stream to recover uranium above the economic discharge levels. Note that even with 

decreased uranium levels the contamination default can still be a ratio of uranium.

The Fernald Health and Safety staff was aware that TRU and fission product contaminants were 

present in the RU, although they believed it to be less than 0.1% of the total alpha activity. Therefore, 

the biological concerns were thought to represent an increase of less than 10% to 15% in the hazard 

level. Radiological control was administered continually on the basis of both air analysis (gross alpha 

counts) and urine bioassay for uranium. Before 1989, no TRU analyses for radiological safety were 

performed on a routine basis for either airborne or urine activity, and exposure controls remained 

based on chemical toxicity under the assumption that these controls would be sufficient for all the 

radiological issues (Bassett et al. 1989). Although the alpha activity from the TRU alpha emitters 

would have been collected and detected on the air samples, the reported results were all considered 

to be uranium and compared to the MAC (WMCO 1986a).

Over time, much of the POOS material was blended with the NU feedstock and with RU of typical 

concentration. This was done to maintain the plantwide limits that had been determined to provide an 

adequate level of safety (i.e., staying near the average complexwide plutonium specification of 

10 ppb). Workers handling the POOS materials directly before blending with other materials (Plant 1 

and other locations) were provided with special protection. As a background comment, a small 

amount of this POOS material from the GDP was inadvertently stored for a period of years. The 

discovery of this material in 1985 was categorized as an unusual event and the material was 

subsequently processed and disposed.

After processing the PGDP POOS, the blended uranium oxide resulted in temporary storage of 

uranium with a maximum concentration of plutonium of 43 ppb U. In addition, several of the uranium 

purification processes resulted in concentrating the TRU contaminants in waste streams. Formal 
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investigations were conducted that traced the TRU contaminants through the various plant processes. 

One study in 1975 defined contaminant affinity for the metal production furnace mold materials (some 

TRU studies are referenced in DOE 2000b). These analyses demonstrated that the extraction and 

metal conversion processes concentrated plutonium, neptunium, and uranium progeny contaminants 

in the raffinate waste of the liquid extraction columns and in the MgF2 from the thermite metal 

conversion furnaces. MgF2 slag from the thermite metal reduction process was also processed 

through a slag leach process and then through the chemical extraction plant. These concentration 

points or areas were identified by analytical methods in early years.

The plutonium contaminant mass concentrations are therefore assumed to be 400 ppb U after 1972. 

The basis for assuming 400 ppb U rather than the 1,732 ppb U for subgroup 10A is briefly described 

here. The high values in the activity distribution came from PGDP in a single shipment in 1980. In 

June of 1980 a single shipment of approximately 24 MTU PGDP process plant ash was shipped in 16 

T-hoppers that contained 25.3 g of plutonium. This shipment was sampled and stored primarily to 

develop blending material inventories and processes to accommodate this unusual shipment. In 1982 

after about 2 years, there is a record of repackaging five of the T-hoppers with the highest 

contaminant levels into large containers in Plant 4 to facilitate process needs in Plants 8 and 2/3. 

Based on the recorded times of the five hoppers with the greatest problems, 150 hours was 

conservatively estimated for repackaging those eight hoppers with concentrations above 400 ppb U 

plutonium. The next step (after another delay) was blending with sump cake in Plant 8 of appropriate 

content to create a calcium uranate feed for further dissolution and process in Plant 2/3, these being 

semi-remote operations. The final recorded processing resulted in a formal approval to process the 

Plant 2/3 UO3 product, which was in the range of 20 to 42 ppb U plutonium). Therefore, the period of 

time material in excess of 400 ppb U was handled was relatively brief. Airline respirators and 

associated protective clothing were used, which ensures that 400 ppb is sufficiently conservative for 

dose reconstruction. Attachment A, Rationale for 400 ppb U Pu for 10A Process Stream, provides 

greater details.

From April 1986 to 1989, 675 bioassay samples from 441 workers were collected for plutonium 

analysis. “Those samples were collected primarily from workers who were expected to be and, in 

some cases, were actually involved in POOS processing and cleanup campaigns in Plants 4 and 8” 

(Bassett et al. 1989). Only 10 individuals exhibited quantities of plutonium in the urine above 

0.02 dpm/sample. [This contractual urine minimum detectable activity (MDA) was determined to be 

unrealistically low and later adjusted to 0.13 dpm/sample for plutonium urinalyses.] Further 

investigation (follow-up samples) revealed that 3 of the 10 were below the specified detection limit of 

0.02 dpm/sample. All of the 10 individuals were flown to Hanford, where in vivo counts were 

conducted for plutonium. All results were below the detection limit of the Hanford in vivo analyses. 

The conservative assumption was made by the site contractor that 7 of the 10 were technically at the 

urine MDA. Directions were given to “regard those results as valid indicators of exposure and to 

calculate the plutonium dose commitments to those workers based on these worst-case assumptions. 

The dose commitments are to be reported in each employee’s permanent exposure record” (Bassett 

et al. 1989). Fernald radiation protection practices were adjusted to more rigorously account for those 

trace isotopes within the DOE radiation protection standards in effect at the time. The information 

above is taken from the report of a DOE team of investigators and documents an organized effort by 

an incoming contractor to verify the previous assumptions that TRU exposures were not a major 

exposure source.

In 1986 and later when 238Pu and 239/240Pu urine sample results are available for full-time POOS 

employees, the RU contaminants are based on the uranium intake. The purpose of these samples 

was for the site to address the issue of exposures from RU contaminants and that there was no 

indication of work with pure plutonium at Fernald. Therefore, because plutonium is addressed as part 

of the RU contaminants, missed dose associated with plutonium bioassay results are not assigned. 

The only exception is for a positive 239/240Pu sample that is greater than the derived intake for uranium. 
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As noted above, only 10 individuals exhibited quantities of plutonium in the urine above 

0.02 dpm/sample (the initial MDA).

Dose reconstructors should contact the Site Lead and PID for guidance if a claim file contains positive 

plutonium bioassay results.

5.3.2.4 Other Contaminants in Recycled Uranium

Routine process chemical analysis at Hanford included gross beta and gamma measurements. 

These results were compared to the baseline measurements from aged NU (DOE 2000a). Limits on 

these other contaminants were set to 0.685 pCi/μg NU (beta/gamma), 50 μCi/lb U for 103Ru or 106Ru, 

15 μCi/lb U for 95Zr and 95Nb, and 2 μCi/lb U for all other isotopes.

The mass balance reports established that most of the RU came from Hanford and was recycled from 

weapons-grade plutonium. Therefore, a 6% weapons-grade mixture that had been chemically purified 

in 1961 was selected for bounding calculations. For dose reconstruction, the plutonium alpha dose 

from the plutonium mixtures in RU is assumed to be from 100% 239Pu. Plutonium mixture ratios for 

fresh, 10-, 20-, and 30-year aged intervals after purification were used to estimate the 241Pu and 241Am 

contaminant levels.

Specific analytical information for trace concentrations of other fission products as documented in a 

Hanford reference (DOE 2000a) are listed in Table 5-7.

Table 5-7. Other radioactive contaminants in RU from Hanford. 

Element/isotope Observed range

Plutonium (primarily Pu-239) <1 to 2 ppb U

Neptunium (primarily Np-237) 0.04 to 0.16 ppm U

Thorium 8 to10 ppm U

Technetium 3 to 4 ppm U

Ru-103/106 <6 μCi/lb U

Zr/Nb-95 <4 μCi/lb U

Americium-241 40 to 320 Bq/g U

Other gamma emitters excluding Tc-99 0.09 to 0.75 μCi/lb U

5.3.2.5 Summary Default Values for Recycled Uranium

RU was first processed in significant amounts at FMPC beginning in 1961, so none of the associated 

contaminants were present before 1961. Therefore, uranium intakes in 1961 and later should include 

default factors to account for exposure to the RU contaminants. Table 5-8 lists the uranium 

enrichment assumptions during the RU period.

Table 5-8. RU enrichment. 

Period Enrichment

1961–1964 1%

1965–2006 2%

Table 5-9 lists the plutonium mixture ages assumed during the RU period.

Contaminant concentrations between 1961 and 1972 for plutonium alpha, 237Np, and 99Tc are based 

on the pre-POOS concentrations explained in Section 5.3.2.2. Contaminant concentrations from 1973 

on for plutonium alpha, 237Np and 99Tc are based on bounding POOS concentrations explained 

Section 5.3.2.3. Contaminant concentrations for 232Th, 103/106Ru, 95Zr/95Nb, and 90Sr are based on the 

Hanford limits and observed concentrations explained Section 5.3.2.4. These are summarized in 

Table 5-10.
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Table 5-9. Age of plutonium mixture in 

RU. 

Period Age

1961–1965 Fresh

1966–1975 10 Year

1976–1985 20 Year

1986–2006 30 Year

Table 5-10. RU contaminant mass concentrations. 

RU contaminant 1961–1972 1973–present

Pu-alphaa 10 ppb U 400 ppb U

Np-237 400 ppb U 11,000 ppb U

Tc-99 6,000 ppb U 20,000 ppb U

Th-232b 10,000 ppb U 10,000 ppb U

Ru-103/106c 50 μCi/lb 50 μCi/lb

Zr-95/Nb-95d 15 μCi/lb 15 μCi/lb

Sr-90 2 μCi/lb 2 μCi/lb

a. The plutonium alpha mixture is assessed as 100% Pu-239. Am-241 and 

Pu-241 are assessed based on 6% weapons-grade plutonium mixture.

b. An intake of Th-228 should be added and assumed to be at 70% equilibrium 

with Th-232.

c. Ru-103/106 is assumed to be 100% Ru-106 due to its longer half-life.

d. Zr-95/Nb-95 pair is assumed to be in equilibrium with equal activities totaling 

the mass concentrations indicated in the table. 

5.3.3 Thorium

5.3.3.1 Thorium Operations, 1954 to 1979

According to a timeline of thorium processing operations (ORAUT 2008a), thorium processing began 

at Fernald in 1954 and continued through the first quarter of 1979. 

SEC classes have been designated based on inability to reconstruct thorium dose with sufficient 

accuracy for periods beginning in 1954 and ending in 1978 (Section 5.1.3). Therefore, no detailed 

information or process description for activities before 1979 is provided.

5.3.3.2 Thorium Operations, 1979 to 2006

Thorium processing was completed in 1979, after which thorium exposure was due to repackaging 

and shipping operations.

In 1988, an environmental assessment indicated that the inventory of thorium materials at Fernald 

consisted of approximately 1,100 MT (as thorium) of various thorium compounds. In 1988, the Plant 8 

silo and bins contained approximately 175 MT of bulk thorium oxide materials plus other inert 

materials. The remainder of the thorium inventory was stored in various sizes of container in 

warehouse Buildings 64, 65, 67, and 68 and as thorium nitrate in Tank 2 at the Pilot Plant. A small 

number of drums were in outside storage adjacent to Building 65 (DOE 1987a, p. 10). Table 5-11 

summarizes Fernald’s stored inventory of thorium in 1987 (Reafsnyder 1987, p. 3).

This includes a total of 13,329 drums and 2,062.2 MT of thorium material (Grumski 1987, p. 13), plus 

the Plant 8 silo and bins material and the thorium nitrate solution in the Pilot Plant tank. In 1985, NLO 

identified the need to repackage much of the thorium stored in 55-gal drums due to “serious 
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Table 5-11. Inventory of stored thorium, 1987. 

Location Description or Quantity

Weight

(MTTh)

Plant 8 silo and bins Bulk thorium oxide material and 

other inert materials

175

Pilot Plant Tank 2 Thorium nitrate solution 9

Building 64 (Plant 9 warehouse) 181drums 64.2

Building 65 (Plant 5 warehouse) 5,599 drums 1,246.3

Building 67 (Plant 1 storage building) 5,992 drums 213.5

Building 68 (Pilot Plant warehouse) 1,317 drums 485.7

Pad west of Building 65 212 (240 overpacked) drums 52.5

Totals Not applicable 2246.2

deterioration” (Adams 1985). Some repackaging was done in 1985 involving only a few individuals, 

during which breathing zone (BZ) samples were taken (Mengel 1985). In 1986, possible overstressed 

conditions in the Plant 8 silo and bins used to store thorium oxide were identified (WMCO 1986b).

In 1988, Fernald initiated a “three Project plan” for improving the temporary storage conditions for 

thorium (WMCO 1992a, p. 203; Reafsnyder 1987). During the execution of the plan, approximately 

1,094 MTTh were declared waste (WMCO 1992a). The “three Project plan” was followed by two 

additional phases which resulted in the removal of all thorium from the site.

• Project 1. June 1988 (anticipated schedule, see DOE 1987b, p. 19) to March 1989, addressed 

the bulk thorium materials in the Plant 8 silo and bins. As the bulk thorium was removed from 

the silo and bins, it was placed in double-containment drums called overpacks (a 48-gal drum 

was packaged inside a 55-gal drum), inventoried, and monitored. The drums were then stored 

in an onsite warehouse along the northern edge of the production area, away from daily plant 

operations. The silo and bins were then decontaminated and demolished (WMCO 1992a). 

This work was subcontracted to International Technology (IT) Corporation, which participated 

in the site’s radiation safety program, and the employees assigned to the project were required 

to receive in vivo counts at 6-month intervals (see whole-body counts in February 1989 in 

Walker 1989).

• Project 2. January to March 1990, addressed overpacking of the 241 containers stored 

outdoors; 212 of the containers were drums. A remote system to handle, identify, and 

overpack the thorium drums and containers was designed. Each container was inventoried, 

weighed, overpacked, and then placed in temporary indoor storage (WMCO 1990a). For 

Project 2, isotopic fecal sampling was recognized as the only method capable of monitoring 

intakes below the annual limit on intake (WMCO 1991, p. 4). The detection limit of this 

analysis was 1.0 dpm (WMCO 1991, p. 6 and others). As a result, starting with this project, 

baseline fecal samples were taken from the directly exposed workers (chemical operators, 

forklift operators, etc.). These results are included in the site’s HIS-20 database. General 

area (GA) samples were taken and each operator was required to wear a BZ sampler when 

inside the radiological barrier (Walker 1990, p. 3). These samples are not in the HIS-20 

database and are therefore unlikely to be reported to NIOSH by DOE.

• Project 3. 1991 to 1995, involved overpacking 13,000 containers of thorium in warehouses. 

This project included overpacking in the Thorium Warehouse, the Plant 1 Thorium Warehouse, 

the Pilot Plant Warehouse, and the (old) Plant 5 Warehouse (WMCO 1992a). Radiological 

controls similar to those in Project 2 were in place during these operations. The first thorium 

was shipped to the Nevada Test Site (NTS) in 1992 (FERMCO 1993a), and shipping 

continued each year thereafter. Almost 6,000 gal of thorium nitrate, a contaminated acid 

waste stream, were treated and solidified in 1995 (FERMCO 1996).
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• Thorium Overpack Project (TOP). Project 3 was succeeded by the TOP in 1996, which 

completed the overpacking of the drums stored in warehouses in 1997. This did not include 

some 8,500 drums of thorium waste that required Resource Conservation and Recovery Act of 

1976 (RCRA) classification (FDF 1998, p. 54).

• RCRA classification and shipping for treatment or disposal, 1997 to 2006. Characterization of 

the remaining 8,500 containers of thorium legacy waste was initiated in 1997. Through the 

end of 2003, over 8,400 of these containers were shipped off the site for treatment, with 

subsequent disposal at NTS. The remaining thorium inventory of approximately 

100 containers had been evaluated. Of these, approximately 90 containers were non-RCRA, 

low-level radioactive waste and 10 were RCRA hazardous waste. From 2004 to 2006, the 

remainder of the mixed thorium waste was shipped to Envirocare of Utah for treatment and 

disposal; the nonhazardous thorium waste was shipped to NTS for disposal. At the end of 

2004, there were only a few remaining containers of thorium waste. The last shipment of 

hazardous waste occurred October 2, 2006, ending hazardous waste management activities

(FDF 1998, 1999; Fluor Fernald 2000, 2001, 2002, 2003a, 2004, 2005, 2006; DOE 2007).

The major Fernald thorium work and regulatory requirements for thorium after the SEC periods (that 

is, for 1979 to 2006) are summarized below in Table 5-12.

Table 5-12. Post-SEC thorium activities and regulatory requirements, 1988 to 2006. 

Project Project Description Timeframe Regulatory requirements

1 Overpacking Plant 8 and silo 

thorium material

November 1988–

March 1989

DOE 5480.11 (DOE 1988)

DOE 5480.1A

2 Overpacking 241 containers 

(212 were drums)

January 1990–

May 1990

DOE 5480.11 (DOE 1988)

3 Overpacking 13,000 containers 1991–2006 DOE 5480.11 (DOE 1988);

Radiological Control 

Manual (Fluor Fernald 

2003b);

10 CFR Part 835

5.3.3.3 Thoron

Thoron was present at Fernald and a portion was released during processing, repackaging, and other 

tasks, as well as during storage, of the metric tons of a variety of forms of thorium.

Because the amount of time workers spent in a particular location varied in terms of the operating and 

handling, and because passive storage and remediation categories varied, the determination of 

working level months (WLMs) per year incorporated occupancy factors for the fraction of time spent 

by workers in these areas. Attachment B, Analysis of Potential Thoron Exposures, provides additional 

information on thoron exposures. It does not provide precise estimates because of the lack of sitespecific thoron data by time and location, but it provides a reasonably conservative estimate of thoron 

exposures. The attachment indicates that the nature of storage facilities results in a minimal 

occupancy time for required functions. The time of occupancy during production periods is assumed 

to be 1,750 hr/yr, based on an average of a continuous 1-MTTh production rate assumption. At a 

1-MTTh daily rate, the recorded annual quantities processed would be completed in just a fraction of a 

year. For long-term storage of high-integrity storage containers, 1 mo/yr is assumed. For routine 

storage conditions, including required repackaging, 500 hr/yr (or 3 mo/yr) is assumed.

Doses for thoron potential were separated into four categories: (1) various operating and handling 

period (2) long-term storage period, (3) passive storage period, and (4) remediation period. The 

highest WLM per yr values for 1954 to 1985 from Attachment B, Tables B-5 and B-6, were used to 

provide a bounding thoron dose. Passive storage occurred from 1986 to 1987, when the potential for 
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no significant thoron dose occurred. Remediation occurred from 1988 to 2006, along with increased 

radiological controls. The potential for thoron exposure was greatly reduced in comparison with the 

earlier years at the site, as discussed below.

5.3.3.3.1 Operations and Handling, 1954 to 1971

Thorium processing began in 1954 (ORAUT 2008a), and there were at least potential exposures from 

levels of thoron during processing and storage of thorium. The thoron exposure potential varied 

between plant operations and other handling areas, and for some years and areas was insignificant. 

There were various storage sites that had the highest potential for thoron exposure. Although Fernald 

did not store large quantities of thorium until 1972 when it became the DOE repository, because work 

locations are unknown, the maximum WLM per year will be assumed.

5.3.3.3.2 Long-Term DOE Repository Storage, 1972 to 1985

Fernald became the national thorium repository for DOE starting in 1972. An inventory of thorium

production orders shows that there were no orders after 1985 and only a few from 1979 to 1985 

(Bonfer 1988). It is likely that most or all of these production orders involved only taking stored 

material out of a warehouse and shipping it to a customer. The occasional thorium repackaging effort 

(which typically involved placing eroding containers into larger containers) would have been of short 

duration and would have involved very few individuals.

5.3.3.3.3 Passive Storage, 1986 to 1987

A period of passive storage occurred at the site from 1986 through 1987. This was a time for planning 

for the disposition and off-site shipping of the stored thorium inventory (Reafsnyder 1987). In 1985, 

Lockwood Greene was under contract to DOE to perform a conceptual design for environmental, 

health, and safety improvements for a number of FMPC facilities. Included in this contract was the 

development of a conceptual design for a thorium handling system to remove the materials from the 

Plant 8 storage vessels. As part of the design process, Lockwood Greene performed a preliminary 

structural analysis of the storage vessels. On June 9, 1986, a Thorium Silo Remediation/Thorium 

Materials Disposition Task Force was convened (WMCO 1986b). Structural remediation of the 

Plant 8 silo occurred from June 1986 to September 1986 (Cook 1987, p. 7). In September 1987, 

Fernald began the design and construction of the handling system necessary to remove and package 

the bulk thorium materials. After construction, the removing, handling, and packaging of thorium 

began in November 1988 (Aas et al. 1987, p. 451).

5.3.3.3.4 Remediation, 1988 to 2006

The majority of efforts at the site involved remediation and cleanup activities from 1988 through 

October 29, 2006. In 1988, Fernald initiated a “three Project plan” for improving the temporary 

storage conditions for thorium (WMCO 1992a, p. 203; Reafsnyder 1987). The removal, handling, and 

packaging of thorium began in November 1988 (Aas et al. 1987, p. 451). Remediation was completed

at Fernald on October 29, 2006 (Stoller 2007, p. 20). By the late 1980s, the respiratory protection 

program was well established. Those who worked on thorium remediation wore respirators. Even 

allowing a respiratory protection factor of only 10 is equivalent to reducing the thorium workers’ 

approximate full-time occupancy to about 1 mo/yr occupancy. Therefore, the value for 1988 to 2006 

is appropriate for all radiological workers, especially since the WLMs derived in Attachment B took no 

credit for respiratory protection. Within the remediation period, a two-tiered approach should be 

taken: (1) radiological workers and (2) nonradiological workers. Nonradiological workers are those 

who clearly did not work in the process area (cafeteria, accounting, legal staff, etc.) and should not be 

assigned thoron dose. Radiological workers are those who could have worked in the process area, 

and are assumed to have the potential for thoron exposure.
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5.3.4 Q-11, K-65, Domestic Uranium, and Raffinates

A detailed processing history of pitchblende and domestic uranium feedstock, as well as the available 

monitoring data, can be found in ORAUT-RPRT-0052 (ORAUT 2011a). A brief summary is contained 

here.

During the Manhattan Project, extremely high-grade uranium ore (50% to 80%) was imported from 

African Metal Company (AMC) in the Belgian Congo and was designated with the code “Q-11” 

(Robinson 1973). At this level of uranium content, the ore could be processed directly in the modified 

PUREX process at the Fernald site. From 1946 until 1953, MCW in St. Louis, Missouri, produced 

uranium from the Q-11 ore. The Q-11 ore contained all of the uranium progeny, including 226Ra. 

When the ore was refined, 226Ra and other progeny were concentrated in the waste stream, which 

was identified with the code “K-65.” The term raffinate was used to describe the waste stream. The 

K-65 residues were returned to Belgium until March 1949 (NLO 1979; Heatherton 1949). In 1949, an 

agreement was reached wherein AEC would store the K-65 residues for later return to AMC. 

Beginning on March 22, 1949, the K-65 residues from MCW were sent to Lake Ontario Ordnance 

Works (LOOW) to be stored (Heatherton 1949). Construction of FMPC began in 1951 and MCW 

began shipping K-65 material to FMPC rather than LOOW (Heatherton 1952; Walden 1952; 

Vogel 1989).

When construction was completed, storage of all K-65 material was assigned to FMPC and two large 

ground-level silos were built for this purpose. Most U.S. and Canadian ores were low grade and 

required processing near the mine sites. The product from these mills was yellowcake (U3O8) of 

approximately 60% uranium content. After 1958, when processing of the Q-11 and other pitchblende 

ores at FMPC had been completed, the source of the uranium for processing at FMPC was this 

yellowcake. The yellowcake had most of the radium and other impurities removed before being 

shipped to FMPC. This raffinate was called cold raffinate because the waste stream had much lower 

concentrations of 226Ra and progeny in comparison with the K-65 raffinate. The cold raffinate was a 

calcined dry, dispersible powder that was transferred to Silo 3 using an enclosed airlift 

(ORAUT 2008b).

The silo waste became the property of the United States in 1983 and was a storage problem. The 

concrete silos cracked, leaked, and were the object of several upgrade efforts to reduce both the 

liquid leaks and the release of radon and its progeny. A dirt berm was added around the silos 

(primarily as silo wall support, but it also provides shielding), as were roof sealants and a 

semipermeable cap on the waste itself, all of which reduced the routine releases of radon and 

progeny from the silos.

The operation of handling the large number of drums of K-65 waste materials and dumping them into 

the silos in the period from July 1953 to September 1958 involved external and internal exposure to 

the radium and other associated contaminants. In addition, there were several operations to improve 

the confinement of the silos for the radon gases and associated progeny, although the waste 

materials have not been directly disturbed since they were put into storage. Air sample data sheets 

from around1953 provide insight into the operation and exposure levels. Bioassay information about 

internal exposures to radium or the associated contaminants consists of over 600 radon breath 

samples. In vivo counts were periodically performed by the Y-12 mobile counting laboratory, but 

those did not begin until 1968, well after completion of the handling and dumping of the drums of 

waste.

The isotopic contents of the three silos are shown in Table 5-13.

Silos 1 and 2 contained the K-65 (hot) raffinates and included relatively high concentrations of 226Ra. 

Silo 3 contained cold raffinate with very little 226Ra. Because uranium was the extraction product, the 
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Table 5-13. Isotopic composition of K-65 silos 1, 2, and 3 (Soldano 1997, Tomes 2001). 

Isotope

Silo 1 

activity 

concentration

(nCi/g raffinate)

Silo 1 

activity 

fraction

Silo 2 

activity 

concentration

(nCi/g raffinate)

Silo2 

activity 

fraction

Silo 3 

activity 

concentration 

(nCi/g raffinate)

Silo 3 

activity 

fraction)

Uranium 1.68 1.61E-03 2.37 3.03E-03 3.63 4.83E-02

Ac-227 7.67 7.36E-03 6.64 8.50E-03 0.93 1.24E-02

Pa-231 Not applicable 0.0 4.04 5.17E-03 0.63 8.39E-03

Pb-210 202 1.94E-01 190 2.43E-01 3.48 4.63E-02

Po-210 281 2.70E-01 231 2.96E-01 Not applicable 0.0

Ra-224 Not applicable 0.0 Not applicable 0.0 0.37 4.93E-03

Ra-226 477 4.58E-01 263 3.36E-01 3.87 5.15E-02

Ra-228 Not applicable 0.0 Not applicable 0.0 0.41 5.46E-03

Th-228 2.28 2.19E-03 7.36 9.42E-03 0.75 9.99E-03

Th-230 68.9 6.61E-02 76.2 9.75E-02 60.2 8.01E-01

Th-232 1.11 1.07E-03 0.99 1.27E-03 0.84 1.12E-02

Total 1,041.64 Not 

applicable

781.6 Not 

applicable

75.11 Not 

applicable

raffinates in the silos contained low levels of uranium. Thorium-230 was present in each silo in similar 

concentrations, indicating that the uranium mills effectively removed radium but were not effective in 

removing thorium from the yellowcake. The total activity concentration in silo 3 was approximately 

10% of the activity concentration in silos 1 and 2. However, 230Th is a dominant radionuclide in Silo 3 

and requires consideration for dose evaluation purposes. The concentration of 230Th is essentially the 

same in all three silos. The 230Th in Silo 3 constitutes a high fraction of the total activity in comparison 

with the fractions in silos 1 and 2, due not to high 230Th levels, but due to the absence of radium and 

its progeny. If the 226Ra and progeny were removed from silos 1 and 2, the 230Th would constitute 

about 80% of the remaining relative hazard, as it does in the Silo 3 material.

RAC (1998) provides estimates of the annual emission of 222Rn from the K-65 silos for the period from 

1952 through 1988. The 95th-percentile value of these estimates, which defines upper estimates of 

emission rates, has been used in the analysis to calculate the annual environmental intake rates 

shown in Table 4-2 in ORAUT-TKBS-0017-4, Feed Materials Production Center – Environmental 

Dose (ORAUT 2015b).

5.3.4.1 Radon Breath Analysis

During 1952 through 1954, to infer radium intakes, 609 breath samples were taken for radon at the 

site. The samples were sent for analysis to the AEC Health and Safety Laboratory (HASL) in New 

York City. At this time, no specific information is available about the measurement method; the 

assumption is that the ambient radon concentration present during the sampling was included in the 

result. Therefore, any dose reconstruction based on these uncorrected results would include a bias 

favorable to the claimant.

A typical radon breath analysis result identifies the individual and provides the radon breath 

concentration measurement results in curies per liter. Annual and composite parametric values for 

the data are shown in Table 5-14. In all years, the minimum reported value is 1 × 10-13 Ci/L, which is 

assumed to be the reporting limit for the analysis. When the reported result is “<1 × 10-13 Ci/L” a 

value half of that (5 × 10-14 Ci/L) was used in calculations. This method is endorsed in NIOSH (1993), 

which is incorporated by reference in the implementing rules for EEOICPA [42 CFR Part 82.16(a)]. 

Approximately 25% of the values were imputed by this method. In addition, approximately 26% of the 

submitted samples were not successfully analyzed due to processing errors or equipment failures. 
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Table 5-14. Radon breath analysis statistical parameters.

Parameter 1952 1953 1954

Composite 1952 

through 1954

Number of attempted measurements 140 238 231 609

N, number of valid measurements 84 183 182 449

Number of valid measurements less than the 

reporting limit

55 54 46 155

Minimum, Ci/L (assumed to be the reporting limit) 5.00E–14 5.00E–14 5.00E–14 5.00E–14

Maximum, Ci/L 9.00E–13 3.30E–12 1.20E–12 3.30E–12

Mean, Ci/L 2.66E–13 2.84E–13 2.47E–13 2.66E–13

Standard deviation 2.01E–13 3.26E–13 1.79E–13 2.53E–13

Coefficient of variation (standard deviation ÷ mean) 0.76 1.14 0.72 0.95

GM, Ci/L 2.01E–13 2.07E–13 1.91E–13 1.99E–13

GSD 2.18 2.13 2.11 2.13

Median, Ci/L 2.00E–13 2.00E–13 2.00E–13 2.00E–13

84th percentile, Ci/L 4.00E–13 4.00E–13 4.00E–13 4.00E–13

95th percentile, Ci/L 6.85E–13 6.90E–13 6.00E–13 6.00E–13

Lacking data, these were removed from consideration, resulting in a total of 449 valid samples over 

the period from 1952 through 1954.

ORAUT-OTIB-0025, Estimation of Radium-226 Activity in the Body from Breath Radon-222 

Measurements, provides information on partitioning of the whole-body radium activity among bone, 

lung, and soft tissue compartments (ORAUT 2005a). The radiation dose to various organs from 226Ra 

in the body can be derived by selecting an appropriate intake scenario and applying appropriate dose 

factors to the derived 226Ra activity level. ORAUT-OTIB-0025 defines a whole-body 226Ra activity 

conversion factor for radon breath data as 2.52 × 105 pCi of 226Ra per pCi/L of exhaled 222Rn.

The median whole-body estimate developed from the radon breath analysis is 0.05 μCi of 226Ra 

(assuming a 1-year chronic inhalation).

Fernald Plant 2/3 workers were also exposed to pitchblende ores and their raffinates from about 1954 

through 1958 (Q-11 through Australian pitchblende). However, processing was done in the Fernald 

plant with the resultant raffinates going directly to the K-65 silos. The raffinate handling and transfer 

operation in Plant 2/3 was of a slurry form, and the raffinates were transferred in process lines directly 

to the silos. This operation was quite different from the drum handling described above. The 

bounding process is taken to be the opening of the drums in Plant 2. At this point in the process, the 

uranium, its progeny, and the associated contaminants are all present. Therefore, uranium bioassay 

was adequate and the other components contributing to the dose can be assigned by using ratios to 

uranium.

After the processing of the pitchblende ores ended in 1958, Plant 2/3 workers processed only ore 

concentrates (yellowcake) from ore processing plants (uranium “mills”) with the resulting “cold” 

raffinates. These were transferred in calcined, dry, dispersible powder form to Silo 3, using an 

enclosed air lift. Again, the bounding process is taken to be the opening of the drums in Plant 2. At 

this point in the process, the uranium, its progeny, and the associated contaminants are all present. 

Therefore, uranium bioassay was adequate and the other components contributing to the dose can be 

assigned by using ratios to uranium.
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5.3.4.2 Raffinate Exposures

5.3.4.2.1 K-65 Raffinate Drum Transfer Workers, 1952 to 1956

Radon breath analysis results exist for some workers at Fernald (NLO 1952, 1953a). Radon breath 

analysis results, titled “Personnel Radon Breath Analysis Measurement Results,” are evaluated and 

included in the dose reconstruction as follows:

1. Convert the radon breath analysis result to picocuries per liter.

2. Apply the ORAUT-OTIB-0025 conversion factor for the whole body of 2.52 × 105 pCi/(pCi/L) to 

get a whole-body content of 226Ra in picocuries (ORAUT 2005a).

3. Use the whole-body 226Ra content in IMBA to derive a chronic intake of type M 226Ra from 

January 1, 1952, through the date of the last radon breath analysis result.

4. In addition to the 226Ra intake estimate, apply the derived intake to the isotopes and ratios in 

Table 5-16.

If there are no radon breath analysis results, dose reconstructors should use the 226Ra default intake 

rate from Table 5-15. This is based on the assumption of the median whole-body estimate developed 

from the radon breath analysis (0.05 μCi of 226Ra) and assuming a 1-year chronic inhalation intake. 

The intake rate of 226Ra is then used with activity ratios of 226Ra to other raffinate nuclides to assign 

intakes of the other radionuclides in Table 5-16.

Table 5-15. Radium intake rate based on radon breath analysis data (pCi/d). 

Years Ra-226 intake estimatea IREPb distribution

1952–1954 5,184 Lognormal GSD=3

a. Ra-226 is solubility type M.

b. Interactive RadioEpidemiological Program. 

Table 5-16. Ratios of silos 1 and 2 nuclides to 226Ra. 

Isotope

Activity ratio of nuclide to 

Ra-226 in silo 1 material

Activity ratio of nuclide to 

Ra-226 in silo 2 material

Ac-227 0.016 0.025

Pa-231 Not applicable 0.015

Pb-210 0.424 0.723

Po-210 0.589 0.881

Ra-224 Not applicable Not applicable

Ra-228 Not applicable Not applicable

Th-228 0.005 0.028

Th-230 0.144 0.290

Th-232 0.002 0.004

Table 5-16 shows the ratio of the isotopes in the silo 1 and 2 materials to 226Ra. These ratios are 

based on data in Table 5-13. The ratios from both silos 1 and 2 are similar, but the ratios to the 

radium isotopes from the silo 1 and 2 data should be compared for dose reconstruction, and the most 

favorable to the claimant should be used to assign dose because it is not possible to know which 

material was included with the radium intake. It is not necessary to consider the ratio from silo 3, 

since the processing of the yellowcake (slurry) did not begin in Silo 3 until 1959 (ORAUT 2011a).

These estimates establish an upper bound of intake for workers involved in the transfer operation of 

the 13,000 barrels of the stored MCW raffinates to the K-65 silos. An examination of external 

penetrating radiation dose for workers who were known to have worked with and handled these 
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drums of raffinate wastes show significant penetrating dose accumulation (several hundreds of 

millirem per week). Therefore, a criterion to determine or verify that a worker had indeed been 

exposed to internal intake from raffinate dusts would be a record of penetrating external dose (i.e., no 

detectable dose would clearly indicate little direct contact or work with the barrels of waste). However, 

the assumption that the population of K-65 raffinate workers could be limited by external dose 

because the high 226Ra content would have produced high dose rates did not prove to be true [1]. 

Because a method for identifying workers on this project could not be identified, all workers should be 

assigned an intake. 

Workers opening drums of the K-65 raffinates would have also been exposed to the radon escaping 

from the drums. Natural ventilation outdoors would have decreased the concentrations in a few 

minutes. No radon beyond the environmental doses need be assigned.

5.3.4.2.2 Plant 2/3 Workers Handling or Processing Pitchblende Ores, 1954 to 1958

Dose reconstructors should assign uranium intakes from uranium urine bioassay. Unmonitored 

exposure to the impurities in pitchblende might have occurred to workers who sampled or handled 

these materials. Assume that the 230Th in silos 1, 2, and 3 is in equilibrium with the total uranium 

activity and assign an intake of 230Th that is equivalent to the derived uranium intake.

In addition, use the ratio to 230Th activity of the isotopes in silos 1 and 2 materials from Table 5-17. 

These ratios are based on the activity concentration of raffinate data in Table 5-13. The ratios from 

the silo 1 and silo 2 data should be compared for dose reconstruction, and the most favorable to the 

claimant used to assign dose, if it is unknown to which silo a worker was exposed. Because the 

activity ratios in Silo 3 are lower (and not as favorable to the claimant), a comparison to Silo 3 is not 

necessary. For dose reconstructions with overlapping periods in 1955-1956, a comparison between 

Table 5-16 and Table 5-17 doses are made, and the most favorable to the claimant should be 

selected for assigning dose.

Table 5-17. Ratios of silos 1, 2, and 3 nuclides to 230Th. 

Isotope

Activity ratio of nuclide to 

Th-230 in silo 1 material

Activity ratio of nuclide to 

Th-230 in silo 2 material

Activity ratio of nuclide to 

Th-230 in silo 3 material

Ac-227 0.111 0.087 0.015

Pa-231 Not applicable 0.053 0.010

Pb-210 2.935 2.492 0.058

Po-210 4.078 3.036 Not applicable

Ra-224 Not applicable Not applicable 0.006

Ra-226 6.929 3.446 0.064

Ra-228 Not applicable Not applicable 0.007

Th-228 0.033 0.097 0.012

Th-232 0.016 0.013 0.014

5.3.4.2.3 Plant 1, 2, 3, and 8 Workers Handling or Processing Yellowcake, 1959 to 1961

Unmonitored exposure to the impurities in yellowcake might have occurred to workers who sampled 

or handled these materials. Workers in Plants 1, 2, 3, and 8 might have been exposed. The intake 

rates from these exposures can be bounded by adding an intake from these impurities in proportion to 

the uranium intake as measured by bioassay. The ratio from silo 3 is used for dose reconstruction 

because the processing of the pitchblende ores (whose residues were stored in Silos 1 and 2) ended 

in 1958 (ORAUT 2011a).

Assume that the 230Th in silo 3 is in equilibrium with the calculated uranium activity and ratio the rest 

of the isotopes to 230Th, as given in Table 5-17. These ratios are based on the activity concentration 

of raffinate data in Table 5-13. Beginning in 1961, RU contaminants must also be included.
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Air monitoring results show that no significant exposure to radon from cold raffinates occurred. The 

milling process removed nearly all of the radium from the ore concentrates. Therefore, the 

concentrate materials presented no significant hazard from penetrating radiation or from the 

accumulation of radon gas during drum opening and sampling operations (Christofano and 

Harris 1960).

5.4 RADIOLOGICAL CONTROLS PROGRAM

The initial health and safety organization (the Industrial Hygiene and Radiation Division) at Fernald 

was organized and directed by an occupational medical physician and staffed primarily with industrial 

hygienists. The Fernald staff worked with the Y-12 Plant staff because of similar radiation safety 

issues associated with their uranium processing, although Y-12 typically had a more highly enriched 

uranium source term. Fernald also used Y-12’s Mobile In Vivo Radiation Monitoring Laboratory 

(MIVRML) for over 20 years. A review of the Y-12 radiation protection program and internal dosimetry 

technology (Cofield 1959; Scott and West 1964; Steckel and West 1966; West, Scott, and Schultz 

1979) indicated that the internal dose technology, techniques, procedures, and philosophy similar to 

Y-12’s were used at Fernald. The fact that the Y-12 MIVRML provided routine service to the Fernald 

in vivo internal dosimetry program is an indicator of the close working relationship the Fernald staff 

had with the Y-12 program.

The safety organization’s staff was aware that TRU elements and fission product contaminants were 

present in the RU. The biological concerns were thought to represent an increase of less than 10% in 

hazard level in comparison with uranium that had never been irradiated. Radiological controls were 

determined and adjusted based on routine air sample analysis and urine bioassay for uranium.

Before 1986, routine TRU analyses were not performed for either airborne or urine activity, although a 

few examples of specific analyses of plutonium on air samples were recorded and indicated that the 

uranium concentrations were controlling. Exposure controls were set at levels that prevented 

chemical toxicity of uranium, assuming that these controls were sufficient for all the radiological 

hazards (Bassett et al. 1989). Although the alpha activity from the TRU alpha emitters would have 

been collected and detected in the air samples, the reported results were all considered to be 

primarily uranium and compared to the MAC for uranium (Wing and Halcomb 1959a, 1959b). As a 

consequence, bioassay measurements were not routinely performed on workers for the contaminants 

associated with RU until 1986. Heatherton (1975, p. 159) includes a report that discusses the results 

of the FMPC uranium monitoring program.

Radioactive contamination was routinely measured in occupied areas of the plant, and there were 

significant radioactive material releases to the environment. It was necessary for every facility at the 

Fernald site to address issues involving airborne contamination, even though engineered confinement 

barriers were used in conjunction with process and work area ventilation. Elevated airborne 

radioactivity resulted from processing thousands of metric tons of dispersible radioactive materials 

with a variety of chemical and physical characteristics. Throughout the process history of the site 

there were high potentials for intakes of uranium, thorium, and their radioactive progeny.

A radiological controls program was in place from the beginning of FMPC operations. The internal 

dose control program consisted of:

• An air sampling program in all processing areas to evaluate internal exposure potential via 

inhalation,

• Urine samples submitted after at least a 2-day work break to allow elimination of uranium 

cleared rapidly via the gastrointestinal tract (this material causes relatively little dose), and
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• In vivo analysis once a month for workers with high exposure potential on a frequent urinalysis 

program and once a year for workers with a low potential for internal intake.

Other elements of the protection program included routine monitoring of the workplace and personnel 

for radiation (NLO 1953b) and contamination, personnel protection in the form of protective clothing 

and respiratory protection in all of the operational areas as needed, and restricting workers from 

workplaces with elevated airborne radioactivity concentrations when the level of uranium in the urine 

or in vivo counting results exceeded specified plant action limits.

5.4.1 Air Monitoring Program

The large quantity of data in the archives shows that Fernald maintained an aggressive air monitoring 

program from the beginning of operations (AEC 1955, Halcomb and Huesing 1957). Both high and 

[primarily] low volume GA and BZ air samples were collected (most for 3 to 30 minutes) and counted 

for alpha contamination in the 1950s. A few sample records and claim file records indicate that some 

beta counts were performed. In the 1960s the samples were counted for both alpha and beta activity. 

The results were compared to the NLO MAC guidelines (a.k.a. the NCG) of 100 dpm/m3 [70 dpm/m3

was used as the MAC (NCG) until the 1970s]. The 100 dpm/m3 equates to about 4.5 × 10-11 μCi/cm3

. 

Routine air samples were taken in every plant and operational area. This program was the primary 

means of controlling intakes. Workers were directed to use respiratory protection in the form of dust 

masks or supplied air respirators depending on the anticipated or measured airborne radioactivity 

concentration. From a historical viewpoint, extensive, long-term air activity summary sheets that 

covered 15 or more years were periodically prepared that indicated routine detectable air activity in all 

working areas of each plant. These summaries detailed annual average exposures to workers 

without respiratory protection and average air activities associated with job assignments that required 

respirators. The air activity ranged from a fraction of the MAC levels to hundreds of times those 

levels.

From 1953 to 1986, the air monitoring program was used primarily to establish work controls, such as 

respiratory protection requirements for workers. However, the measured air concentration levels from 

the routine sampling program typically were not used to establish worker intakes, and workers were 

required to submit routine urine samples for uranium analysis only for the purpose of verification of 

site controls based on the air sample. The MDL for routine air sampling was in the 0.02 to 0.2 MAC 

range. When multiplied by an 8-hour workday, an MDL of approximately 1.5 MAC-hr/d is derived 

(based on informal calculations conducted from data taken from recorded sample sheet volume and 

radiation counting data). The most typical sample volume was a 3-minute sample (0.06 m3

) in dusty 

areas, which equates to the 1.6 MAC-hr MDL, and up to a 30-minute sample (0.6 m3

) with an MDL of 

0.2 MAC-hr. BZ samples were generally of the same flow rate but shorter duration. Long-term 

averages at 0.1-MAC levels were reported, but should be interpreted as numerical averages only.

Beginning in 1989, BZ samples were converted to DAC-hours in the case of thorium, uranium, and 

custom nuclide exposures. On occasion, these air sample analyses were converted to intake, dose, 

or both. The calculated doses observed in the records were low [i.e., tens of millirem committed 

effective dose equivalent (CEDE)]. The air sample results are used with the isotope and a conversion 

to a percentage of the DAC, then converted directly to dose in CEDE.

For reference, Table 5-18 lists the abbreviations and codes on the sample and dose calculation 

datasheets.
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Table 5-18. Air monitoring analysis sheet codes. 

Code Description

MAC NLO MAC (NCG) –100 dpm/m3 and 70 dpm/m3 before 1970 

NCG NLO concentration guide–used interchangeably with MAC

R Sample collection rate in m3/min

T Sample time in minutes

Q Sample volume in m3 Q = R × T

GA Sample collected in a GA

BZ Sample collected as close to the BZ as practicable

Analyzed for–alpha Gross alpha count

Analyzed for–beta Gross beta count

Analyzed for–Ra Alpha count on radium sample separation

Analyzed for–Th (#33) Alpha count on thorium sample separation

Analyzed for–all others Chemical analysis for nonradiological samples

Later (1993 to 2006) air monitoring was performed with the intent of calculating intake and dose if 

necessary for low dose cases in addition to maintaining site control of personnel exposures. The air 

sample data were used to calculate a specific concentration of a specific nuclide in terms of 

microcuries per cubic centimeter and conversion to a specific DAC percentage. This value was 

directly converted to intake and organ dose, as appropriate.

Table 5-19 cross-references Fernald air monitoring methods with the applicable MDLs.

Table 5-19. Air monitoring methods with applicable MDLs. 

Type of 

analysis Method Period Frequency MDLa,c 

Air sampling BZ sampling 1952–1993 Variable and frequent, 

depending on work

0.02–0.2 MAC depending on 

sample volume

Air sampling BZ lapel 

samplers

1993–2006 Job specific, when Th-230 or 

Th-232 exposure potential 

exists. BZ sampling for 

uranium is also performed for 

work controls.

1–2 DAC-hr with a minimum 

detectable intakeb of 0.001–

0.002 corrected DAC-hr for 

thorium

a. From 1952 to 1993, BZ samples were taken for 5 to 30 minutes at 20 L/min (0.1 to 0.6 m3 of air); after 1993, the lapel 

sampler was typically worn continuously at about 4 L/min for 8 hr/d (approximately 2 m3 of air collected), then counted 

for longer times with more sensitive laboratory counters.

b. Detectable Intake is lower due to the application of a respiratory protection factor.

c. MDLs reported in worker’s records take precedence over the default values in this table. 

Fernald’s HIS-20 database was implemented at the site in the mid-1990s and is the source of BZ air 

monitoring data. Some of the data in HIS-20 were migrated from legacy health and safety databases. 

DOE has also provided data extracted from HIS-20 to NIOSH in the form of database tables. Thorium 

BZ results begin in 1993 in the HIS-20 database. Beginning in 1995 (HIS-20 database), BZ air 

monitoring data are consistently available.

5.4.2 Bioassay Program

A uranium urinalysis program was administered at Fernald starting in 1953 or possibly earlier. 

Uranium is an unusual radioactive material in that its chemical toxicity can be the dominant hazard in 

cases of readily or moderately soluble compounds of depleted uranium (DU), NU, and LEU. Uranium 

is a heavy metal and a sufficient heavy metal exposure can lead to kidney damage.

In 1959, a chemical toxicity threshold concentration of 3 μg U/g of kidney tissue (based on animal 

studies and some human data) was established by the International Commission on Radiation 

Protection (ICRP), and has been used since that time as the reference limit. Concentrations in 
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excess of 3 μg U/g of kidney tissue have been shown to result in renal tubule damage. The no-effect 

level of uranium in the kidney has been calculated to be 1.1 μg U/g of kidney tissue, which averages 

to a kidney burden of 0.337 mg (Rich et al. 1988). Table 5-20 lists a summary of the Fernald 

chemical toxicity limits with their associated radiological limits.

Table 5-20. Chemical toxicity and associated limits.

a,b

Solubility and enrichment

OSHA limit for 

toxicity

(mg/m3 )

OSHA limit for

equivalent

radiation exposure 

(μCi/cm3)

Limit

based on

radiation exposure

(μCi/cm3)

Soluble DU 0.05 1.8E–11 6.0E–10

Soluble NU 0.05 3.5E–11 3.0E–10

Soluble 2% EUb 0.05 8.1E–11 2.0E–11

Insoluble DU 0.25 9.2E–11 6.0E–10

Insoluble NU 0.25 1.8E–10 3.0E–10

Insoluble 2% EUb 0.25 1.8E–10 3.0E–10

a. A 1-milligram intake is approximately 0.337 of the kidney toxicity limit. The intakes are assumed to be 7 mg for type F, 

37 mg for type M, and 696 mg for type S.

b. Italicized entries, and the type S intake of 696 mg, indicate limits that are controlled by radiological concerns.

The early basis for conducting routine urine analysis was to ensure that uranium exposure controls 

were adequate to prevent chemical toxicity. In addition, the urinalysis results were used to restrict 

further worker exposure when control levels were exceeded. Urine samples were taken as frequently 

as weekly (or more frequently as indicated by unusual events). Internal radiation doses, which were 

derived from urine sample results, were not calculated until the mid to late 1980s.

Although the primary exposure control during the early years of operation was based on toxicity 

concerns, some radiological exposures were reported in addition to a few work restrictions based on 

radiological dose limits.

Nearly all employees provided urine samples for uranium analysis at the time of their annual 

physicals. Workers with work assignments such that exposure could be expected on a routine basis 

were sampled weekly, monthly, or (at the least) bimonthly. The frequency for routine sampling for 

operations personnel changed in the early years from quarterly to a variable frequency depending on 

previous urinary uranium results, knowledge of plant operations, and possible exposure. This 

frequency might have varied from 3 times per year to once per month for any particular worker. 

Routine samples were collected at the beginning of the shift, but not necessarily at the beginning of 

the workweek. In the early years, any routine result that exceeded 40 μg/L of uranium was followed 

up by obtaining start-of-shift urine samples for 3 successive workdays. If one of these results were 

greater than 40 μg/L, the individual’s dose was investigated for cause (Heatherton 1975, p. 160)

As early as 1958, FMPC reported internal dose exposure to the AEC in an annual report. Table 5-21 

summarizes the data from 1958 to 1966. The data provide indication of recorded annual urinary 

results in summary form.

The value of fecal samples was recognized even in the early years and has been well understood 

since 1986. As an example, fecal sampling for uranium was performed on several workers in 1968 as 

a part of a DOE HASL-NLO study (as recorded on analytical data sheets of October 4, 1968). 

However, because it was recognized that the natural environmental levels vary considerably, fecal 

samples were determined to be less reliable so were not a part of the routine bioassay program at the 

Fernald site. Fecal samples were collected in special cases and beginning in the mid-1990s, they 

were analyzed under contract with a specified maximum MDA of 0.1 pCi/sample. If lower sample 

MDAs are found in records associated with contracted fecal uranium analyses, these lower MDAs 

should be used.
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Table 5-21. Urinary uranium averages summary. 

Urine average 

(μg/L)a 

Total number of exposed workers in each category

1958 1959 1960 1961 1962 1963 1964 1965 1966

25–30 22 51 125 75 25 26 18 5 3

31–35 18 37 44 41 5 9 9 1 1

36–40 17 21 25 18 5 1 6 0 0

41–45 9 5 15 6 1 1 0 0 0

46–50 7 4 6 3 1 0 0 1 0

51–55 5 3 7 1 0 0 0 0 0

56–60 2 1 3 1 0 0 0 0 0

61–70 3 4 4 0 0 0 0 0 0

71–90 3 5 0 0 0 0 0 0 0

91–110 3 3 0 0 0 0 0 0 0

111–155 45 2 0 0 0 0 0 0 0

Total 94 136 229 145 37 37 33 7 4

a. Various annual reports reported the units in mg/L, which is an obvious typographical error. The permissible urine 

concentration, averaged throughout the year, was 0.05 mg/L (which is equivalent to 50 μg/L).

In vivo lung counts were routine from 1968 to 1989 using a MIVRML from the Y-12 Plant in Oak 

Ridge, Tennessee. Typical results of the MIVRML were uranium lung burdens, although a few 

thorium analyses results were reported at specific times. The results were reported in micrograms of 

235U and milligrams of total uranium (mg U) and were recorded in individual dosimetry files and 

summarized in formal reports. TRU analyses were attempted; however, the limitations of the 

MIVRML, which included limited detection sensitivities for TRU isotopes and infrequent counts, and

lacked the ability to detect the anticipated levels at Fernald or in fact could not detect levels that met 

regulatory limits for TRU isotopes.

From 1989 to 2002, the lung counts took place in the Fernald In Vivo Examination Center (IVEC) 

(Waligora 2003). Results from the IVEC were reported in nCi if a nuclide was detected.

5.4.3 Uranium Urinalysis Program

The urinalysis procedure in place until 1993 was conducted using the fluorometric fusion process, 

which fused uranium from raw urine with sodium fluoride and measured the fluorescence created by 

ultraviolet light. This method provided a measure of the total amount of elemental uranium in the 

sample. Comparison studies between the other laboratories provided assurance that the analytical 

process was consistent with industry standards. The MDA was not formally established in the early 

periods (as evidenced by a records search). The values listed in the bioassay data sheets generally 

range from 0.003 mg/L to 0.008 mg/L in urine. The effective MDA is now understood to have been 

near 0.014 mg/L.

In early 1993, other DOE laboratories were changing to kinetic phosphorescence analysis (KPA), with 

detection capabilities as low as 0.02 μg/L, which is far below the natural environmental background in 

most people. The MDA of 0.17 μg/L was established consistent with other laboratories. A decision 

level (DL) of 0.8 μg/L at Fernald since 1993 represents the best estimate of the nonoccupational 

excretion of uranium, although values below the DL are reported in the dose history records. 

Specifically, when Fernald changed to the KPA, 0.8 μg/L became the reporting level until 1997, after 

which all calculated results were reported as derived. Some database printouts for samples after 

1997 include less-than symbols (<) for results that are less than 0.17 μg/L (e.g., <0.06 μg U/L). These 

results should be interpreted as less than 0.17 μg/L. This represents a database formatting issue 

rather than a lowering of detection limits.

The method used at Fernald from 2002 to 2006 for urinalysis was inductively coupled plasma mass 

spectrometry (ICP-MS), which has an a priori MDA of 0.15 μg/L.
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Table 5-22 lists the MDLs, which remained essentially unchanged throughout the history of the site.

Table 5-22. Routine uranium in vitro bioassay capabilities. 

Type and method of analysis Period Frequency MDL

Urine, uranium, 

fluorophotometry

1952–1993 Weekly to annual–

job specific

14 μg/L(a)

Urine, uranium, Chemchek KPA 1993–09/2002 Bimonthly 0.17 μg/L (total U)

Urine, uranium, ICP-MS 09/2002–2006 As requested 0.15 μg/L (total U)

Feces,

b fluorophotometry Various No routine 

schedule 

Unknown; assume environmental 

levels of 2 μg/sample 

Feces,

b contract laboratory Occasional No routine 

schedule

0.1 pCi/sample

a. Y-12 listed a sensitivity of 1.6 μg/L in 1973 using the fluorometric process for 0.7% U-235. Fernald frequently listed less 

than 0.003 mg/L in the bioassay data reports. Several blank samples on intercomparison studies also list results as 

0.003 mg/L. A value of 0.008 mg/L has also been quoted in the records as the MDL. However, a formal response on 

January 21, 1993 (Blalock and Neton 1993), to a deficiency in the ability to detect 100 mrem CEDE with the existing 

0.014-mg/L MDA is accepted as the most reliable representation for historical MDAs for this analytical procedure.

b. Fecal sampling was performed as a part of special studies or specific incident investigations on occasion. Records of 

this activity are generally not present in the dosimetry records for individuals.

To be favorable to the claimant, intakes of uranium due to environmental background are not 

considered. That is, any detectable uranium is assumed to be due to occupational intakes. All 

uranium bioassay results (whether urine or fecal) that are greater than the MDA are assumed to be 

indicative of occupational exposure, regardless of any larger value that the site might have used.

Table 5-23 lists the uranium urinalysis sample type codes, and Table 5-24 lists sample type coding 

system and personnel assignment codes.

Note that Code 30 urine samples are believed to have been spot samples rather than 24-hour 

samples. All uranium bioassay samples should be considered to be spot samples, unless there is 

something in the record to indicate otherwise. Because the results are almost always reported in 

micrograms per liter rather than micrograms per sample, this will typically not make a difference when 

evaluating the results. However, this needs to be considered for incidents and for days when there 

are multiple results.

5.4.4 In Vitro Bioassay Procedures for Other Radionuclides

The fundamental and primary bioassay for the first 35 years (1951 through 1986) of Fernald 

operational experience was urine analysis for uranium metal, reported in milligrams per liter. 

Radionuclides other than uranium have been analyzed on occasion through the years, predominantly 

by contract laboratories. Even for those special cases, they have been so few in number that the 

review of records for this TBD did not reveal a dose record with nonuranium urinalysis results. There 

are records of special studies, but no documented intent to analyze for radionuclides other than 

uranium. The primary contract laboratory for FMPC in vitro analyses was United States Testing 

Company in Richland, Washington. Table 5-25 lists MDAs for various nonuranium radionuclides.

5.4.5 Gross Beta Urine Sample Results, 1957

In 1957, there appeared to be a concern for workers in Plant 5 who were suspected of unusual 

intakes due to the presence of recorded gross beta urine results. These urine samples were analyzed 

for uranium, and an aliquot from the sample was then counted for gross beta. An expressed concern 

was noted in a 1957 monthly report of “unusual uranium intakes.” FMPC likely counted an aliquot for 

beta due to their sensitivity for “other” isotopes, although it is doubtful that the concern was specifically 

for thorium – perhaps a concern for fission products, considering that Plant 5 had not been involved in 

thorium processing, as far as is recorded.
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Table 5-23. Fernald uranium urinalysis sample type codes. 

Code Description

00 No Code

10 Pre-Employment Sample – collected during pre-employment physical 

exam or before beginning work on first day. Establishes individual U 

background

20 Annual Sample – collected as a part of each employee’s annual 

physical exam

30 Routine Sample – samples from plant workers who are on a routine 

schedule for the purpose of insuring that airborne levels of uranium in 

the work place are being controlled within safe limits

40 Incident – Follow-up Sample – samples from employees involved in 

an event or circumstances which presents a potential for elevated 

exposure

49 An incident sample left at the end of the shift on the day of the 

incident.

50 Special Sample – samples collected as a part of a study to provide 

data related to uranium exposure and/or excretion characteristics of 

specific areas and/or conditions. Also includes samples taken from 

subcontractors. In general, results for subcontractors are not in the 

HIS-20 database before 1986.

5C Special Correlation Sample

60 Termination/Retirement Sample – sample obtained from employee 

during post-employment physical exam.

70 Rehire – sample obtained during a former employee’s physical exam 

before being re-employed

R Recall Sample (example: 3R – Routine Sample Recall) – samples 

taken to verify positive sample results and/or to follow elimination 

pattern

VF Visitor First Sample

VR Visitor Routine Sample

VE Visitor Exit Sample

BDL Below Detection Level

NA Not applicable/available

These beta counts are recorded as “gross” beta, which could contain 40K and a variety of other 

“natural” contaminants (because they were in the height of the atmospheric weapons testing fallout, 

the generation of which started to decline in 1962 to 1963). Claimant files typically show that those 

who worked beyond 1968 had routine in vivo lung counts with reported thorium levels. Therefore, the 

gross beta counts are not well-enough defined to be of value. There are hundreds of uranium urine 

analyses, a defined RU contaminant default approach, thorium lung counts, and a defined intake 

analyses based on DWE air sampling analyses before 1968. This appears to be a justifiable basis for 

ignoring the 2 to 4 gross beta counts, which resulted from a single occurrence in Plant 5 [2].

5.4.6 In Vivo Bioassay

Lung counting became available to FMPC in 1968 in the form of the MIVRML. The mobile van visited 

the Fernald plant on a routine schedule and counted the workers on a schedule based on their 

internal exposure potential and their urine sampling results. When lung counting became available, 

the annual reports to AEC listed the number of workers who exceeded 50% of the maximum 

permissible lung burden (MPLB) and the calculated annual doses to the lung in rem. The uranium 

MPLB was 0.0175 μCi, assuming 1% enrichment. Table 5-26 presents the typical reporting of internal 

dose as determined by direct lung counting.
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Table 5-24. Fernald sample type coding system and personnel 

assignment codes.

Sample type coding system – specific

Code Definition

1 During 1st hour of shift

2 During 2nd hour of shift

3 During 3rd hour of shift

4 During 4th hour of shift

5 During 5th hour of shift

6 During 6th hour of shift

7 During 7th hour of shift

8 During 8th hour of shift

9 After completing work

Sample type coding system – further definition

Code Definition

5A Off the job, overnight composite sample

5B Off the job, overnight individual specimen

5C Sample to test for possible correlation with abnormal clinical lab 

findings

5D 24-hour individual sample from confined patients

5E 24-hour composite sample from confined patients

5F 24-hour individual sample from unconfined patients

5G 24-hour composite sample from unconfined patients

5H On-the-job individual sample collected in the work area

Personnel assignment codes

Code Definition

150 Plant 6 Inspection

200 General Project (Plant 8)

210 Plant 1

220 Plant 2 and 3

240 Plant 4

250 Plant 5 and 9

261 Plant 6 Rolling Mill

262 Plant 6 and 9

2623 Plant 6 and 9 Chemical

2625 Plant 6 and 9 Machining

270 Pilot Plant

280 Plant 8

430 Boiler Plant

436 Mechanical Department

452 Cafeteria

455 Security Department

462 Stores

463 Transportation Department

465 Garage

466 Service – Porters & Laundry

The Industrial Hygiene & Radiation Department had Internal Deposition Action Levels procedures 

from about 1970 that indicate actions related to the determination of percent MPLB to either uranium 

or thorium. Uranium-235 was detected primarily by the emission of its 186-keV photon. Uranium-238 

was calculated from measurement of the 234Th progeny assumed to be in equilibrium with the 238U. 

Thorium-232 and 228Th activities were determined based on equilibrium assumptions and detection of 

their progeny, most likely 228Ac for 232Th, but 212Pb might have been used for assessment of both 

thorium isotopes. Thorium-230 is not readily detectable by in vivo measurements. There appeared to 

be no attempt to detect TRU contaminants with the MIVRML. The only quantification routinely made 
Document No. ORAUT-TKBS-0017-5 Revision No. 02 Effective Date: 09/30/2016 Page 43 of 189

Table 5-25. MDAs for nonuranium radionuclides.c

Type of analysis Method Period Sample frequency MDA

Urine–plutonium

(Pu-239/240)

Chemical extraction 

and gross count

Before 1988 No routine schedule–

unknown

~0.1 pCi/sample

Urine–plutonium

(Pu-239/240)

Extraction and alpha 

spectroscopy count

1990s–2006 Special study samples 

only

0.1 pCi/sample

Urine–plutonium

(Pu-238/239/240)

Extraction and alpha 

spectroscopy count

1986–1990 Workers potentially 

exposed to POOS in 

Plants 4 and 8

0.02 dpm/sample 

(0.22 pCi/sample)b

Fecal–plutonium

(Pu-239/240)

Extraction and alpha 

spectroscopy count

1990s–2006 Special study samples 

only

0.1 pCi/sample

Urine–thorium

Th-230 or Th-232

Extraction alpha 

spectroscopy count

1990s–2006 Not performed 0.1 pCi/sample

Fecal–thorium

Th-230 or Th-232

Extraction and alpha 

spectroscopy count

1990s–2006 Only in cases of 

significant exposure

0.1 pCi/sample

Urine–radiuma

Ra-226

Unknown 1952–1953 Special study samples 30 dpm/sample

Urine–radiuma

Ra-226

Unknown 1954–1976 Special study samples 1.3 dpm/sample

Urine–radiuma

Ra-226

Unknown 1971–1989 Special study samples 0.07 dpm/sample

Urine or fecal –

radium (Ra-226)

Unknown 1997–2006 Contingency only 0.1 pCi/sample

a. Contractor for radium samples not known. The MDAs listed are typical of existing industry standards.

b. Initial contractual MDA; this MDA could not be maintained and was increased to 0.13 dpm/sample (Bassett et al., p. 76). 

Suspected positive urine samples were followed with additional urine samples, in vivo counts, and fecal samples.

c. MDAs for the methods are found in Tomes (2001). For plutonium in urine and in feces, by extraction and alpha 

spectroscopy count, for the period from the 1990s to 2006, the value is on p. 57. That for thorium in feces is also on 

p. 57. For radium-226 in urine or feces, the value is on p. 93.

Table 5-26. Typical reporting of internal 

dose determined by direct lung count.

Year Number above 50% MPLB

1974 21

1975 21

1976 9

1977 8

1978 10

1979 13

1980 6

with the MIVRML was uranium lung burden in units of micrograms of uranium and occasionally 

thorium activity. Uranium quantification required an assumption about uranium enrichment. The 

results from the MIVRML were reported in micrograms of 235U and milligrams of total uranium in the 

lung, which was translated to MPLB based on the assumed enrichment (generally 1%). The percent 

lung burden was then multiplied by 15 rem to obtain the assigned annual lung dose in rem.

Workers who had known exposures to high air concentrations, had high urine results, or were 

involved in an incident were counted on first priority each time the MIVRML visited the site. Other 

workers were counted based on their job exposure potentials, as shown in Table 5-27.

The IVEC operated at the Fernald site from 1989 to 2002. A subject with a 2.5-cm chest wall 

thickness had the MDAs listed in Table 5-28 at the 95% confidence interval for a 3,600-second count. 

The previous mobile counting system, which serviced Fernald from 1968 to 1989, provided reports to 

the site. However, no system performance characteristics have been discovered to date. Table 5-28 

lists the MDAs for Fernald in vivo analyses.
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Table 5-27. Typical MIVRML counting schedule at Fernald in the 1970s.

Labor category description In vivo counting schedule

All chemical operators Once per year

Members of Project Labor Pool Once per year

Mechanical Department crafts,

Mechanical Department laborer,

Laundry group,

Industrial truck operator,

Locomotive operator,

Switchman,

Graphite shop machinist,

Machine tool operator,

Degreaser,

Crane operator,

Stamper,

Plant 6 laborer,

Furnace operator heater,

Mill man,

Decontaminator,

Transportation laborer

During each MIVRML visit, 25% of the employees in these 

classifications were scheduled to be counted, and each worker 

would be counted at least once during a 2-year period.

Cafeteria,

Water treatment group,

Power plant group,

Heavy equipment operator,

Motor vehicle operator,

Stores Warehouse attendant,

Checker,

Industrial mechanic,

Security Police officer,

Porter,

Toolmaker,

Machine set-up,

Tool room machinist,

Gauge set-up,

Inspector

Salaried personnel and workers in these classifications were 

not routinely counted because of low chronic exposure and low 

potential for unobserved acute exposures.

Summaries of the chest count MDAs by uranium enrichment and mode of counting are shown in 

Tables 5-29 and 5-30 for the MIVRML and IVEC facilities, respectively.
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Table 5-28. In vivo bioassay MDAs for the lung.

Radionuclide Facilitya

MDA

(nCi)

DU MDA

(mg)

NU MDA

(mg)

2% EU MDA 

(mg)

U-235 MIVRML

1968–1989

Not available Not available Not available 100 μg (20%)d

U-238 MIVRML

1968–1989

Not available Not available Not available 6.5b (6 nCi for 

1% EU)

U-238 (Th-234) FMPC IVECc

1989–2001

2.5 7.4 (3 nCi) 7.5 (5 nCi) 7.6 (12 nCi)

U-235 FMPC IVECc

1989–2001

0.18 36 (15 nCi) 11.3 (8 nCi) 4.2 (7 nCi)

Pu-239 FMPC IVECc

1989–2001

190 Not available Not available Not available

Am-241 FMPC IVECc

1989–2001

0.25 Not available Not available Not available

Ra-226 FMPC IVECc

1989–2001

~3.0 Not available Not available Not available

a. The values for the FMPC IVEC MDAs are taken from Soldano (1997). The U-235 and U-238 values are on p. 45. 

Values for Pu-239 and Am-241 are on p. 63. Values for Ac-228 and Pb-212 are on p. 81 and are based on an 

assumption of equilibrium with Th-232. The MDA for Ra-226 is on p. 92.

b. The recorded MDA in mass quantities were reported in Scott et al. (1969, p. 7) on the basis of the enrichments 

assumption shown - and converted to activity units in this table. Values reported in a worker’s record were below 10% 

MPLB, which implies an MDA of less than 1 nCi total 1% (assumed) U-235.

c. IVEC MDAs are from site internal dosimetry TBDs (Tomes 2001; Soldano 1997; WMCO 1990b). 

d. The MDA is based on 20% enrichment, but is being applied to 2% enrichment from Scott et al. (1969, p. 7).
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2 Effective Date: 

09/30/2016 Page 46 of 189

Table 5-29. Summary of MIVRML chest count MDAs by uranium enrichment, 1968 to 1989.

Enrichment Isotopes Mass % Activity %

Total U SA 

(pCi/μg)

Isotopic MDA 

(μg)

Isotopic MDA 

(pCi)

Total U 

MDA 

(pCi)

Total U 

MDA 

(mg)

DU U-235 0.20% 1.07% 0.402 100 215 20,100 50.00

DU U-238 99.80% 83.42% 0.402 6,500 2,184 2,618 6.51

NU U-235 0.72% 2.28% 0.683 100 216 9,481 13.88

NU U-238 99.27% 48.86% 0.683 6,500 2,185 4,472 6.55

1% EU U-235 1.00% 2.10% 0.973 100 204 9,730 10.00

1% EU U-238 98.99% 34.20% 0.973 6,500 2,185 6,389 6.57

2% EU U-235 2.00% 2.68% 1.616 100 217 8,080 5.00

2% EU U-238 97.98% 20.40% 1.616 6,500 2,187 10,721 6.63

Table 5-30. Summary of IVEC chest count MDAs by uranium enrichment, 1989 to 2002.

Enrichment Isotopes Mass % Activity %

Total U SA 

(pCi/μg)

Isotopic MDA 

(nCi)

Total U MDA 

(nCi)

Total U 

MDA 

(pCi)

Total U 

MDA 

(mg)

DU U-235 0.20% 1.07% 0.402 0.18 17 16,822 41.85

DU U-238 99.80% 83.42% 0.402 2.5 3 2,997 7.45

NU U-235 0.72% 2.28% 0.683 0.18 8 7,895 11.56

NU U-238 99.27% 48.86% 0.683 2.5 5 5,117 7.49

1% EU U-235 1.00% 2.10% 0.973 0.18 9 8,571 8.81

1% EU U-238 98.99% 34.20% 0.973 2.5 7 7,310 7.51

2% EU U-235 2.00% 2.68% 1.616 0.18 7 6,716 4.16

2% EU U-238 97.98% 20.40% 1.616 2.5 12 12,255 7.58
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5.5 DOSE RECONSTRUCTION

5.5.1 Uranium Dose Assessments

Note that employees at the FMPC who were not employed by NLO (or by DOE or its predecessor 

agencies), and who worked at FMPC from January 1, 1951, through December 31, 1983, are included 

in the SEC due to the fact that their uranium urinalysis results might not have been included in the 

database used for the coworker study. While internal uranium doses for members of this class cannot 

be completely reconstructed, NIOSH intends to use any internal monitoring data that may become 

available for an individual claim (and that can be interpreted using existing NIOSH dose 

reconstruction processes or procedures). 

5.5.1.1 Bioassay Data

A uranium urinalysis program was administered at Fernald starting in 1952. Data should be available 

for exposed workers because of the extensive urine bioassay program at Fernald. Table 5-31 

contains a summary of the MDAs for routine urinalysis samples. See Section 5.4.3 and Table 5-29 for 

additional information.

Table 5-31. Routine uranium urinalysis capabilities. 

Method of analysis Period MDA

Fluorophotometry 1952–1993 14 μg/L

Chemchek KPA 1993–09/2002 0.17 μg/L (total U)

ICP-MS 09/2002–2006 0.15 μg/L (total U)

In vivo bioassay, in the form of lung counts, was also performed. From 1968 to 1989, measurements 

from the Y-12 MIVRML are available. From 1989 to 2002, the lung counts took place in the Fernald 

IVEC. Refer to Tables 5-29 and 5-30 for the MDAs for each of these counters.

5.5.1.2 Uranium Breathing Zone Results

Uranium urine results and chest count measurements take precedence over air data, and uranium BZ 

results do not have to be evaluated if there are positive urine or chest count measurements. 

However, when those results are all negative and there are positive uranium BZ results, include the 

intake based on BZ sample results in addition to the missed dose based on urine or chest count 

measurements. Ingestion intakes are also assigned when only BZ data are used to assign inhalation 

intakes (i.e. there are no in vitro or in vivo bioassay), as described in OCAS-TIB-009, Estimation of 

Ingestion Intakes (NIOSH 2004).

5.5.1.3 Uranium Enrichment

For converting uranium mass to activity, use the values in Table 5-2. Table 5-32 lists the default 

assumptions for cases where the uranium enrichment is not known (this will be most, if not all, cases).

Table 5-32. Default uranium enrichment. 

Years Enrichment

1952–1964 1% 

1965–2006 2% 

5.5.1.4 Recycled Uranium

Starting with 1961, RU components should be added to uranium intakes. Intakes are assigned by 

applying the Table 5-33 (ratio to uranium mass) or Table 5-34 (ratio to activity) contaminants and 

ratios to the calculated uranium intake. The selected material types for RU contaminants are 
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assigned using the direction in ORAUT (2014a). These contaminants are added to all uranium 

intakes regardless of monitoring method (urine, chest counts, BZ air samples, etc.). 

Table 5-33. RU contaminant intakes per unit mass of uranium.

Isotope

1961–1964

(Bq/g U)

1965

(Bq/g U)

1966–1972

(Bq/g U)

1973–1975

(Bq/g U)

1976–1985

(Bq/g U)

1986 on

(Bq/g U)

U 3.61E+04 5.98E+04 5.98E+04 5.98E+04 5.98E+04 5.98E+04

Pualphaa 2.96E+01 2.96E+01 2.93E+01 1.17E+03 1.16E+03 1.15E+03

Pu-241 3.05E+02 3.05E+02 1.88E+02 7.53E+03 4.66E+03 2.89E+03

Am-241 1.96E–02 1.96E–02 3.85E+00 1.54E+02 2.46E+02 3.00E+02

Np-237 1.02E+01 1.02E+01 1.02E+01 2.81E+02 2.81E+02 2.81E+02

Tc-99 3.77E+03 3.77E+03 3.77E+03 1.26E+04 1.26E+04 1.26E+04

Th-232 4.07E–02 4.07E–02 4.07E–02 4.07E–02 4.07E–02 4.07E–02

Th-228 2.85E–02 2.85E–02 2.85E–02 2.85E–02 2.85E–02 2.85E–02

Ru-106 4.08E+03 4.08E+03 4.08E+03 4.08E+03 4.08E+03 4.08E+03

Zr-95 6.12E+02 6.12E+02 6.12E+02 6.12E+02 6.12E+02 6.12E+02

Nb-95 6.12E+02 6.12E+02 6.12E+02 6.12E+02 6.12E+02 6.12E+02

Sr-90 1.63E+02 1.63E+02 1.63E+02 1.63E+02 1.63E+02 1.63E+02

a. The plutonium alpha mixture is assessed as 100% Pu-239. Am-241 and Pu-241 are assessed based on 6% weaponsgrade plutonium mixture.

Table 5-34. RU contaminant intakes per unit activity of uranium (ORAUT 2016). 

Isotope

1961–1964

(Bq/Bq U)

1965

(Bq/Bq U)

1966–1972

(Bq/Bq U)

1973–1975

(Bq/Bq U)

1976–1985

(Bq/Bq U)

1986 on

(Bq/Bq U)

U 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Pu-alphaa 8.21E–04 4.94E–04 4.90E–04 1.96E–02 1.94E–02 1.93E–02

Pu-241 8.47E–03 5.10E–03 3.15E–03 1.26E–01 7.80E–02 4.83E–02

Am-241 5.45E–07 3.28E–07 6.44E–05 2.57E–03 4.11E–03 5.02E–03

Np-237 2.84E–04 1.71E–04 1.71E–04 4.70E–03 4.70E–03 4.70E–03

Tc-99 1.05E–01 6.31E-02 6.31E-02 2.10E–01 2.10E–01 2.10E–01

Th-232 1.13E–06 6.81E-07 6.81E-07 6.81E–07 6.81E–07 6.81E–07

Th-228 7.91E–07 4.76E–07 4.76E–07 4.76E–07 4.76E–07 4.76E–07

Ru-106 1.13E–01 6.82E-02 6.82E-02 6.82E-02 6.82E-02 6.82E–02

Zr-95 1.70E–02 1.02E–02 1.02E–02 1.02E–02 1.02E–02 1.02E–02

Nb-95 1.70E–02 1.02E–02 1.02E–02 1.02E–02 1.02E–02 1.02E–02

Sr-90 4.53E–03 2.73E–03 2.73E–03 2.73E–03 2.73E–03 2.73E–03

a. The plutonium alpha mixture is assessed as 100% Pu-239. Am-241 and Pu-241 are assessed based on 6% weaponsgrade plutonium mixture.

5.5.1.5 Chemical Form

If the chemical compound is known for a particular exposure scenario, then Table 5-3 can be 

consulted for uranium chemical forms. However, dose reconstructors must typically evaluate each 

absorption type to determine the type most favorable to the claimant.

5.5.1.6 Unmonitored Uranium Workers

For unmonitored uranium workers, dose reconstructors should use the uranium intake rates from 

Attachment C, Internal Dosimetry Uranium Coworker Data, in combination with the special 

instructions (e.g., the addition of RU contaminants) in this section and subsections.
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5.5.2 Thorium and Thoron Dose Assessments

Thorium processing did not occur in 1953; therefore thorium dose reconstruction is not required for 

1953. There have been two SEC classes approved because of the infeasibility of constructing the 

internal dose due to thorium. Therefore, thorium doses for 1954 through 1978 are also not 

reconstructed.

The method for assigning thoron (220Rn) does not rely on those used for thorium assessment, so 

although thoron exposures may be associated with thorium, thoron dose is assigned during the 

thorium SEC periods.

The dose reconstruction method varies depending on the available data and timeframe. In many 

instances thorium workers were routinely monitored and chest count data are available. In other 

instances BZ air sampling data are available, and in other times the demonstrated effectiveness of 

radiological controls makes it possible to assume that no worker was exposed to concentrations in 

excess of 10% of the derived air concentration (DAC). Sustained airborne activity levels in excess of 

10% of the DAC triggered airborne area radiological control posting.

Table 5-35 shows the dose reconstruction method for various times and different situations of data 

availability.

Table 5-35. Methods for thorium dose determination. 

Timeframe

In vivo data 

exists (yes or no) Evaluation method Section

1979–1994 Yes In vivo results 5.5.2.1.1

1979–1989 No Coworker data 5.5.2.3.1

1990–1994 No Assume full-time occupancy in areas at 10% class W 

232Th DAC

5.5.2.3.2

1995–2006 Yes Evaluate in vivo data to discover previous intakes. If a 

previous intake is discovered, use BZ data. For new 

intakes, use positive in vivo data instead of negative BZ 

data. If in vivo results are below MDA, use BZ data. 

5.5.2.1.3

1995–2006 No Use BZ data 5.5.2.1.2

5.5.2.1 Monitored Thorium Assessment Methods, 1979 to 2006

5.5.2.1.1 In Vivo Bioassay, 1979 to 1994

As previously discussed, in vivo lung counts were performed from 1968 to 1989 using the Y-12 Plant 

MIVRML. From 1989 to 2002, the counts took place in the Fernald IVEC. Thorium-232 and its 

progeny 228Th cannot be detected when inside the body, but the gamma emissions of the subsequent 

228Ac and 212Pb progeny, are of sufficient energy to be detected. For this reason, in vivo bioassay for 

232Th is often performed by quantifying the activity of the 228Ac or 212Pb in the body and then 

calculating the amount of 232Th (and relatively long-lived progeny 228Ra and 228Th) present by making 

assumptions about the relative ratios of the nuclides.

The FMPC 228Ac and 212Pb chest count results appeared to exhibit a bias, so an analysis of the data 

was performed and bias correction value calculated. MDAs were also generated. Attachment D, 

Evaluation of Fernald Actinium-228/Lead-212 Chest Count Data, contains the details of this analysis. 

Table 5-36 summarizes the bias corrections while Table 5-37 summarizes the MDA values for 228Ac 

and 212Pb. MDAs from Table 5-37 should be used to evaluate chest count measurements unless 

there are claim-specific chest count MDAs. Positive chest count results (i.e., those >MDA) are 

adjusted by subtracting the bias values in Table 5-36 before the results are used.
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Table 5-36. Summary of bias corrections for 

228Ac and 212Pb (nCi) in chest counts.a 

Year 212Pb bias 228Ac bias

1979 –0.014 0.031

1980 –0.045 0.025

1981 –0.056 0.022

1982 –0.058 0.010

1983 –0.101 –0.008

1984 –0.089 0.001

1985 –0.086 0.017

1986 –0.085 0.020

1987 –0.047 0.008

1988 0.002 0.058

1989–2001b 0.072 0.182

a. These values are to be subtracted from a positive 

(i.e., >MDA) result.

b. ORAUT (2014d) provides the basis for these 

values.

Table 5-37. Thorium chest count MDAs (nCi), 1979 to 

2001. 

Counter Years 212Pb MDA 228Ac MDA

MIVRML 1979–1988 0.24 0.24

IVEC 1989–2001 0.42 1.10

Because IMBA is not designed to be used with multiple-nuclide chains, it cannot be used for an exact 

assessment when progeny are used to determine the parent intake. However, an approximation can 

be made that can be used for many cases. ORAUT-OTIB-0076, Guiding Reconstruction of Intakes of 

Thorium Resulting from Nuclear Weapons Programs (ORAUT 2014b), derives the basis and provides 

examples of how to evaluate thorium chest count data using IMBA. Assumptions about the process 

history of the material are made such that the derived intakes are favorable to the claimant. In most 

cases, the 212Pb activity is a more reliable indicator of a thorium intake, so it is used for the thorium for 

dose reconstruction. Triple-separated thorium is assumed because it results in intakes that are 

favorable to the claimant.

Because IMBA does not correctly calculate 232Th and 228Th doses, the following steps apply only to 

the intake assessment. Alternate methods must be used for the dose assessment.

Given a 212Pb chest count result for a chronic intake longer than 1 year or a chest count more than 

30 days after an acute intake:

1. Evaluate the chest burden using the 228Th biokinetic model in IMBA (i.e., use the 212Pb result to 

model the 228Th intake in IMBA).

2. Multiply the intake rate obtained in step 1 by a factor of 1.1 and assign it as the intake rate of 

228Th.

3. Divide the 228Th intake rate by 0.19 to obtain the 232Th intake rate.

If the conditions above (chronic intake longer than 1 year or chest count more than 30 days after an 

acute intake) are not met and an over- or underestimate cannot be performed, contact the PID for a 

best estimate.
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5.5.2.1.2 Breathing Zone Air Samples, 1995 to 2006

Thorium BZ samples are more sensitive than chest count measurements. Therefore, for 1995 

through 2006, thorium doses are typically based on BZ data. Ingestion intakes are also assigned 

when only BZ data are used to assign inhalation intakes (i.e. there are no in vitro or in vivo bioassay), 

as described in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004). See Section 5.5.2.1.3 

for a discussion of how to handle cases with both chest counts and BZ sample results.

When calculating intakes based on BZ data, include the missed dose [see Table 5-19 for minimum 

detection levels (MDLs)]. Use the highest value in the range of MDLs in Table 5-19 to assess the 

missed dose. Reported BZ values less than the MDL are set equal to the MDL (corrected for any 

respiratory protection, to be consistent with the reported measured results). Results are summed and 

assessed as an annual chronic intake rather than being assessed individually.

The solubility class (W or Y) that is in the claimant’s records for each radionuclide and year is used for 

converting reported DAC-hr to activity. The DAC hours “with respirator” column (there are no totals in 

the “without resp” column) should be used, and the DAC-hr for each combination of radionuclide and 

class for each year must be summed before calculating the intake. The solubility type that is most 

favorable to the claimant should then be selected for calculating the dose [3].

For thorium, the BZ results are isotopic for 228Th, 230Th, and 232Th, although there are few 228Th results 

and these were always analyzed in conjunction with 232Th. In addition to isotopic thorium results in 

HIS-20, a review in August 2013 indicated five custom nuclides identified as BL-13, BL-65, CELL 8, 

KS-65 and RT-210 [4]. Fernald used these custom nuclides for known mixtures of nuclides by 

calculating an effective DAC (EDAC) for the mixture to assign a more accurate DAC-hr exposure total 

for BZ air monitoring. The likely definitions for BL-13, BL-65, and KS-65 have been located. These 

and potential definitions for CELL-8 and RT-210 are described in Attachment E, Technical Basis for 

Assigning Doses from Effective Derived Air Concentration Breathing Zone Results. The attachment 

also provides a method of calculating intakes from the EDAC data for dose reconstruction. A 

summary of the custom nuclides is shown in Table 5-38.

Dose reconstructors should apply the activity fraction for each isotope or element, then multiply the 

reported DAC-hr by the EDAC and the breathing rate (1.2 × 106 mL/hr) to get the total intake per year.

5.5.2.1.3 In Vivo Bioassay and BZ Data, 1995 to 2006

There might be chest count data and BZ air monitoring results for a thorium worker from 1995 to 

2006. If all chest count data are below the MDA, then only the BZ air monitoring results should be 

used to evaluate thorium dose. If thorium BZ air monitoring data are negative, but there are positive 

chest count results, then chest counts should be reviewed to determine if there were thorium intakes 

before 1995. An evaluation of the previous in vivo counts is performed to determine if the in vivo 

results reflect a lung burden that had been previously identified from earlier exposure. If a previous 

intake is discovered, then use the BZ data. For new intakes, use positive in vivo data instead of 

negative BZ data. 

5.5.2.2 Assignment of Unsupported Radium

For triple-separated thorium, there is no 228Ra intake associated with the thorium intake. However, in 

some unusual cases, there might have been an intake of 228Ra that was not associated with a thorium 

intake. This is called an intake of unsupported radium (and not part of the thorium mixture), and is

evaluated separately, as discussed below. When a thorium intake occurs, the activity ratio between 

228Ac and 212Pb can be up to 1:1. If the ratio exceeds 1.5:1, this is an indicator of an intake of 

unsupported radium and the 228Ra intake rate is estimated from the 228Ac activity.
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Table 5-38. Custom nuclides from EDAC BZ results. 

Custom 

nuclide EDAC (μCi/mL)

Isotope or element, 

activity fraction

BL-13 3.0E–12 Th-232, 0.0504;

Th-230, 0.248;

Th-228, 0.0565;

U-totalc, 0.645

BL-65 2.22E–12 Th-232, 0.20;

Ra-228, 0.20;

Ac-228, 0.20;

Th-228, 0.20;

Ra-224, 0.20

Cell 8a Reserved Reserved

KS-65 silo 1b 3.2E–11 K-65 silo 1, 

see Table 5-13

KS-65 silo 2b 2.0E–11 K-65 silo 2, 

see Table 5-13

RT-210 2.25E–10 Pb-210, 0.333;

Bi-210, 0.333;

Po-210, 0.333;

a. Contact Site Lead for guidance.

b. Dose reconstructors should use the mixture of either silo 1, 2 or 3 that is 

most favorable to the claimant, if unknown.

c. Because BL-13 primarily consists of uranium, it is not a replacement for 

thorium bioassay.

When there was an intake of unsupported radium, the following approach is taken.

Because 228Ac has a 6.15-hr half-life, it quickly (about 2 days) comes to equilibrium with its 228Ra 

parent. It can therefore be used to directly assess the 228Ra-228 intake. Determine the 228Ra intake 

by assuming a 228Ra chest count equal to the 228Ac measured value. Assume a type M material 

because this is not associated with a thorium intake.

5.5.2.3 Assignment of Unmonitored Thorium Dose

5.5.2.3.1 Thorium Coworker Model, 1979 to 1989

The derivation of the thorium intakes is described in Attachment F, Thorium In Vivo Coworker Study, 

1979 to 1989. As noted in the attachment, as well as discussed in Section 5.5.2.1.1, triple-separated 

thorium is assumed because it results in intakes that are favorable to the claimant. The coworker 

thorium intake rates are summarized in Table 5-39.

Table 5-39. Thorium coworker intake rates, 1979 to 1989. 

Intake solubility type

50th percentile

(pCi/d) GSD

95th percentile

(pCi/d)

232Th Type M 35.3 3.00 215

228Th Type M 6.71 3.00 40.85

232Th Type S 5.46 3.00 33.3

228Th Type S 1.04 3.00 6.33

Thorium coworker doses are assigned to all unmonitored workers, unless there is a reason why they 

should be excluded (e.g., a secretary or administrative individual who worked only in nonradiological 

areas). The 95th-percentile intake rate, with a constant distribution, is assigned to those with the 

highest potential for exposure. Workers with a baseline thorium fecal sample are included in this 

group, as well as subcontractors from IT Corporation working during 1988 and 1989. All others are 

assigned the 50th-percentile intake rate with a lognormal distribution. The same material type is 
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assigned for both isotopes of thorium (i.e., both are assigned as type M or both are assigned as 

type S).

An unsupported radium intake is not part of the coworker model and is not assigned.

5.5.2.3.2 Airborne Thorium Concentration, 1990 to 1994

The implementation of DOE 5480.11 (DOE 1988) requirements by January 1990 was readily apparent 

from the site documentation of the thorium overpacking work (access control, respiratory protection 

requirements, air monitoring with GA and BZ samples, and pre-job and follow-up fecal samples if 

required.) This provides confidence that no thorium workers were exposed to greater than 10% of a 

DAC (with respiratory protection being taken into account) on a sustained or average basis. Routine 

workplace air monitoring in nonairborne radioactivity areas was a site requirement, which provides a 

general assurance that workers were not chronically exposed to airborne levels greater than or 

equivalent to 10% of a DAC.

Because the thorium overpacking procedures did not appear to change much over time, it was 

considered useful to compare the thorium BZ results from the later periods to the 10% of a DAC 

assumption proposed for 1990 to 1994. The total annual thorium exposures in DAC-hr from 1993 to 

2006 in the HIS-20 database were fit to a lognormal distribution. The GM of the results without the 

respiratory protection factor did not exceed 17 DAC-hr/yr for any of the three data-weighting methods. 

This is much less than 0.1 DAC times 2,000 hr/yr (200 DAC-hr/yr), which would be the upper bound. 

With respiratory protection factors applied, the 95th percentile of the BZ results did not exceed 14 

DAC-hr/yr for any of the three data-weighting methods [5].

Therefore, the use of 10% of the thorium DAC air limit is appropriate. The thorium dose is based on 

the assumption of 100% 232Th as favorable to claimants. Airborne areas would have been controlled 

to the lower Class W DAC. It is a standard industry practice for the lowest DAC value to be used 

because, for a given air concentration, this results in a higher exposure in DAC-hours. An inhalation 

and ingestion intake of 232Th is assumed. Although the thorium exposure in 1990 could be limited to 

January to May 1990 (due to 5 months of Project 2 duration), a full year of exposure should be 

assumed as favorable to claimants. Attachment G, Class W Thorium-232 Inhalation Intake Rate 

Based on 10% of the Derived Air Concentration, provides the assumptions used to derive these intake 

rates. Ingestion doses are assigned as described in OCAS-TIB-009, Estimation of Ingestion Intakes

(NIOSH 2004), where the amount of activity ingested on a daily basis can be approximated assuming 

it to be 0.2 times the activity per cubic meter of air.

For 1990 through 1994, unless case-specific in vivo measurements are available, assume that each 

thorium worker was a full-time occupant of an area where the airborne thorium activity concentration 

was 10% of the DAC for Class W 232Th, which is 5 × 10-14 μCi/mL.

The derived inhalation and ingestion intake rates are shown in Table 5-40. Dose reconstructors 

should assign these intake rates as a constant distribution in IREP.

Table 5-40. 232Th inhalation and ingestion rates based on 

10% DAC. 

Years

Inhalation rate 

(dpm/yr)

Ingestion rate 

(dpm/yr)

1990–1994 266.4 5.55
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5.5.2.4 Assignment of Thoron Dose, 1954 to 2006

As previously discussed, thoron was present at Fernald. The primary dose from thoron is to the 

respiratory tract organs and is delivered primarily by the short-lived progeny (NIOSH 2013).

Thoron doses are assigned to all workers, unless there is a reason why they should be excluded (e.g., 

a cafeteria worker, accounting, legal staff individual, etc. who worked only in nonradiological areas). 

Thoron doses are summarized in Table 5-41. A constant distribution is assumed in IREP.

Table 5-41. Thoron worker exposure. 

Years Category WLM/yr

1954–1971 Operations and Handling 1.6

1972-1985 Long-term DOE repository storage 1.6

1986–1987 Passive storage No significant dose

1988–2006 Remediation (radiological workers)a 0.53

1988–2006 Remediation (non-radiological workers)b No significant dose

a. Radiological workers are those that could have worked in a process area and are 

assigned thoron dose.

b. Nonradiological workers are not assigned thoron exposures because they would 

not have received any significant dose.

Table 1 of DCAS-TIB-011, Lung Dose Conversion Factor for Thoron WLM (NIOSH 2013), provides 

dose conversion factors for converting WLMs to dose in rem for the lung, ET1, and ET2. An 

equilibrium of 80% 212Bi/212Pb is assumed.

5.5.3 Raffinate Dose Assessments

5.5.3.1 1952 to 1954

Assess radon breath measurements in this timeframe as follows:

1. Convert the radon breath analysis result to units of picocuries per liter.

2. Convert the value in step 1 to 226Ra whole-body content in pCi by multiplying by 2.52 × 105

pCi/(pCi/L).

3. Use the whole-body 226Ra content in IMBA to calculate a chronic intake of type M 226Ra 

beginning on January 1, 1952 (or date of employment, if later), through the date of the last 

radon breath analysis result.

4. In addition to the 226Ra intake estimate, apply the derived intake to the isotopes and ratios in 

Table 5-16.

If there are no radon breath analysis results for a worker, assign the 226Ra intake rate from Table 5-15. 

This intake rate is then used with the activity ratios in Table 5-16 to assign intakes of associated 

raffinate nuclides.

5.5.3.2 1954 to 1958 (Pitchblende Ores)

To account for the impurities in pitchblende, assume a 230Th intake equal to the calculated uranium 

(total uranium activity) intake. Assign the additional impurities by applying the ratios in Table 5-17. 

Dose reconstructions with overlapping periods in 1955 and 1956 require a comparison between 

Table 5-16 and Table 5-17 doses for silos 1 and 2, and the most favorable to the claimant selected for 

assigning dose.
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5.5.3.3 1959 to 1961 (Yellowcake)

To account for the handling or processing of yellowcake, assume a 230Th intake equal to the 

calculated uranium (total uranium activity) intake. Assign the additional impurities by applying the 

ratios in Table 5-17 for silo 3. Include the RU contaminants beginning in 1961.

5.5.4 Radon Exposure Assessment

These records typically appear in Personnel Exposure files, separate from the DOE Response file in 

the NIOSH-DCAS Claims Tracking System (NOCTS) and are radon exposure records in WLMs. 

Follow ORAUT-TKBS-0017-4 (ORAUT 2015b) guidance when evaluating these records.

5.5.5 Site Closure

In 2006, the site remediation was complete, and it was transferred to the DOE Office of Legacy 

Management. This ended the site bioassay and BZ measurements. Internal doses after that date 

should be assessed using the current version of ORAUT-TKBS-0017-4 (ORAUT 2015b).

5.6 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

[1] Potter, Eugene. ORAUT Team. Principal Health Physicist. November 2014.

An investigation was conducted into the possibility of limiting the workers to which 226Ra 

intakes would apply by external dose. NLO (1952) includes radon breath results for 1952. 

Assuming that workers doing the K-65 drum handling were subject to high external dose rates, 

the HIS-20 database was queried for 1952 external dose results. A comparison was made of 

the groups with 1952 radon breath results with the other individuals monitored for external 

dose in 1952. A comparison was made of the list of individuals in 1952 with breath data 

(65 total) to claimant files, and showed 10 workers with breath data. DOE files in NOCTS 

were examined; none included breath data. [redacted] of the individuals did have film badge 

investigations in the DOE file, which listed work with K-65 drums. The distributions were 

different with the breath sample group being higher, but the two distributions overlapped 

significantly. In addition, there were [redacted] workers with radon breath samples but no 

external data for 1952 in HIS-20. From this, it was concluded that an external dose trigger 

could not be used to assign 226Ra intakes.

[2] Rich, Bryce. ORAUT Team. Principal Health Physicist. March 2008.

A review conducted on gross beta urine sample results at Fernald that sometimes appear in 

claimant records indicated that they were not well enough defined to be of value.

[3] Brackett, Elizabeth. ORAUT Team. Principal Internal Dosimetrist. August 2013.

Use of the solubility class (absorption type) of BZ results from claimant records, DAC hours 

“with respirator” column, and then selection of the “max” solubility option for dose evaluation of 

BZ samples ensures doses that are favorable to the claimant.
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[4] Potter, Eugene. ORAUT Team. Principal Health Physicist. August 2013.

A review of HIS-20 indicated that there were “custom” air monitoring nuclides identified as 

BL-13, BL-65, CELL 8, KS-65, and RT-210, and these might be in a claimant’s records. A 

methodology for calculating an EDAC was developed.

[5] Potter, Eugene. ORAUT Team. Principal Health Physicist. January 2015.

A comparison and statistical analysis performed by Eugene Potter indicated that thorium BZ 

results from 1993 to 2006 in the HIS-20 database did not exceed the upper-bound limit of 

0.1 DAC times 2,000 hr/yr.

[6] Bryce, Rich L. ORAU Team. Principal Health Physicist.

The 233U and trace quantities of 232U came from thorium from Hanford from 1977 to 1979. The 

thorium was “recycled thorium” with a contamination level of approximately 130 g 233U per 10.8 

MTTh. The total 233U plant inventory spread through the 351 MTTh was approximately 4.2 kg. 

Trace levels of 232U could have been present.

[7] Bryce, Rich L. ORAU Team. Principal Health Physicist.

The quantities stored in the process buildings of 15 MTTh awaiting immediate processing are 

not known from records or interviews. It is reasonable (based upon a knowledge of 

processing practices) to move materials to a ready point and the amount assumed is 

considered adequately conservative. The primary storage locations were outside on pads or 

in adjoining storage facilities.

[8] Bryce, Rich L. ORAU Team. Principal Health Physicist. 

Thorium feed materials were received from a variety of separations facilities stockpiles. A 

1-year delay since separation was judged to be a reasonable and conservative average.

[9] Arno, Matthew. ORAU Team. Dose Reconstructor. June 18, 2007.

Lognormal distributions typically provide the best fit to the available data and are a distribution 

suitable for input into IREP.

[10] Arno, Matthew. ORAU Team. Dose Reconstructor. June 18, 2007.

The error in individual bioassay results has a normal distribution because the dominant source 

of uncertainty is the counting statistics. Although the underlying group statistics have a 

lognormal distribution, each result was treated as if it had a normal distribution to match what 

was done for analysis of an individual’s bioassay data and because the lognormal distribution 

of the data is addressed by analyzing both the 50th and 84th percentiles of the data.

[11] LaBone, Thomas. ORAU Team. Deputy PID. December 18, 2007.

The assumption of equilibrium between 212Pb and 228

Th requires an assumption that none of 

the noble gas 220Rn escapes the lung. Radon-220 has a half-life of only 55.6 s, which makes 

this a reasonable assumption.

[12] Morris, Robert. ORAU Team. Principal Health Physicist. November 14, 2007.

Inspection of in vivo monitoring records confirms that many of the workers counted during the 

first visit had “thorium worker” or “former thorium worker” noted on their individual MIVRML log 

sheets.

[13] LaBone, Thomas. ORAU Team. Deputy PID. May 19, 2014.

Bias factors were developed based on the Fernald post-SEC thorium methodology.
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[14] Arno, Matthew. ORAU Team. Dose Reconstructor. June 18, 2007.

Lognormal distributions typically provide the best fit to the available data and are a distribution 

suitable for input into IREP.

[15] Morris, Robert. ORAU Team. Principal Health Physicist. November 14, 2007.

The Fernald SEC petition includes 1989. The thorium operations in 1989 were unchanged 

from those in 1988 and it is judged reasonable and efficient to extend the 1980 through 1988 

period to include 1989.
Document No. ORAUT-TKBS-0017-5 Revision No. 02 Effective Date: 09/30/2016 Page 58 of 189

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GLOSSARY

absorption type

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively insoluble in the 

respiratory tract (slow solubilization). Also called solubility type.

acute exposure

Radiation exposure to the body delivered in a short period. See chronic exposure.

becquerel (Bq)

A unit of radioactivity equal to one disintegration per second.

chronic exposure

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years). 

See acute exposure.

class

See inhalation class and absorption type.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

decision level (DL)

Minimum level at which a particular device can detect and quantify exposure or radiation. Also 

called lower limit of detection and detection limit or level.

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of rad, rep, or grays.

dose equivalent

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dosimetry

Measurement and calculation of internal and external radiation doses.

enriched uranium (EU)

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7% by mass. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade 

uranium contains greater than 90% 235U.
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exposure

(1) In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.

inhalation class

Former respiratory tract inhalation classification scheme developed by the International 

Council on Radiological Protection for inhaled material according to its rate of clearance from 

the pulmonary region of the lung. Materials were classified as D (days, half-life less than 10 

days), W (weeks, 10 to 100 days), or Y (years, more than 100 days). See absorption type, 

which superseded this concept.

in vitro bioassay

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo bioassay

Measurements of radioactive material in the human body utilizing instrumentation that detects 

radiation emitted from the radioactive material in the body.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.

maximum permissible lung burden (MPLB)

Historical occupational limit for radionuclides in the lung defined as the quantity of the 

radionuclide that could be present at any given time to deliver 15 rem per year to the lung at 

the end of a 50-year period of chronic exposure.

microcurie

A measure of radioactivity equal to one-millionth of a curie. See curie.

millirem

A unit of radiation dose equal to one-thousandth of a rem. See rem.

minimum detectable activity (MDA)

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error).

natural uranium (NU)

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcuries 

per gram).

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.
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potential alpha energy concentration (PAEC)

Kinetic energy in units of working levels potentially released in a unit volume of air by alpha 

particles emitted by the short-lived radioactive progeny of 222Rn (218Po, 214Pb, 214Bi, and 214Po) 

and 220Rn (216Po, 212Pb, 212Bi, 212Po). See potential alpha energy exposure and working level.

potential alpha energy exposure (PAEE)

Average potential alpha energy concentration to which a worker is exposed multiplied by the 

time of exposure in working months of 170 hours (units of working level months). PAEE is the 

potential alpha energy concentration multiplied by time. See potential alpha energy 

concentration and working level month.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer.

radioactive

Of, caused by, or exhibiting radioactivity.

radionuclide

Radioactive nuclide. See radioactive and nuclide.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

solubility type

See absorption type.

transuranic (TRU) elements

Elements with atomic numbers above 92 (uranium). Examples include plutonium and 

americium.

working level (WL)

Unit of concentration in air of the short-lived decay products of 222Rn (218Po, 214Pb, 214Bi, and 

214Po) and 220Rn (216Po, 212Pb, 212Bi, 212Po) defined as any combination of the short-lived 

radioactive progeny of radon or thoron in 1 liter of air, without regard to the degree of 

equilibrium, that results in the ultimate emission of 130,000 megaelectron-volts of alpha 

energy; 1 WL equals 2.083 × 10-5 joules per cubic meter. See potential alpha energy 

concentration.

working level month (WLM)

Unit of exposure to radon progeny defined as exposure for 1 working month (170 working 

hours) to a potential alpha energy concentration of 1 WL; 1 WLM equals 1 WL times 
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170 hours, which is 0.00354 joule-hours per cubic meter. See potential alpha energy 

exposure and working level.
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ATTACHMENT A 

RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM

TABLE OF CONTENTS

SECTION TITLE PAGE

A.2 The Issue............................................................................................................................... 76

A.3 General Background .............................................................................................................. 77

A.4 Timeline ................................................................................................................................. 77

A.5 Discussion Points................................................................................................................... 78

A.6 Conclusion ............................................................................................................................. 81

A.7 Discussion of References....................................................................................................... 81

LIST OF TABLES 

TABLE TITLE PAGE

A-1 Timeline of PGDP shipments to Fernald ................................................................................ 78

A-2 Totals of recycled feeds to Fernald ........................................................................................ 78
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ATTACHMENT A 

RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM (continued)

A.1 SUMMARY

The concern that the recommended sitewide default of 400 ppb U Pu2 for the 1970s and 1980s might 

not bound the dose from RU contaminants for all categories of workers because of the higher values 

in process subgroup 10A is addressed in this attachment. The 10A process categorization represents 

“tower ash and decontamination residues from Portsmouth and Paducah GDPs” with the high values 

in the activity distribution coming from PGDP in a single shipment in 1980.

In the 1970s (until 1977) the ash and residues from PGDP had elevated contaminant concentrations, 

but they were generally under 10 ppb U Pu and were sampled and blended directly with uranium 

concentrates to attain operational concentrations equivalent to those processed directly from the 

primary sites. (See the Table 5-6 values of process subgroup 6A at 4.6 ppb U Pu at the 

95th percentile.) During this period there were approximately 10 g of plutonium received, which 

generally increased the levels in the Fernald process streams, but the latter were still held to levels 

below 10 ppb U Pu. The concentrates ran out in 1977, requiring blending with the lowest process 

streams available to stay under the 10-ppb U level. In June of 1980 a single shipment of 

approximately 24 MTU PGDP process plant ash was shipped in 16 T-hoppers and contained 25.3 g of 

plutonium. This essentially doubled the plutonium inventory for the entire operational history of the 

plant to that point.

This shipment was sampled and stored primarily to develop blending material inventories and 

processes to accommodate this unusual shipment. After approximately a 2-year period, there is a 

1982 record of repackaging five of the T-hoppers with the highest contaminant levels into large 

containers in Plant 4 to facilitate process needs in Plants 8 and 2/3. Air sampling records associated 

with this repackaging are recorded as well as the time required, which was done primarily on the first 

shift of several days. It is assumed that this record was generated to document an unusual operation 

with specific contamination control problems. Although these were five of the 16 T-hoppers with the 

greatest concerns and difficulties, the times of repackaging were used to estimate the repackaging of 

the hoppers that contained contaminants above the 400 ppb U Pu plantwide default. It is important to 

note that of the 16 T-hopper containers in the 1980 shipment, two contained 66% of the total 

plutonium contaminants, while six contained 88% of the plutonium itself. It is also of note that only 

eight of the hoppers contained plutonium contaminant concentration above 400 ppb U.

Based on the recorded times of the five hoppers with the greatest problems, 150 hours was 

conservatively estimated for repackaging those eight hoppers with concentrations above 

400 ppb U Pu. This would be the handling time in April and May of 1982. There are recorded 

descriptions of further processing, the next step (after another delay) being blending with sump cake 

in Plant 8 of appropriate content to create a calcium uranate feed for further dissolution and process in 

Plant 2/3, these being semi-remote operations. The final recorded processing resulted in a formal 

approval to process the Plant 2/3 UO3 product, which was in the range of 20 to 42 ppb U Pu.

A.2 THE ISSUE

The remaining issue related to acceptance of 400 ppb U Pu as a sitewide default is the question 

relating to the NIOSH decision not to use the Fernald process stream 10A (95th percentile of 

1,732 ppb U Pu), as described in the Ohio Mass Balance Report (DOE 2000a) in establishing the 

 

2 For the sake of simplicity, this attachment discusses only plutonium. The same principles apply to the other 

primary contaminants (neptunium and technetium), which are also listed with appropriate defaults in the 

guidance to the dose reconstructors.
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ATTACHMENT A 

RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM (continued)

values of the defaults. This process stream represents the wastes received from the GDPs (but 

primarily and specifically PGDP) starting in about 1975 or 1976, but realistically starting in 1980. 

Because the contaminant distribution within the materials of this process stream were handled 

episodically (as opposed to continuously) and of short duration in nature and blended with low-level 

feed materials to attain approved limits, the materials in this process stream were not used to 

establish the defaults.

The primary remaining question relates to the length of time that can reasonably be established as the 

maximum exposure time for any given individual worker to the contaminant ratios in the 10A process 

stream. Though the materials with the high levels of contaminants above the 400 ppb U Pu default 

were handled for a short period of time in 1982, were recognized as materials above permissible 

levels, and preventative measures were used to protect the workers from exposure, the presence of 

the materials of this process stream is a concern to FMPC Working Group reviewers as a potential 

exposure above those established.

A.3 GENERAL BACKGROUND

Quoting from Attachment E.2, “Selected Non-Routine Campaigns,” of the Ohio Mass Balance Report 

(DOE 2000a, p. 900):

Over a period of several years, a large inventory of uranium-bearing residues was 

accumulated at the PGDP. Most of the residues were near-normal in isotopic assay and 

were highly diverse in physical and chemical characteristics. Although much of the 

residue inventory was generated by Paducah site operations, a significant quantity of the 

UF4-type residues was generated at the FEMP and was shipped to Paducah for possible 

use as feed material in their operations. Very little of the UF4 shipped from the FEMP 

was used at Paducah because their operation was suspended shortly after the transfer 

was completed.

By early 1975, the Paducah residue inventory contained approx. 400 MTU. After 

assessing the operational and environmental impacts of processing this inventory at 

Fernald, DOE-ORO directed that the inventory be shipped to the FEMP for recovering 

the uranium in conjunction with the ore concentrate campaign that was in progress in 

Plant 2/3. The introduction of this inventory as a blend with ore concentrates began in 

1976. Paducah continued shipping various types of residues after 1976. These 

materials were placed into FEMP’s inventory of recoverable residues for eventual 

enriched uranium process campaigns in Plant 2/3.

A.4 TIMELINE

The brief timeline and mass summary Tables A-1 and A-2 indicate two important points relevant to 

this discussion:

1. The concentration of contaminants relative to uranium was generally below 10 ppb U Pu 

before the single shipment of 16 T-hoppers of feed plant waste in 1980.

2. The total mass of 22.5 MTU in this shipment was only 0.3% of the total MTU received during 

this period.
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ATTACHMENT A 

RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM (continued)

Table A-1. Timeline of PGDP shipments to Fernald.

Date Discussion References

1966–1974 UO3, U3O8 incinerator ash materials – Pu at 6 ppb 

U Pu

Spenceley 1985; DOE 2000a, 

p. 889

1975–1976 291 MTU scrap and fluoride residues – 2.6 g Pu at 

9 ppb U

Start of major transfer of PGDP scraps and waste 

materials to Fernald

Spenceley 1985; Spenceley 1985; 

NLO 1957–1984; Author unknown 

undated b

1979, 1981–1983 1,814 MTU of UO3, UNH –

10 g Pu at 5.6 ppb U Pu

Spenceley 1985

1980 22.5 MT Feed Plant ash –

25.3 g Pu at 1,124 ppb U Pu. This shipment 

consisted of 16 T-hopper containers, 2 of which 

contained 66% of the Pu. The highest container 

contained 39% of the Pu.

Spenceley 1985; DOE 2000a; 

NLO 1957–1984; Author unknown 

undated b; Robinson 1989; Author 

unknown undated a; WMCO 

1986c; Walker 1986

Table A-2. Totals of recycled feeds to Fernald (Spenceley 1985).

Materials MTU/% Pu-239 g Pu-239 ppb U

UO3 6,112/82% 17.4 2.9

UNH 699/9% 4.7 6.7

Incinerator/tower ash 311/4% 0.2 0.7

UO2 39/0.5% 0.01 Not significant

Feed plant ash 22.5/0.3% 25.3 1,124

1975–1976 miscellaneous scrap 291/4% 2.6 9

A.5 DISCUSSION POINTS

1. PGDP wastes had accumulated after shutdown of processing facilities in the early 1970s. The 

approximately 400 MTU were directed to Fernald for recovery after considering the practicality, 

radiological impact, and ability to blend the materials to acceptable levels for processing. 

During the period from the early 1970s through 1979 the concentrations from PGDP were 

generally below 10 ppb, though higher than the feed materials from the primary RU source 

facilities (Hanford and SRS specifically) by perhaps a factor of 2 or greater (DOE 2000a, 

p. 1,099; see process 6A values in Attachment F.1; 95th percentile value is 4.6 ppb U Pu). 

This material was easily blended directly with ore concentrates to bring the contaminant 

concentrations relative to uranium into the levels that were being processed straight from the 

PUREX processes at Hanford and SRS. However the supply of concentrates ran out in about 

1977, which made it necessary to blend with lower levels of feed materials (but not NU).

2. However, this did introduce an added total quantity of plutonium to Fernald from the 

“secondary” sources (GDPs) that had not been through the decontaminating mechanisms of 

the diffusion cascades. That is, the materials that were sent during this time were scraps and 

residues and were in ranged from 6 to 10 ppb Pu rather than the parts per ton range.

3. The 1980 shipment of primarily feed plant ash was unusually high in contaminant 

concentrations, adding over 25 g of plutonium to the plant inventory (essentially doubling the 

inventory for the entire plant history), and presenting an increased challenge to effectively 

blend the materials to acceptable levels (Spenceley 1985, pp. 7–17, 19, 20, 34; Mengel 

1982, p. 4).
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ATTACHMENT A 

RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM (continued)

4. This shipment was not processed immediately: “These materials were placed into FEMP’s 

inventory of recoverable residues for eventual EU process campaigns in Plant 2/3.” The 

shipments were received at Plant 1 for immediate sampling and then storage until needed.

The following from Robinson (1989, p. 50) is instructive:

D.C. Bonfer answer – Paducah Feed Plant Ash was prepackaged from hoppers into 

drums in Plant 4. Drummed material was later blended with sump cake in the rotary kiln 

operation in Plant 8 producing seven lots of calcium uranate, which was used as feed 

for the refinery. Most of this material was converted to UO3; a quantity remains as UNH 

in the refinery. UO3, which was contaminated with the plutonium, remains on inventory 

at NLO. The remainder of the UO3 was converted to UF4 and blended with three parts 

of UF4 produced from Hanford UO3 in the metal reduction operations.

5. Interviews with four senior former managers (ORAUT 2008b, 2011b, 2011c, 2011d) during the 

period in question were productive in providing operational insights and perceptions, but there 

was not a direct recollection of the specific time required for the processing the 16 T-hoppers 

to the point that the contaminants would be below the default level of 400 ppb U Pu.

6. The T-hoppers were stored for approximately 2 years (until 1982) when five of the highest 

level T-hoppers were taken to Plant 4 for repackaging into containers compatible with the 

other plant facilities (Mengel 1982). Repackaging took longer than expected due to some of 

the material having become solid in at least some of the containers. Instead of processing at a 

rate of two to three hoppers per shift (ORAUT 2011d), two of the containers required two shifts 

each (with an estimated 5 hours per shift due to airline respiratory protection and associated 

protective clothing requirements, for a total of 10 hours; Hinnefeld 1988), one container 

required three shifts (15 hours), one required five shifts (25 hours), and the remaining 

container required eight shifts (40 hours). This time requirement was complicated and the

activities delayed by the need to correct procedures that were spreading contamination 

(Mengel 1982, p. 4, Table I). The first shift on April 28, 1982, was the only time removing 

material from the hopper to a tray on the floor was indicated.

7. This repackaging effort took place mainly on the first shift, presumably to minimize routine 

operations impact. Conclusions from these records place an upper bound on the time of 

potential exposure to the highest contamination in process 10A materials in the repackaging 

operation. All of the POOS from PGDP came in 16 T-hoppers. (39% of the activity was in the 

highest container, 67% in the highest two containers, 88% in the highest seven containers). 

The hopper with the highest contamination levels required five shifts or approximately 25 

hours on five different shifts.

8. The Bonfer quotation in item 4 above indicates that the material in the 1980 shipment was 

taken to Plant 8 “sometime later” after repackaging and blended with “sump cake” in the rotary 

kiln. From the same quotation it is apparent that the blended calcium uranate (no longer 

process 10A contamination levels; that is, below the 400 ppb U Pu default level) was dissolved 

in Plant 2/3 resulting in liquid UNH. Again the materials were stored until appropriate 

processing to UO3 was available (but again the materials had lower levels of contaminants 

below the recommended 400 ppb U default).

9. The next set of records are from about 1985 to 1987 when the material had been dissolved 

and run through Plant 2/3 extraction with UO3 product that had plutonium levels of 26 to 
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ATTACHMENT A 

RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM (continued)

42 ppb (Jelinek and Thomas 1985). In addition, a training program specifically for the POOS 

processing in Plants 4 and 5 discussed the processing of 168 MTU (UO3), including blending 

of the 42 ppb U Pu UO3 on a “one-to-one” basis with “in spec. materials” (Author unknown 

undated a). The training course also discussed of the fact that FMPC had raised the “target” 

limit to 20 ppb, but DOE headquarters had reduced the limit to 10 ppb again. However, the 

training was for the procedure of processing the 42 ppb U. This operation is not of interest in 

answering the question of whether the 400 ppb U is adequately bounding.

10. Conclusions about time of potential exposure to the materials after repackaging are based on 

the following discussion. After the repackaging in Plant 4, the material was again stored for a 

convenient time and circumstances for further processing and blending in Plant 8. It is 

possible that one or a few of the drums of material that were not free flowing could have been 

milled in Plant 1 before sending to Plant 8. This possibility is supported in DOE (2000a, 

p. 556), which indicates that one of the data points in the 10A process stream is from a Plant 1 

dust filter. Both of these operations involved handling the large drums on automatic lifts and in 

enclosed systems. Time of exposure would be less than the repackaging times. The Plant 8 

blended material was no longer above the 400 ppb U in question.

Therefore, the exposure times to the highest contaminant levels would be 20 shifts multiplied 

by 5 hours per shift or 100 hours for any single worker that was present for each of the 

20 shifts required for repackaging five hoppers, or 150 hours to process the eight hoppers that 

exceeded the 400 ppb U. This assumes that each of the eight hoppers above 400 ppb 

required the same time as the problematic five recorded. This is extremely conservative 

due to:

• Only 3 of the hoppers were above 1,000 ppb, and only 8 of the 16 were above 

400 ppb.

• The time of continuous exposure is significantly overestimated and particularly due to 

the use of airline respiratory protection and the equivalent protective clothing.

• The other 11 hoppers probably required much less time, as indicated in the recording 

of the five highest and more problematic.

This repackaging time would have been spent in 1982, assuming that the same worker was 

present during the repackaging of each of the 16 hoppers. Therefore, only 8% of the annual 

operating time in that year could have been spent in contact with this repackaging operation, 

and then only 50% of the containers had contaminants above 400 ppb U. After the 

repackaging the workers would be assigned to work with materials that were probably an order 

of magnitude or 2 lower than the 400 ppb U Pu.

11. The mass of the 1980 shipment, although containing 25.3 g of plutonium, was only 22.5 MTU. 

As time went on and after 4 to 5 years, much of this contamination was spread through the 

Fernald processes, raising many of them above previous levels. However, handling of the 

material did not require a consistent effort, as has been discussed, because it was in storage 

for most of this period. The mass did not account for more than a fraction of a percent of the 

material in process.
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ATTACHMENT A 

RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM (continued)

A.6 CONCLUSION

It is the conclusion of this investigation that the levels of 1,732 ppb U Pu (at the 95th percentile for 

process group 10A) is adequately covered by the 400 ppb U Pu plantwide default. The 400 ppb Pu 

default was calculated by conservatively considering the maximum levels in any of those process 

groups that represent continuous operations. The highest levels in the ash and residues from PGDP 

were contained in less than 0.3% of the wastes from PGDP and required a maximum of a few 

hundred hours of handling before downblending. This represents a short exposure time with 

documented safety airline respiratory protection. The remaining exposure time in 1982 for any worker 

would have been with materials below the recommended defaults. There is no reason to believe that 

any worker could have an exposure for which the recommended default would not be adequately 

conservative.

A.7 DISCUSSION OF REFERENCES

Ohio Mass Balance Report (DOE 2000a)

Transfer of PGDP wastes began in 1976 and continued. “These materials were placed into FEMP’s 

inventory of recoverable residues for eventual EU process campaigns in Plant 2/3.” Some of the 

wastes were blended and some were not (p. 900). See p. 901 for blending ratios, feed evaluations, 

and classification of scrap residues. Table E.2-1 is a listing of PGDP scrap classification, amounts, 

descriptions, pretreatments, etc. Table E.2-2 gives the Fernald permissible levels. The wastes were 

blended with ore concentrates until 1977, when the supply ran out. Between 1978 and 1986 

77.5 MTU of residues were recovered in Plant 2/3 (Table E.2-4). The scraps from K-25 and 

Portsmouth GDP are discussed on p. 908.

“Joint Task Force on Recycle Material Processing” (Spenceley 1985)

See pp. 4, 5, 7 to 20, and 34 particularly for a comparative, time-based summary of the PGDP wastes 

as received. 1980 was the high point. Table 1 on p. 34 gives the listing of the 16 T-hoppers with the 

25 g of plutonium that came in 1980.

“Air Sampling at Plant 4 Repackaging of Paducah Reprocessed Feed Plant Ash” (Mengel 1982)

This is a four-page summary of the air sample levels during repackaging the wastes from five 

T-hoppers into large drums during 1982 (2 years after receipt). The effort took about 20 days on the 

day shift for a total of around 100 hours. The implication is that the effort was interrupted for other 

routine processes. A “mistake” of “breaking up” materials outside the containers on a single shift on 

April 28, 1982, was corrected. Personnel wore airline masks, etc. This is a clear indication of the fact 

that the processing was not done all at once after 1980, but occurred as they identified materials 

satisfactory for blending.

Evaluation of Plutonium Bioassay Data (Robinson 1989)

A variety of files dealing with:

• Evaluation of bioassay analytical data, p. 2, including detection limits, p. 9;

• Responses to 1985 Task Force Question set, p. 20, see Table 1, p. 25, and Bonfer response 

on p. 50, note Table I, p. 51;

• POOS blending processes, see pp. 61, 63, and 86;

• Precautions during blending, see p. 68; authorization to process >20 ppb U Pu in 1985;
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ATTACHMENT A 

RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM (continued)

• Summaries of plutonium levels by year;

• Neton recommendations, p. 91; and

• Others.

“POOS History and Risk Assessment” (Hinnefeld 1988) 

See report starting on page 121 for another analytical description of the repackaging operation in 

Plant 4, including Table 1 on page 123. Again, this is documentation of the results of this particular 

operation of repackaging 5 of the 16 T-hoppers.

DOE Contact Report, Processing 168 MTU of UO3 containing Recycle Contaminants (Jelinek 

and Thomas 1985)

This is documentation in 1985 (3 years after the repackaging of the five T hoppers; at that point it was 

through the Plant 2/3 and at 26 to 34 ppb U Pu. There is a hazards analysis with a call for bioassay, 

etc.

Interview with Former FEMP employees (ORAUT 2008b)

A description of blending PGDP ash (p. 21). It was sampled and milled in Plant 1 then blended in 

Plant 4. Statement made that Plant 1 equipment was not “large scale” for blending.

“Discussion of and Questions Relating to the Specific Processes and Time Required to Down 

Blend the 10A Sub Group Process” (ORAUT 2011b)

A former Fernald employee suggested that his recollection was that the POOS materials were 

dissolved in nitric acid in the Plant 2/3 front end and stored as UNH until suitable blending materials 

could be used to meet processing limitations.

Former Fernald Employee, September 9, 2011 (ORAUT 2011c)

A worker indicated that his recollections were not exact, but that the nature of the materials 

(nonflowing, etc.) would require milling in Plant 1 and some processing in Plant 8 before final 

processing in Plant 2/3.

Interview with Former Fernald Employee, October 4, 2011 (ORAUT 2011d)

To his recollection, the repackaging and blending operation was conducted in Plant 4 because this 

facility had the equipment and ability to handle the large containers. The containers were emptied 

into a receiver on the fourth floor, which fed in an enclosed transfer facility to the receivers on the first 

floor. All of the equipment was enclosed and vented to filters. The only contamination occurred 

during unusual leaks or spills. He indicated that while he did not recall specifically, he would estimate 

that under normal conditions two or three hoppers could be processed per shift.

Processing Uranium Containing Plutonium Above Target Limit (Author unknown undated a)

Specifically mentions training for the 168 MTU (UO3) at 42 ppb U and blending 67 MT with 67 MT of 

“in spec” material in Plant 4 (p. 5). Also mentions 75,000 gal of POOS raffinate to be processed. 

Page 6 discussion of FMPC 20 ppb U target limit, then DOE set target limit at 10 ppb U, with the 

materials to process at FMPC at 42 ppb U.

Collection of Reports and Data Including Discussion of Early RU Programs (Author unknown 

undated b)

Statement of blending 1976 PGDP wastes with the last of the ore concentrates on p. 128. Page 139 

shows a 1999 spreadsheet listing of analytical results of T-hopper T-449 container (which had the 
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ATTACHMENT A 

RATIONALE FOR 400 PPB U PU FOR 10A PROCESS STREAM (continued)

highest activity levels) with an average of approximately 7,000 ppb U (multiple sampling of each 

container, inhomogeneous results, two containers above 7,000 ppb U, other two below 1,000). A 

table of document references for enriched RU programs at FMPC is listed on pp. 193–198. Page 365 

shows a report of 99Tc at FEMP.

Type C Investigation: Discovery of Out-Of-Specification Plutonium Bearing Material During 

Inspection and Handling of Plant 1 Pad Residues for Waste Shipment (WMCO 1988)

Drums of 1977 PGDP wastes that had been mislabeled and set aside. Levels were 17.6 ppb U.

POOS Handling and Urinalyses Hazards Analyses (Traub 1985)

1985 analysis by Pacific Northwest National Laboratory about change in maximum permissible 

concentrations based on contaminants.

Plant 4 Smear Results from October to November, 1986 (WMCO 1986c)

Pre- and post-POOS processing sample results.

“POOS Raffinate Spill” (Walker 1986)

This reference contains a technical discussion of 40 ppb U versus 10 ppb U.
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES

TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Introduction ............................................................................................................................ 86

B.2 Thorium Process Inventories and Thoron Potential................................................................ 88

B.3 Thorium Waste Repository Locations and Quantities ............................................................. 89

B.4 Factors to Consider in Determining Potential Alpha Energy Concentration ............................ 90

B.5 NIOSH Responses to SC&A Comments on Methods for Reconstructing Thoron Doses 

at Fernald Presented at the 12/4/14 Working Group Meeting................................................. 96

B.5.1 Background ................................................................................................................ 96

B.5.2 Discussion .................................................................................................................. 97

B.5.3 Thoron Measurements................................................................................................ 99

B.5.4 Thoron Daughter WL Measurements........................................................................ 101

B.5.5 Conclusion................................................................................................................ 105

B.5.6 1954 Thoron/Thoron Daughter Results..................................................................... 109

LIST OF TABLES 

TABLE TITLE PAGE

B-1 Historical thorium mass balance............................................................................................. 86

B-2 Thorium storage inventory, 1987............................................................................................ 90

B-3 PAEC for thoron and progeny ................................................................................................ 90

B-4 Building dimensions and ventilation systems.......................................................................... 91

B-5 Summary of thoron exposure estimates................................................................................. 96

B-6 Thoron exposure....................................................................................................................97

B-7 Summary of the 1996 through 1998 thoron measurement data............................................ 100

B-8 Summary of WL measurements........................................................................................... 103

B-9 Three high measurements near Plant 8 silo and bins in 1986 .............................................. 104

B-10 Locations and numbers of samples...................................................................................... 109

B-11 Thoron and thoron daughter results ..................................................................................... 110

LIST OF FIGURES 

FIGURES TITLE PAGE

B-1 232Th decay diagram, including the 220Rn (thoron) decay chain .............................................. 87

B-2 Decay and buildup of 228Th and progeny after processing...................................................... 88

B-3 Thoron comparison, 1996 to 1998........................................................................................ 106

B-4 Measurements and lognormal fits, 1996............................................................................... 107

B-5 WL comparisons, 1986 through 2003................................................................................... 108

B-6 GA center of rod storage area.............................................................................................. 111

B-7 GA north end of rod storage area......................................................................................... 112

B-8 GA TNT and sump cake storage area.................................................................................. 112
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ATTACHMENT B

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

B-9 GA wet area, main floor ....................................................................................................... 113

B-10 GA Wet area, top deck......................................................................................................... 113
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ATTACHMENT B

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

B.1 INTRODUCTION

From the beginning of FMPC operations in late 1953 until the “completion of the disposition of the 

containerized waste inventory” (DOE 2007), there have been uranium and thorium processes 

(summarized in Table B-1) with feed materials and the associated waste materials that have resulted 

in elevated levels of 3.823-day half-life 222Rn (radon) and 56-second half-life 220Rn (thoron) along with 

their particulate progeny. Radon is the first progeny of 226Ra, which was present in significant 

quantities in the raffinates from processing high uranium-bearing pitchblende ores. The elevated 

radon exposure levels at Fernald from this source are addressed in ORAUT-TKBS-0017-4 (ORAUT 

2016b). This attachment addresses the potential exposures from elevated levels of thoron during 

processing and storage of thorium at Fernald during the post-SEC periods.

Table B-1. Historical thorium mass balance (MTTh).a,b

Primary 

buildings Period Daily Annual Total Operations and materials

Plant 9 1954–1956 0.5–0.6 100–200 380 ThF4 to metal and machining of cores 

for reactors. Wastes stored in Plants 1 

and 9 storage areas.

Plant 4 1954–1956 0.5–0.6 150–300 461 Thorium oxide conversion to ThF4 for 

use in Plant 9.

Plant 6 1959–1963 0.02 10–20 80 Converted furnace for Plant 9 1954 to 

1956 Th recovery from scraps and 

waste.

Plant 1 1954–1979 <1 Unknown Unknown Receive, weigh, sample, and store 

source and special materials.

Pilot Plant 1954–1956 

and

1964–1979

0.01–0.75 8–381 2558 68% of FMPC Th processed in Pilot 

Plant. Metal production, dense thoria, 

thorium nitrate to thoria gel, thorium 

hydroxide production, processes 

developed for scale up to production 

levels.

Plant 8 1965 

and 

1968–1972

0.025 92 303 ThF4 to thorium hydroxide, thorium 

sump cake calcined, thorium residue 

processing to thorium hydroxide

Plant 2/3 1968 0.12 45 45 Short-term denitration experiment and 

uranium recovery from thorium 

hydroxide from plant 8.

Total 

processed

Not 

applicable

Not

applicable

Not 

applicable

3,827c Not applicable

FMPC 

waste pits

1954–2006 

(closure)

Not 

applicable

Not 

applicable

65 Discarded type Super S thorium 

material as waste

a. Sources: Dolan and Hill (1988); DOE (1987b); Mead, Savage, and Fugate (ca. 1986); Clark et al. (1989); Grumski 

(1987); Bogar (1987); Starkey (1964).

b. For perspective, a metric ton of pure thorium metal at the theoretical density occupies a relatively small volume. For 

example a 10- to 11-gal can (about 15 in. diameter and 15-in. tall) would weigh approximately 1,000 lb or 0.45 MT. 

Three of these relatively small cans could hold in excess of a metric ton of thorium.

c. Not including the unknown amount from Plant 1.

As illustrated in the decay diagram of 232Th (Figure B-1), thoron is the second progeny of 228Th and, in 

a couple of weeks after processing for thorium purification, can be considered to be in full equilibrium 

with the parent 228Th. Thorium-228 is a third progeny of the long-lived 232Th and generally in 40% to 

65% equilibrium for materials processed at FMPC (Figure B-2). The degree of equilibrium is 

dependent on both (1) the decay of 228Th without replenishment from the 5.7-year half-life 228Ra after 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

removal of the thorium progeny and (2) the time it takes the 228Ra to build into equilibrium with 232Th 

(Figure B-2). For the post-SEC period (after 1979), 65% equilibrium is assumed as the default. 

Thorium-228 is also a first progeny of 232U, which could have been present in small quantities from 

recycled thorium materials from Hanford in the 1977 to 1979 process, but not in sufficient levels from 

this source to result in a significant change in the defaults defined in this attachment [6].

Figure B-1. 232Th decay diagram, including the 220Rn (thoron) decay chain.

Thoron was present and a portion released during the processing and storage of metric tons of a 

variety of forms of thorium. Although this radioactive gas and its short-lived progeny are present in 

variable levels as a natural background, elevated levels with those of radon have been identified and 

studied and associated with increased respiratory cancer potential (i.e., uranium miner studies and 

others listed).

The primary focus of Fernald radiological safety programs for thorium was to define the thorium air 

concentrations in air in the work place coupled with in vivo lung counts for thorium, the results of 

which provided the means of controlling worker exposures to levels below the then-current 

permissible levels (Tomes 2001). In addition the metric ton quantities of thorium in process (with the 

measured particulate releases), or in storage, provided sources of elevated thoron gas release with its 

subsequent progeny. The records indicate that there was an awareness of the potential for exposure 

to thoron and its progeny through documentation of reported “thoron” measurements and holding of 

the routine BZ and GA samples for up to 96 hours to allow the short-lived radon and thoron progeny 

to decay (Weaver 1987; Starkey 1964). This provided a more meaningful measure of the long-lived 

0.64

0.36
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

thorium concentrations (Starkey 1964). Because air activity sample analyses specifically for thoron 

progeny are insufficient to directly document the routine concentrations in the work places at all times, 

an approach for bounding the possible workplace concentrations of thoron and progeny is presented. 

0 5 10 15 20 25

Time since chemical purification, years

0.1

1

10

100

Activity of Isotopes, % of Equilibrium Value

Th-232

Ra-228 or Ac-228 ingrowth

Original Th-228 decay

Th-228 ingrowth from Ac-228

Total Th-228 or Pb-212

Activity of Natural Thorium Following Removal of Impurities

Minimumum Value of Th-228 is 42% at 4.5 Years

Figure B-2. Decay and buildup of 228Th and progeny after processing.

This attachment presents the methods and considerations of bounding the work place exposures to 

thoron and primarily the short-lived progeny, based on theoretical analyses and conservative 

assumptions.

B.2 THORIUM PROCESS INVENTORIES AND THORON POTENTIAL

During the 35-year operating history of the Fernald site, approximately 70% of the time (24 years) 

involved the production of tens to hundreds of MTTh parts and products on an annual basis for a total 

of over 3,800 MTTh (Table B-1). The production of that quantity of products required processing of 

approximately 4,200 MTTh, and leaving over 750 MTTh of scrap and waste materials (see Tables B-1 

and B-5). On the surface this appears to be a large amount, although the primary material at FMPC 

was uranium. Several historical texts refer to amounts of thorium being “minor” in nature because the 

mass of thorium was generally less than 1% of the mass of RU receipts alone. Still, the tens to 

hundreds of metric tons of thorium in most of the FMPC facilities at any given time during the 

operational history was a recognized hazard and a source of thoron gas (220Rn), a very short-lived 

progeny of 228Th (thoron has a half-life of 56 seconds). Table B-1 outlines the rough thorium mass 

balance and estimated inventories during the indicated periods. 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

As shown above, the processing averages during the 24 years of thorium operations were in the 

range of 15 to 20 MTTh/mo or <1 MTTh/d, while process feedstocks (quantities in the process facility 

preparatory to processing) could have been in the range of 2 to 15 MT levels. That is, during 

processing campaigns small quantities were probably placed in the process facilities temporarily in 

queue for the operations [7]. Most of the storage was held in containers in adjacent storage buildings 

as well as barreled wastes held on pads outside the buildings

The primary products were metal cores for the Hanford Site and Savannah River Plant reactors in 

1954 to 1956 and a wide range of chemical forms and compounds in the succeeding periods as 

indicated in Table B-1. Process scraps, out-of-specification product, etc. resulted and became a 

storage or process category (or both). The processes were also varied, ranging from chemical, 

metallurgical, metal machining, waste storage, repackaging, and shipping.

The primary process from 1977 to 1979 was unusual in that it was primarily disposal and storage of 

excess reprocessed thorium materials from Hanford. It involved 6,900-gal tanks of nitrate solution, 

each with 10.8 MTTh for conversion to a stable and therefore storable thoria gel. These tanks were 

received at a rate of up to two per month for a maximum process rate of 21.6 MTTh/mo, which is less 

than 1 per day with an overall total of 351 MTTh. Uranium-233 was a contaminant of the thorium at 

approximately 130 g per tank load or approximately 12 ppm Th. The total mass of 4.2 kg of 233U was 

not separated from the thorium and was stored with the thorium (Grumski 1987).

The total mass inventory is only instructive in relation to the thoron inventory when the equilibrium 

ratio of 228Th to 232Th is known in addition to the fact that thorium-bearing materials were seldom 

100% thorium. The theoretical activity of 232Th is 1.09 × 105 pCi/g, but for consistency with other 

reports, including the safety analysis report for the thorium storage facility, a specific activity for 232Th 

and it progeny of 0.989 × 105 pCi/g of thorium materials will be used as an upper limit, which accounts 

for lesser densities of thorium compounds and other materials in the product.

The equilibrium age of the thorium processed at FMPC ranged from recently separated to 1 or 2 years 

since separation of the nonthorium progeny [8]. Figure B-2 illustrates the disequilibrium that occurs 

when thorium is separated from or purified from its progeny. Radon-220 is a progeny of the 1.9-yr 

half-life 228Th, which decreases to approximately 65% of equilibrium with 232Th in 1 year after 

separation of the radium and actinium progeny and 43% in 5 years after separation. Full isotopic 

equilibrium of 232Th and 228Th can be assumed for long-term storage materials after 35 years or so. 

For the functional period, 65% will be assumed, which represents a material 6 months to 2 years after 

separation and should be adequately conservative. Assuming 65% equilibrium would make 220Rn 

inventory equivalent to 0.64 × 105 pCi/g of 228Th compounds in the processes, and assuming full 

equilibrium with each of its progeny.

B.3 THORIUM WASTE REPOSITORY LOCATIONS AND QUANTITIES

FMPC was appointed the DOE thorium materials repository in 1972, and thorium processing was shut 

down in 1979. During the storage period up to 2,300 MTTh were stored pending final disposal off site 

in Nevada and Utah. The repository thorium wastes and surplus thorium materials were accumulated 

from Fernald and other sites in the repository era from 1972 until the “completion of remediation and 

the disposition of the containerized waste inventory” or final closure in 2006. See Table B-2 for a 

tabulated summary as of 1987of 13,329 drums of thorium plus 184 MTTh in bulk powders (or 

granules) or liquid. Over 689 MTTh were stored in the Pilot Plant warehouse and 747 MT in four 

other designated storage buildings. The first shipment from this inventory was to the EnviroCare 

facility in Utah.
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Table B-2. Thorium storage inventory (MTTh), 1987.a

Plant 

area

Storage 

building Drums

Net weight

(lb)

Th weight

(lb) % Th MTTh

Plant 1 67 5,992 426,978 333,044 78 151

Plant 5 65 5,599 2,492,505 711,210 29 323

Plant 5 65W pad 240 105,032 92,955 89 42

Plant 8 Silo and bins Bulk material Not applicable Unknown 50 175

Plant 9 64 181 128,345 123,141 96 56

Pilot 

Plant

68 1,317 971,398 745,785 77 338

Pilot 

Plant

Liquid tank Nitrate solution Not applicable Unknown Not applicable 351

Total Not applicable 13,329 Not applicable Not applicable Not applicable 1,436

a. Sources: Dolan and Hill (1988); DOE (1987b); Mead, Savage, and Fugate (ca. 1986); Clark et al. (1989); Grumski 

(1987); Bogar (1987); Starkey (1964).

B.4 FACTORS TO CONSIDER IN DETERMINING POTENTIAL ALPHA ENERGY 

CONCENTRATION

The primary dose from thoron is to the respiratory tract organs and is delivered primarily by the shortlived progeny (Table B-3). Due to the short radiological half-lives and lung clearances of the 

particulate progeny, there are no practical bioassay measurements to define intake (NIOSH 2013; 

UNSCEAR 1994; Bodansky, Robkin, and Stadler 1987; Evans and Goodman 1940; Strom and Reif 

1996). Therefore, the method of choice for dose evaluation is to measure the potential alpha energy 

exposure (PAEE; see Glossary), which is the potential alpha energy concentration (PAEC) in air 

multiplied by the time the worker was exposed. The PAEC is measured in working levels (WLs), and 

the PAEE is measured in WLM. The bounding defaults recommended in this section are listed as 

WLM, which relate directly to total alpha energy exposure.

Table B-3. PAEC for thoron and progeny.

Isotope Half-life Atoms/7.5 pCi α MeV/atom

PAEC

(MeV/7.5 pCi)

Rn-220 56 s 23 14.6 335

Po-216 0.15 s <1 Included with Rn-220 Included with Rn-220

Pb-212 10.64 hr 15,476 7.8 1.21E+5

Bi-212 60.6 min 1,469 7.8 0.12E+5

Po-212 0.3 μs Not required Included with Bi-212 Included with Bi-212

Tl-208 3 min 73 No alpha 0

Total Not applicable Not applicable Not applicable 1.33E+5

Air Monitoring

From the beginning of operations in the early 1950s the radiological safety program evaluated 222Rn 

and its progeny in work areas with specific radon and progeny measurements. The health and safety 

staff was aware of thoron dose potential as evidenced by a number of samples and references to 

specific thoron analyses, which have been recovered or mentioned in staff reports (Weaver 1987; 

Starkey 1964). During the period after 1989, when WEMCO assumed the contractor role at the 

Fernald site, extensive and recorded radon and thoron monitoring was performed across the site 

(Weaver 1987). Air monitoring for particulate activity was also performed throughout the production 

period, and enough air sampling data has been recovered for realistic conclusions and assumptions 

about exposure potentials.
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Work Place Control Functions

To better understand the workplace ventilation dynamics and air movements, Table B-4 provides a 

summary of the dimensions of the buildings in the Fernald complex that handled thorium in any 

amount. The process ventilation systems, which include an air moving system for maintaining a 

negative pressure within the process equipment, a baghouse particulate filter (a few liquid scrubbers 

were also used), and individual stacks for release of filtered air to the environment. It was reported 

that there were a total of 94 of these ventilation systems and stacks during the early periods, the 

number of which changed with changing processes and projects. A total of 114 ventilation systems is 

indicated in Table B-4, and the total stack flow for each building is a summary of all the individual 

stack flows in that building. Although there were some general building vents for assistance in 

temperature reduction, these were not used continuously. The air supply for the stack discharge 

flows came from the building working environment around the containment systems, which in turn 

enclosed the process equipment. The pressure differential maintained the airflow into the equipment 

in which the thorium was being processed. As indicated in Table B-4, the air changes per hour for the 

entire facility volume would average approximately three when all the individual ventilation systems 

were operating. The air change rate in the immediately vicinity of the individual processes would 

obviously be higher (Dolan and Hill 1988; DOE 1987b; Mead, Savage, and Fugate ca. 1986; Clark et 

al. 1989; Grumski 1987; Bogar 1987; Starkey 1964).

Table B-4. Building dimensions and ventilation systems.

Building Dimensions (ft) Volume (ft3)

Volume

(m3)

Number of 

ventilation 

systems

Total 

stack 

(ft3/min)

Total 

facility 

chgs/hr

Plant 1 100 × 160 × 60 9.6E+5 2.72E+4 15 6.1E+4 3.8

Plant 2/3 60 × 380 × 67 1.53E+6 4.33E+4 6 3.2E+4 1.3

Plant 4 225 × 165 × 94 3.49E+6 9.88E+4 12 4.38E+4 0.8

Plant 5

and Building 55

650 × 100 × 52 3.38E+6 9.57E+4 17 1.5E+5 2.7

60 × 30 × 51 0.92E+5 2.61E+3 1 4.0E+3 2.6

Plant 6 500 × 200 × 39 3.9E+6 11.0E+4 6 1.35E+5 2.1

Plant 7 110 × 80 × 114 1.0E+6 2.83E+4 4 1.7E+4 1.0

Plant 8 60 × 260 × 48 7.49E+5 2.12E+4 24 8.22E+4 6.6

Plant 9 300 × 225 × 40 2.7E+6 7.65E+4 4 7.89E+4 1.8

Pilot Plant 210 × 235 × 54 2.66E+6 7.53E+4 9 4.37E+5 9.9

Building 64 50 × 320 × 22 3.52E+5 1.0E+4 8 1.6E+4 2.7

Building 65 50 × 210 × 22 2.31E+5 0.65E+4 8 1.6E+4 4.2

Average or 

total

Not 

applicable

Not

applicable

5E+4

average

114

 total

Not 

applicable

3.3 

average

Release Fraction of Stored Materials

Four examples are provided to address the release fraction (RF) of both 222Rn and 220Rn gases and 

provide a basis for estimating the release fraction of radon gases to the work environs:

1. To bound the release of radon gas from a source of stored barrels of known waste 

concentration, measurements taken in a shipping car (~6 × 104

-L volume) with 48 barrels of 

raffinate wastes with a 226Ra concentration of 400 nCi/g were used. This is the maximum 

measured isotopic characterization of silo waste – determined at a much later date. A level of 

approximately 25 Ci of 226Ra was calculated as the inventory in the 48-drum waste shipment. 

The maximum measured 222Rn progeny concentration (the highest of a number of samples) in 

the transport carrier was 400 pCi/L in the unventilated shipping car. Using these values, the 

de facto 222Rn RF from the drums is calculated to be 400 pCi/L ÷ (2.5 × 1013 pCi/6 × 104 L) = 

1 × 10-6. These shipping containers were “enclosed containers,” but obviously not vapor tight. 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

This is a measurement of 222Rn with a 3.8-day half-life and the associated higher diffusion 

capabilities, in comparison with 220Rn, which would have shown a much lower RF.

2. Another example and specifically for short-lived thoron gas release is found in the maximum 

measured 220Rn progeny levels in Building 65 (0.7 × 104

 m3) in 1996. Building 65 is one of the 

primary DOE repository storage facilities and the maximum of many measurement 

concentrations was 267 pCi/L. Assuming this measurement was the progeny of 220Rn, the

radon level would be 1.3 × 104 pCi/L. The facility contained approximately 421 MTTh at the 

time of the measurement with the associated 27 Ci of 220Rn. Some of the containers had not 

been double packed, and the lower level drums were in poor repair – corroded and leaking. 

Total release of the 220Rn inventory would result in 2.7 × 1013 pCi/0.7 × 107 L = 4 × 106 pCi/L 

for an RF of 1.3 × 104

 ÷ 4 × 106 = 3 × 10-3. There are many unknowns in this example: 

ventilation of the building, fraction of the stored material containers that were breached, 

location of the air sample in relation to leaking drums, etc. Therefore, it is not be used as a 

quantitative example, but rather to indicate that the RF for some storage facility operations is 

be in the 1 × 10-3 range and is more likely to be comparable to the storage silo example below.

3. A third example can be demonstrated through measurements made at the top of the Plant 8 

3,000-ft3 storage silo (an elevated cylinder, 17 ft in diameter and 23 ft high with conical 

sections on the bottom and top) with approximately 85 MTTh. Thorium oxide and hydroxide 

granular material at approximately 50% thorium had a measured density of 200 lb/ft3 in the top 

of the oxides, which provides a thorium density of 100 lb/ft3

. An air activity concentration of 

71 WL was measured at the top of the tank, which had a 15-ft high by 17-ft diameter air space 

in the tank above the thorium materials (1 × 105 L). The cone-shaped surface area at the top 

thorium material (it was deposited by a conveyor to the center of the silo) was calculated as 

approximately 350 ft2

. 

• With a diffusion distance of approximately 2 cm (Banjanac et al. 2006; and assumed 

1 in.), the effective thoron quantity available for diffusion to the headspace was 

calculated as:

• (350 ft2

)(1/12 ft)(100 lb/ft3)(454 g/lb)(0.64 × 105 pCi/g) = 8.5 × 1010pCi

• This compares to (85 MTTh)(6.4 × 1010 pCi/MTTh) = 5.4 × 1012 pCi in the total silo or 

1.6% of the 220Rn available for release.

• The maximum expected 220Rn in the 1 × 105

-L headspace volume is (8.5 × 1010 pCi) ÷ 

1 × 105 L = 9 × 105 pCi/L (total 220Rn in 1 in. of the surface).

• This is compared to a measured 71 WL= (375 pCi/L 220Rn/WL)(71 WL = 3 × 104 pCi/L.

• This results in an effective RF of (3 × 104

) ÷ (9 × 105 = 3 × 10-2) of that within 1 in. of 

surface.

• Therefore, the RF considering the total 220Rn (from 85 MTTh) in the silo is (3E+4) ÷ 

(5.4 × 1012) ÷ (1E × 105

) = 6 × 10-4. 

From this example we see that the Plant 8 storage silo is essentially a big barrel of thorium, 

and the amount of 220

Rn in the headspace would be on the order of 1 × 10-3 to 1 × 10-4 of the 

total in the container. Migration of the 220Rn from the container (big or small) to the work 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

environment would be in the 1 × 10-6 range (reasonably estimated based on 222Rn migration 

from barrels of waste).

4. The fourth example is found in the Plant 8 elevated double-bin storage tanks. Similar to the 

silo example above, a lesser RF is calculated, which further verifies the conservatism in the 

assumptions.

Therefore, it is judged to be conservative to use an RF of 1 × 10-6 for stored inventories in 

intact but unsealed containers (10- to 50-gal drums, etc.) and 1 × 10-4 for areas immediately in 

the vicinity of open thorium materials such as silo storage, leaking drums in closed buildings, 

etc.

Release Fraction of Materials in Process

For process areas the RF can be conservatively calculated based on general thorium particulate air 

activity in comparison with the total thorium material in process. Routine air sampling in the operating 

areas recorded thorium concentrations ranging from tens to a few samples in the tens of thousands of 

dpm/m3 and mostly in the very early period of 1954 to 1955 when processes were being developed. It 

could then be assumed that a 1-MT (1 × 106

-g) daily process load in a 5 × 104-m3 volume process 

building in which process releases had resulted in 1 × 104 dpm/m3

. However, it is not reasonable to 

assume that the air concentrations in the building would have been at the maximum levels throughout 

the entire area, and these levels should be assumed to exist in a 20-ft hemispherical air space near 

the process equipment (475 m3

). This is based on the fact that the higher value air samples used 

were undoubtedly BZ samples taken during high-level maintenance (e.g., clean-out evolutions, etc.). 

For these reasons it is reasonable to assume that the air samples represented maximum 

concentrations in the immediate vicinity of the equipment rather than average building concentrations. 

Therefore, the concentration in the air, if all the process materials were completely released, would be 

(1 × 106 g) (0.64 × 105 pCi/g) (2.22 dpm/pCi) ÷ 475 m3 = 0.3 × 109 dpm/m3

. The RF would be 

(1 × 104

)(0.3 × 109) = 3 × 10-5. 

Some injection operations (dumping barrels of feed material into hoppers, etc.) could be expected to 

release the free thoron in the headspace of the container, which would be a ratio of 2 cm to the height 

of the materials in the container – in the range of 0.1 to 0.01 as indicated in the discussion of release 

fractions from stored materials. Even then the particulate dusts would serve as an absorption and 

removal mechanism of the progeny and clear in a short period of time (limiting the exposure time). An 

RF of 0.01 was assumed. There was a variety of process equipment that was used, including liquid 

extraction vessels, furnaces, mixers, etc., all of which were designed for containment.

Equilibrium of Thoron and Progeny 212Pb, Bi-212, Po-212 and Tl-208

Quoting from Harley and Chittaporn (2006) of the New York University School of Medicine, “The 

inhaled solid decay products of thoron deposited on the lung airways deliver the bronchial dose.” 

Harley (2004) also reports:

... the Feq [fraction of equilibrium] cannot exceed 0.04 and calculations of Feq for indoor 

radon and thoron confirms the values of 0.4 and 0.02. The value of Feq for thoron is 

supported by two large data bases, one reported by NIRS [National Institute of 

Radiological Science] with measurements in China and one reported by NYU [New York 

University] in residences and also at the former uranium processing facility at Fernald, 

OH.
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Pillai and Paul (1999) reported equilibrium values of thoron and its progeny of 0.002 to 0.007 in a 

monazite processing plant in India. This analysis uses 0.02 as adequately bounding.

Summary Production Rate and Available Thoron Inventory for Release

As indicated above the average daily production rate was <1 MTTh (ranging from 0.03 to 0.8) with 

6 × 10-2 Ci thoron. Temporary storage in the production facilities during processing campaigns could 

be 5 to 15 MTTh with the associated 0.3 to 0.9 Ci thoron. DOE long-term storage at the Fernald site 

was 100 to 450 MTTh in any of the storage facilities with the associated 6 to 27 Ci thoron.

Occupancy Time

The nature of storage facilities results in a minimal occupancy time for required functions. This 

analysis assumes 500 hr/yr (or 3 mo/yr) for routine storage conditions accompanied by required 

repackaging. For long-term storage of high-integrity storage containers, 1 mo/yr is assumed. The 

time of occupancy during production periods will be assumed as 1750 hr/yr, because an average 

1-MTTh continuous production rate is assumed. At this 1-MTTh daily rate, the recorded annual 

quantities processed would be completed in just a fraction of a year.

Demonstrated Local Diffusion Factor

It has been demonstrated that for thoron progeny, the concentrations are approximately a factor of 10 

higher near the source of the release than the concentration at a distance of 3 ft (Stranden 1980). 

Because this is probably dependent on ventilation, this demonstrated reduction factor to the BZ of 

workers is ignored.

Respiratory Protection

It was established practice to wear respiratory protection during the processing of large volumes or 

masses of hazardous materials. The wearing of respirators would remove essentially all of the 

progeny through filtration and electrostatic attraction in the filter. However, in the interest of 

favorability to the claimant due to a few recorded cases of procedure violation (there were sample 

sheets and other logs that made note of personnel not wearing masks), it is assumed that no 

respirators were worn. The notes themselves indicate an unusual condition. Therefore, this is an 

admitted overly conservative assumption based on both (1) the established protection policies and 

practices and (2) demonstrated difference in calculated intakes of thorium based on air monitoring 

results in comparison with in vivo measured lung burdens.

The following summarizes the assumptions for thoron exposures:

1. Materials in process averaged less than 1 MT/d and less than 400 MT/yr.

2. Materials in process facilities in temporary storage ranged from 2 to 15 MT, and 15 is 

assumed.

3. Long-term DOE storage ranged from 100 to 450 MT in any given storage location, so 450 is 

assumed.

4. Specific activity of thoron in thorium materials at Fernald is 0.64 × 105 pCi/g (e.g., 228Th/232Th = 

0.65).

5. Facility volumes were large (averaging 5 × 104

 m3). In the interest of conservatism, it is 

assumed that the mixing or exposure volume was near the process equipment and 

represented approximately 20-ft radius, hemispherical volume in the total facility (475 m3

).
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

6. RF in temporary or long-term storage is 1 × 10-6. 

7. RF in temporary open storage or with compromised containers is 1 × 10-3 to 1 × 10-4. 

8. RF in general process areas is 1 × 10-5; 1 × 10-4 was assumed for Plant 9 during process 

development period in 1954 to 1955.

9. Fraction of equilibrium (Feq) for thoron is 0.02. In a plant configuration with large buildings and 

engineered ventilation, the recommended value of 0.02 is judged to be conservative. In this 

case, it would require a concentration of 220Rn/216Po at 375 pCi/L to produce 212Pb/212Bi 

concentrations at 7.5 pCi/L and therefore result in 1 WL.

10. Diffusion factor of thoron progeny from source release point is 1 (no reduction). Fraction of 

progeny near source compared to 3 ft away is 0.1, but 1 is assumed.

11. Occupancy time in storage facilities is 500 hr/yr and 1 mo/yr for long-term storage.

12. Occupancy time in process facilities is a maximum of 1,750 hours.

13. Protective respiratory protection is assumed not to have been used.

14. Thoron WL is 7.5 pCi/L with 100% equilibrium with progeny, which is 1.3 × 105 MeV/L 

(Table B-3. WL = 375 pCi/L 220Rn and 216Po to produce measured progeny 212Pb and 212Bi at 

7.5 pCi/L (see the discussion of Feq factor of 3 and summary item 9 above.)

Estimates of possible exposures to thoron indicate that for some periods and in some locations 

(primarily long-term storage areas in which some extended work periods were required), exposures 

could be as high as 1.6 WLM/yr. The work could involve activity with poorly ventilated facilities with 

open containers, such as repackaging failed containers, preparing for shipment, etc. Even though the 

RFs are relatively small for both storage configurations and operational conditions and the short halflife of 220Rn limits the diffusion range in all materials to 2 cm or less, possible quantities and exposures 

can still be of concern. UNSCEAR (1994) indicates that due to the relatively longer half-lives of the 

progeny of thoron in comparison with radon, significant dose can be experienced with relatively small 

WLM exposures.

Table B-5 provides a comparative summary of the WLM intake estimates during the entire operational 

period including the SEC periods. Table B-6 provides a summary of the WLM intake of 220Rn progeny 

that could have occurred during processing of thorium in the post-SEC periods. The 1.6-WLM 

estimate for work in the storage facilities in preparing thorium materials with higher integrity containers 

and preparations for shipment is conservative. The work would require a very limited time and not the 

entire indicated 17-year storage period. In addition, the work was recognized as requiring special 

controls including protective clothing, respiratory protection, and ventilation.
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

B.5 NIOSH RESPONSES TO SC&A COMMENTS ON METHODS FOR RECONSTRUCTING 

THORON DOSES AT FERNALD PRESENTED AT THE 12/4/14 WORKING GROUP 

MEETING

B.5.1 Background

At the Fernald Working Group meeting on September 4, 2014, NIOSH presented a white paper called 

Fernald Dose Reconstruction Methodology for the Post Special Exposure Cohort (SEC) Period, 1979-

2006 (NIOSH 2014c). In addition to thorium recommendations, the paper included thoron exposure 

recommendations (WLM per year), based on the information from Sections B.1 to B.4 to generate 

Tables B-5 and B-6. 

Table B-5. Summary of thoron exposure estimates.

a

Period and location

Available 

thoron 

activity

Reduction and 

calculation 

factors

Effective 

facility 

volume 

(L)b

Rn-220/

Po-216 

activity

(pCi/L)

Pb-212/

Bi-212 

activity

(WL) WLM/yr

1954–1955, Plant 9 and Plant 4

In process per day

1 MTTh/d = 

6.4E+10 

pCi/d Rn-220

RF 1E–4

Feq 2E–2 

Focc 10.2 mo/yr

5E+5 12.8 0.034 0.35

Available for release 15 MTTh/d = 

1E+12 pCi/d 

Rn-220

RF 1E–6

Feq 2E–2 

Focc 2.9 mo/yr

5E+5 2 5.3E–3 0.016

1959–1963, Plant 6

In process per day

0.02 MTTh/d 

= 1.3E+09 

pCi/d Rn-220

RF 1E–5

Feq 2E–2 

Focc 10.2 mo/yr

5E+5 0.026 3.5E–3 Not 

significant

Available for release 2 MTTh/d = 

1.3E+11 pCi 

Rn-220

RF 1E–6

Feq 2E–2 

Focc 2.9 mo/yr

5E+5 0.26 NS Not 

significant

1965 and 1968–1972, Plant 8

In process per day

0.03 MTTh/d 

= 2E+09 

pCi/d Rn-220

RF 1E–5

Feq 2E–2 

Focc 10.2 mo

5E+5 0.04 NS Not 

significant

Available for release 2 MTTh/d = 

1.3E+11 pCi 

Rn-220

RF 1E–6

Feq 2E–2 

Focc 2.9 mo/yr

5E+5 0.26 NS Not 

significant

1954–1956 and 1964–1979, Pilot Plant

In process per day

0.8 MTTh/d = 

5E+10 pCi/d 

Rn-220

RF 1E–5

Feq 2E–2 

Focc 10.2 mo/yr

5E+5 1 0.0027 0.027

Available for release 15 MTTh/d = 

1E+12 pCi 

Rn-220

RF 1E–6

Feq 2E–2 

Focc 3 mo/yr

5E+5 2 0.0053 0.016

1954–1989, various storage sites

Available for release

300 MTTh = 

2E+13 pCi 

Rn-220

RF 1E–4

Feq 2E–2 

Focc 3 mo/yr

1E+7 200 0.53 1.55

1972–2006 ,final closure storage

Available for release

300 MTTh = 

2E+13 pCi

Rn-220

RF 1E–4

Feq 2E–2 

Focc 1 mo/yr

1E+7 200 0.53 0.53

a. Feq = fraction of equilibrium; Focc = occupancy time; NS = not significant.

b. For maximizing purposes, the mixing of the thoron and progeny during the processing of thorium was assumed to be in 

a 20-ft hemispherical volume immediately around the release point, which in turn is assumed to be the process 

workstation and for the short-term storage of thorium in the process facilities. For long-term storage facilities, the 

volume of the storage facilities was used.
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Table B-6. Thoron exposure. 

Period Plant

WLM intake each year 

during period

1954–1956 Plant 4 and 9 0.4

1959–1963 Plant 6 No significant dose

1965 and 

1968–1972

Plant 8 No significant dose

1977–1979 Pilot Plant 0.03

1972–1989 Storage facilities, repackaging, etc. 1.6

1972–2006 Closure, various storage 0.5

Discussions at the December 4, 2014 Fernald Working Group meeting included a Sanford Cohen & 

Associates (SC&A) presentation, SC&A Review of Proposed NIOSH Methods for Reconstructing 

Thorium Doses at Fernald (1979–2006) (SC&A 2014). Slides 32 and 33 of the presentation list 

SC&A’s summary and critique of NIOSH’s “Major assumptions” regarding thoron dose reconstruction:

1. The white paper appears to contradict itself on the assumption of 300 MT thorium in the 

storage sites (the preceding page appears to quote 450 MT, the introduction quotes over 

2,000 MT of thorium materials in addition to the Plant 8 silos and bins and Pilot Plant storage).

2. The release fraction should be better established. Quoted release fractions in the white paper 

appear to range from 1 × 10-6 all the way to 1 × 10-3. 

3. The equilibrium factor of 0.02 (or 2%) is not well founded. The stated reference indicates 

equilibrium factors could range from 2% to 10%, but also notes “more precise studies are 

warranted.”

4. SC&A could not determine the rationale behind occupancy times of 3 months (up through 

1989) and 1 month (during final closure).

5. The specific activity of thoron was given as 6.4 × 10-4 pCi/g assuming exposures occurred 6 to 

12 months after separation and an equilibrium fraction of 228Th/232Th of 0.65. The equilibrium 

fraction of materials stored in Building 65 was at least 0.95.

A related item under “Main Conclusions” on Slide 34 states:

The main parameters for estimating exposure to thoron should be better described and 

established to assure that thoron exposures are calculated in a scientifically defensible, 

claimant-favorable and/or bounding approach.

B.5.2 Discussion

The information from Sections B.1 to B.4 to generate Tables B-5 and B-6 were not intended to provide 

precise estimates. Nor is it even possible to do so because of the general lack of site-specific thoron 

data by time and location. Rather, it was developed to provide a reasonably conservative estimate of 

the exposures. It is clear from the site documents collected that thoron was recognized as a potential 

problem from the start of thorium operations. There are numerous references to thoron in site 

documents. For example, see the measurements in Plant 9 from 1954 that are discussed in the 

following sections (NLO 1954). In March 1980, a standard operating procedure entitled Radiological 

Controls for Repackaging Thorium specified that 220Rn working-level samples be taken at least twice 

per shift (WMCO 1990c, p. 6). In 1998, the Plant 9 Project Radiation Protection Plan specified that a 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

minimum of 25% of the workers in each work group/crew (minimum of 1 worker) would be required to 

wear a pump and filter assembly to monitor for radon and thoron (FEMP undated, p. 112). 

NOTE: The comments are addressed in the following sections. Comments 2 and 4 are 

addressed together, after Comments 1, 3, and 5.

SC&A Comment 1 (Thorium Inventory): The white paper appears to contradict itself on the 

assumption of 300 mT thorium in the storage sites (the preceding page appears to quote 

450 mT, the introduction quotes over 2000 mT of thorium materials in addition to the plant 8 

silos and bins and pilot plant storage).

NIOSH Response: SC&A’s statement that the introduction quotes over 2,000 MT of thorium materials 

apparently refers to the total of the net weights in Table B-2. This table shows the estimates for 

various storage locations. The MT weight in storage locations is the source of the discrepancy 

between the 450 MTTh in Sections B.1 to B.4 and the 300 MTTh used in the calculations for storage 

areas. The data in Table B-2 come primarily from Dunaway (1990). These same data are also in 

Collier (1987) and Weinreich (1987). The 175 MTTh in the Plant 8 Silo and bins is from the 1985 data 

in NLO (1965–1988). The 351 MTTh in the Pilot Plant “Liquid Tank” represents the thorium nitrate 

shipped from Hanford in the late 1970s. There are various descriptions of the quantity of thorium 

nitrate in the site documents (e.g., Bonfer 1988; FMPC 1952–1992; Dunaway 1977). Most of the 

material was converted into dried thoria gel. There were 6,500 gal of thorium nitrate remaining in 

1987, as noted below. Documents also vary on the quantity of thoria gel produced (see Dolan and 

Hill 1988; RCK 1997; Heatherton 1977; Audia 1979). The 338.3 MTTh quantity in Audia (1979) 

matches the quantity stored in Building 68 in 1987 (Dunaway 1990). In 1987, only 9.2 MT of thorium 

in thorium nitrate was stored in Tank T-2 at the Pilot Plant (NLO 1985, p. 176). Therefore, the use of 

351 MTTh represents a bit of double counting of some of the thorium. The value of 450 MTTh

represents an upper bound on the thorium stored in any one location, but the 300 MTTh quantity used 

in the calculations better represents a more typical storage quantity. Using the 450 MTTh estimate 

with the existing assumptions would result in a 50% increase in the thoron and thoron daughter 

concentrations to 300 pCi/L and 0.8 WL, respectively. As will be seen in the measurements 

presented below, this increase would not affect the conclusion of this appendix.

SC&A Comment 3 (Equilibrium Factor): The equilibrium factor of 0.02 (or 2%) is not well 

founded. The stated reference indicates equilibrium factors could range from 2%–10%, but 

also notes “more precise studies are warranted.”

NIOSH Response: SC&A correctly states that the range of equilibrium factors between thoron and its 

daughters was reported by the site to range from 0.02 to 0.10 (Daniels 1995, p. 13). It is not clear 

how these values were determined. In a 2006 paper, Harley et al. reported that the indoor thoron 

equilibrium factor is 0.02 and cannot exceed 0.04. The large diffusion coefficient of the very shortlived 216Po results in a high removal rate to surfaces. The value of the equilibrium factor for thoron 

was supported by two large databases, one of which included measurements made at Fernald 

(Harley and Chittaporn 2006).

SC&A Comment 5 (Assumption of 228Th/232Th Equilibrium Fraction): The specific activity of 

thoron was given as 6.4 × 10-4 pCi/g assuming exposures occurred 6–12 months after 

separation and an equilibrium fraction of Th-228/Th-232 of 0.65. The equilibrium fraction of 

materials stored in Building 65 was at least 0.95.
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ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

NIOSH Response: The estimates were generalized to cover all areas over the site’s history. There 

are individual examples that do not fit the conditions assumed in that discussion. This comment 

points out one of them. However, using the 323 MTTh stored in Building 65 with a 228Th-to232Th 

equilibrium fraction of 0.95 would result in only a 50% increase in the thoron and thoron daughter 

concentrations to 300 pCi/L and 0.8 WL, respectively. As with the 450 MTTh example in the response 

to Comment 1, this increase would not affect the conclusion within this appendix.

Comment 2 (Release Fraction): The release fraction should be better established. Quoted 

release fractions in the white paper appear to range from 10-6 all the way to 10-3.

Comment 4 (Occupancy Time): SC&A could not determine the rationale behind occupancy 

times of 3 months (up through 1989) and 1 month (during final closure).

NIOSH Response: Rather than focus on SC&A’s immediate concerns about release fractions and 

occupancy times, this response seeks to demonstrate that the estimates provided are consistent with 

actual measurements that NIOSH has analyzed. It is acknowledged beforehand that this comparison 

will not be totally satisfying. Because the measurements are either in terms of thoron concentration in 

pCi/L or thoron daughters by WL, the comparison involves comparing dose rates, not total doses. 

However, if the dose rates are shown to be as low as, or lower than, the estimates, the discussion will 

be moved forward. Then, perhaps some agreement can be reached on occupancy factors for select 

populations. In the discussion below, thoron measurements and thoron daughter measurements will 

be addressed separately.

There are some issues with the available data. Most of the data is from 1986 and later. These data 

have the disadvantage of not covering the earlier periods. The period before 1979 is already included 

in the SEC, which means that the respiratory track cancers are already being compensated for much 

of the site’s history. On the other hand, many of the later measurements were made with improved 

technology, such as track-etch detectors and WL monitors, which likely collected more and better data 

than would have been possible in the early years. However, even some fairly recent WL 

measurements were done with short-term (on the order of minutes) air sample measurements that 

were recounted to determine the contribution from thoron daughters. This probably resulted in greater 

variability in the measurement values. Many of the later-period measurements were taken in 

locations where thoron levels were known to be a potential problem which likely resulted in the mean 

of the measurements being higher than the true site mean.

B.5.3 Thoron Measurements

Thoron calculations based on air samples collected in 1954 for Plant 9 are in NLO (1954). The 

concentrations were recalculated from the data on sample worksheets. They are discussed in 

Section B.5.6. The thoron concentrations were 21.8 ±17.2 pCi/L before the ventilation was turned off. 

The 12.8 pCi/L estimate in Sections B.1 to B.4 lies within one standard deviation of these results. The 

underlying thoron daughter measurements will be discussed below.

Thoron measurements for 1996 to 1998 have been recovered. The measurements were obtained 

with track-etch dosimeters and were reported in pCi/L. The dosimeters were calibrated by placing two 

types of dosimeters in a chamber and exposing them to known concentrations of radon and thoron. 

Both dosimeter types prevented the entry of the daughter products. One type (“filter”) allowed both 

radon and thoron gas to enter the dosimeter and expose the CR-39 (or equivalent) plastic. The other 

type (“membrane”) retarded the infiltration so that most of the thoron would decay before exposing the 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

CR-39, resulting in a measurement of radon only. The difference in the two readings (in tracks per 

unit area) was interpreted as the thoron exposure in pCi/L (Pearson and Spangler 1991).

The track-etch dosimeters were deployed for at least one quarter (3 months). Generally, over 40 

locations were measured simultaneously, including both high- and low-exposure-potential locations. 

Over 300 quarters of data were accumulated in the three-year program. For this appendix, the annual 

data were fit to lognormal distributions. All three data-weighting methods produced similar results. 

Table B-7 summarizes the data. The GM and 95th percentile are shown for the “uniform” (unweighted 

least squares) lognormal fit. 

Table B-7. Summary of the 1996 through 1998 thoron measurement data (pCi/L). 

Period Locations Geometric mean 95th percentile Reference Page(s)

1996 68 0.435 7.92 Renk 1988 65-66

1997 42 0.470 1.66 Renk 1988 38

1998a 54 0.524 1.61 Renk 1988 38

a. Annual value was not available in reference; average for 3 quarters was calculated.

The data in Table B-7 may be compared to the thoron concentration estimate of 200 pCi/L for “1954-

1989 Various Storage Sites” and “1972-2006 Final Closure Storage” (Table B-5). This comparison is 

plotted in Figure B-3, which shows the GMs and 95th percentiles for all three fitting methods reported 

by the Lognormal Fitting Utility Program (Version 2006-10-12.0; Battelle 2007). It is apparent from the 

comparison that the Sections B.1 to B.4 estimate of 200 pCi/L is at least a factor of 25 higher than the 

95th percentile of the 1996 to 1998 measurements.

There are two individual measurements cited in the references that are worth discussing. The first is 

a measurement of 267 pCi/L for Building 65, which was included in the 1996 distribution above. It is a 

high outlier, as can be seen in Figure B-4. The measurement was for December 1995 through March 

1996. There were no measurements for the other three calendar quarters in this location. The 

second measurement for discussion is also from Building 65. It was located in the 1995 ALARA 

design review for the Building 65 Thorium Overpacking Project (Allen 1995a, p. 31). This reference 

cites a thoron concentration of 112.75 pCi/L based on “thoron cup data” (another term for track-etch 

dosimeters). This was apparently a value determined with only natural ventilation, since it was used 

in the design review to calculate the effect of forced ventilation through HEPA filters (Allen 1995a, 

p. 32). At the time, 5,600 drums of thorium were stored in Building 65 (Allen 1995a, p. 10) with very 

limited access and with radiological controls imposed. The design parameters in various documents 

assumed that only 20 workers would be involved in the overpacking of Building 65 (e.g., Toler 1995, 

p. 187). The above-cited measurements were the only two values greater than 100 pCi/L found in the 

site documents. From this discussion, it is apparent that Building 65 was a separate distribution. 

Although the 267 pCi/L was included in the 1996 analysis, it really should not be considered as a part 

of a site-wide distribution of thoron results.

Additionally, the 267 pCi/L measurement from Building 65 was used, along with the inventory of 

stored thorium to generate a release fraction of 3 ×103

. This appears to be an error, as the 267 pCi/L 

was assumed to be a progeny measurement, and used an equilibrium factor of 0.02 to revise the 

radon concentration to 1.3 × 104 pCi/L. Renk (1988, p. 67) indicates that this result was a thoron track 

etch result, meaning it was a measurement of thoron gas, not progeny. Consequently, the release 

fraction should be 50 times lower than 3 × 10-3, or more in the 3 × 105 range.
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

B.5.4 Thoron Daughter WL Measurements

It is useful to illustrate how the WL estimate in Sections B.1 to B.4 was calculated from the thoron 

estimate. The parameters are:

• 200 pCi/L, calculated as a conservative value in Sections B.1 to B.4 for “1954-1989 Various 

Storage Sites” and “1972-2006 Final Closure Storage” (Table B-5). This covers the 

periods/locations where measurements are available.

• 7.5 pCi/L/WL, for thoron with daughters in full equilibrium. 

• 0.02 to 0.04 (2% to 4%), equilibrium factor. As Comment 3 points out, the equilibrium factor 

range listed in site documents is 0.02 to 0.10 (e.g., Daniels 1995, p. 13). However, later 

research (which included Fernald data) indicates that the indoor equilibrium factor is 0.02 and 

cannot exceed 0.04 (Harley and Chittaporn 2006, p. 4).

Therefore, for 200 pCi/L thoron, the WL for thoron daughters with an equilibrium factor of 0.02 is:

(B-1) daughters

200pCi L 0 02 0 53 WL pCi 7 5

L

WL . .

. WL

= ×=

Or, for an equilibrium factor of 0.04:

(B-2) daughters

200pCi L 0 04 1 07 WL pCi 7 5 WL

L

WL . .

.

= ×=

Note that a value of 0.53 WL was used in Sections B.1 to B.4 for “1954-1989 Various Storage Sites” 

and “1972-2006 Final Closure Storage” (Table B-5). The 1.07-WL value is also shown on the WL plot 

in this appendix (Figure B-5).

The WL measurements that were obtained are as limited as the thoron measurements. As noted 

above, some WL measurements were done by recounting air samples to determine the contribution 

from thoron daughters. As such, some measurements represent the WL averaged over only several 

minutes. When several measurements of this type were obtained from one location or facility, they 

are the most useful; but, unfortunately, this was not always the case.

As discussed above, thoron daughter measurements in Plant 9 in 1954 were obtained (NLO 1954). 

The results, which required some manipulation, are in Attachment 1 of this appendix. Briefly, the 

measurements were taken at six locations to determine the effect of turning off the ventilation. A 

series of five measurements were taken at five of the locations. The samples were counted twice. 

The difference in concentrations between the first and second counts was due to the 212Pb, 212Bi and 

212Po activity. This concentration can be converted to units of pCi/L and then can be divided by 

7.5 pCi/L per WL to determine the number of WLs. The first 11 measurements were taken before the 

ventilation was disrupted. These measurements averaged 0.058 ±0.05 WL (±1 standard deviation). It 

is not clear whether any additional measures were taken to maximize the effect of the reduced 

ventilation, such as closing windows and doors that would normally be open. Given this uncertainty, 

the results appear to be consistent with the Sections B.1 to B.4 value of 0.034 WL. Because of their 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

unique nature, these measurements are not included in the analyses of the 1986-and-later results 

below.

Due to the lack of data, both individual measurements and distributions of measurements were 

included in the 1986-and-later analysis. For Buildings 64 and 65, WL measurements were obtained in 

1996 by using WL monitors. The detectors recorded hourly measurements for 7 to 8 days. These 

measurements are different from most of the discovered data in two ways. First, they more closely fit 

a normal rather than a lognormal distribution. Second, there was the potential for time dependence to 

interfere with the analysis. This was evident in the 1996 Building 65 measurements where it 

appeared that the WL monitor was started and the building ventilation was subsequently decreased. 

The first 22 measurements collected were the 22 lowest measurements, and they increased in an 

exponential fashion. These measurements were eliminated from the fit, which greatly improved it. 

The remaining points fit a normal distribution well, with an R-squared value of 0.978.

The summary of the WL measurements appears in Table B-8. 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Table B-8. Summary of WL measurements. 

Time and/or 

location

Number

of 

results Distribution

Mean, GM, or 

measurement 

(WL)

95th

percentile 

(WL) Reference Page(s)

1986 Plant 8 

silo & bins

13 Lognormal 0.0650 30.2 WMCO 1996 17, 63

1986 various 

locations

22 Lognormal 0.0217 0.133 Weaver 1987 4

1987 various 

locations

55 Lognormal 0.0117 0.180 Weaver 1987 4, 5

1990-1991 

Building 71 

(beforea)

85 Normal 0.108 Not 

applicable

FEMP 1996–1997 68

1991 Plant 9 182 Lognormal 0.023 Not 

applicable

Renk 1998 118

1992 Building 

3045

1 Not 

applicable

0.002 Not 

applicable

WMCO 1992b 3

1992 Building 

15 C-29A Cave

1 Not 

applicable

0.016 Not 

applicable

FMPC 1992 3

1992 Building

15 N-4

1 Not 

applicable

0.064 Not 

applicable

FMPC 1992 3

1993 Plant 1, 

Ore Silos BZ

1 Not 

applicable

0.394 Not 

applicable

FERMCO 1993a 3

1993 Plant 1, 

Ore Silos BZ

1 Not 

applicable

0.239 Not 

applicable

FERMCO 1993a 8

1993 Plant 1, 

Ore Silos BZ

1 Not 

applicable

3.1 Not 

applicable

FERMCO 1993a 10

1995 Building 

65 w/o 

ventilation

1 Not 

applicable

7 Not 

applicable

Toler 1996a 13

1995 Building 

65 w ventilation

1 Not 

applicable

0.02 Not 

applicable

Toler 1996a 13

1996 Building 

64 C.P. (1-wk)

168 Normal 0.1030 0.1631 Renk 1998 22–26

1996 Building 

64 N. (1-wk)

168 Normal 0.0599 0.0987 Renk 1998 45–49

1996 Building 

65 (8-day)

191 Normal 2.20 2.82 Renk 1998 51–56

1996 Building 

71 (after*)

18 Normal 0.007 0.011 FEMP 1996–1997 68

2002–2003 

Pilot Plant

34 Lognormal 0.0131 0.953 FEMP 1993–2004 175–240

a. ”Before” and “after” refer to sealing of the floor in Building 71.

The data in Table 2 were compared to both the 0.02 equilibrium calculation in Sections B.1 to B.4 and 

the 0.04 equilibrium calculation shown above. The results appear in Figure B-5. Most of the 

individual measurements, means or GMs, and 95th percentiles are less than the 0.53-WL estimate in 

Sections B.1 to B.4. The five exceptions are discussed individually below.

Exception 1: 1986 Plant 8 Silo and Bins (95th percentile = 30.2 WL)

Three of 13 measurements taken before the remediation of the Plant 8 Silos and Bins (aka Project 1) 

were greater than 0.53 WL (two were below detection). The three points were all significantly greater 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

than the remaining measurements, resulting in a GSD of 41.8 and a high 95th percentile. However, 

all three measurements were taken in areas not normally occupied, as shown in Table B-9. 

Table B-9. Three high measurements near Plant 

8 silo and bins in 1986. 

Location WL Measurement

Top of silo 17

Top of silo, upwind side 2.3

Top of silo, downwind side 71

The remaining measurements, excluding the ones shown as “none detected,” ranged from 0.0091 to 

0.44 WL, well within the 0.53 WL estimated in Sections B.1 to B.4. 

Exception 2: 1993 Plant 1, Ore Silos Breathing Zone (BZ) (Individual Measurement = 3.1 WL)

The Plant 1 Ore Silos Project involved the dismantling of 14 silos which had formerly contained cold 

raffinates from uranium concentrate processing (FERMCO 1995, p. 88). One of three samples from a 

BZ sampler recovered for the Plant 1 Ore Silos Project in 1993 was above the 0.53 WL estimate in 

Sections B.1 to B.4. Normally a BZ sample indicates that it was hands-on work with appropriate 

protections provided to the few employees involved. In this case, no monitored employee name was 

on the sample sheet which is unusual for a BZ sample. The location description was “Plant 1 ore 

Silos Project (Silo #14) 3’ inside from top.” It is likely that this was a BZ sampler being used as an 

area sampler and that this was not an occupied area.

Exception 3: 1995 Bldg 65, Without Ventilation (Individual Measurement = 7 WL)

Three WL values were reported were for Building 65 in 1995 for three corresponding ventilation 

conditions. Building 65 was the facility that contained 5,600 drums of thorium materials, some of 

which had deteriorated. The measurement quoted above was apparently the maximum of several 

measurements. The Health Physics Plan for the Thorium Overpacking Project states, “Baseline WLM 

measurements in Building 65 indicate that airborne radon/thoron progeny may approach 7 WLs” 

(Toler 1996b, p. 12). With 16,000 cubic feet per minute ventilation, the value quoted was less than 

0.02 WL. With three of the HEPA ventilation units running, the value quoted was 0.10 WL. The latter 

two values are less than the 0.53 WL used in Sections B.1 to B.4. In any case, the area was 

restricted and protections were in place for the relatively-small workforce that entered the facility for 

overpacking. Respiratory protection was required for building entry (Hazlewood 1995, p. 154). This 

would have been effective for the particulate daughters, although not for the thoron itself. Another 

1995 reference quotes WL values that vary anywhere from 2 WLs to 10 WLs but that are normally in 

the 3 to 6 WLs range. The numbers cited were for radon and thoron, and the reference notes that the 

WL monitors normally indicate 95% to 100% thoron (Allen 1995b, p. 155). Because the range from 2 

to 10 WL is consistent with the 7-WL measurement, it is assumed that these values also apply to 

static or no-ventilation conditions, and that the values were much lower when workers were present.

Exception 4: 1996 Bldg 65, Eight Days of Hourly Measurements by a WL Monitor (mean = 2.2 

WL)

This series of measurements is related to the 1995 Building 65 measurements cited above. As 

previously discussed, the measurements increased with time for the first 22 measurements (about 

one day). This seems to indicate some decrease in the ventilation. After equilibrium is achieved, the 

measurements are normally distributed about a mean of 2.2 WL. The same comments apply to these 

measurements as the 1995 Building 65 measurements; namely, the area was restricted, and 

protections were in place for the relatively-small workforce that entered the facility for overpacking. 

The fact that the levels increased to the range discussed in the previous paragraph probably indicates 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

that this measurement applies to static or no-ventilation conditions. Again, it is assumed that the 

values were much lower when workers were present.

Exception 5: 2002–2003 Pilot Plant (95th Percentile = 0.95 WL)

This series of measurements consisted of both high-volume and low-volume air samples. Almost all 

of the samples were collected for only five minutes. The measurements are in a document that is 

marked as RI/FS samples in support of the Pilot Plant demolition (FEMP 1993–2004, pp. 175, 177, 

179, 181, 183, 185, 187, 189, 191, 193, 195, 197, 199, 201, 203, 205, 207, 209, 211, 213, 215, 217, 

219, 221, 223, 225, 227, 229, 230, 232, 234, 236, 238, and 240). However, the 2002 to 2003 

radon/thoron samples are noted as being either weekly or monthly routine samples. The data did not 

fit either a normal or a lognormal distribution well. The high- and low-volume samples were also 

plotted and fitted separately. The low-volume samples provided a better lognormal fit, but the highvolume samples did not. The main issue with the high-volume samples was a set of three low outliers 

in the data, with a magnitude on the order of 1 × 105 WL. These three points were an order of 

magnitude lower than any of the other WL results reported. The best visual fit to a lognormal was 

obtained by eliminating these data, as well as two negative results and one illegible result. Instead, 

the somewhat arbitrary decision was made to treat these data as less than detection and include them 

in the fit. The resulting estimate of the 95th percentile exceeded the Sections B.1 to B.4 value of 0.53 

WL, but not the 1.07 WL value for a 0.04% equilibrium factor. None of the underlying data exceeded 

the Sections B.1 to B.4 value. 

B.5.5 Conclusion

With only one exception, the thoron measurements in 1996 through 1998 were far below the 

200 pCi/L estimated in Sections B.1 to B.4. The lone exception was for Building 65 for December 

1995 through March 1996. No additional details about this measurement were discovered. However, 

during the ALARA Design Review, a measurement of half that magnitude was considered to be 

typical for Building 65 under natural ventilation conditions (Allen 1995a, p. 31). The value used in the 

ALARA Design Review did not exceed 200 pCi/L.

In general, the thoron daughter measurements were below 0.53 WL. The exceptions were discussed 

in detail above and can be summarized as: being in Building 65, in areas not normally occupied (top 

of Plant 8 silos), or one of three BZ samples from the Plant 1 ore silos project which may have been 

an area sample rather than a BZ sample of an individual.

From the above, one can conclude that, compared to the obtained measurements, the thoron and 

thoron daughter concentration estimates in Sections B.1 to B.4 used to generate Tables B-5 and B-6 

are reasonably conservative.
ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

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Figure B-3. Thoron comparison, 1996 to 1998. 
ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Document No. ORAUT-TKBS-0017-5 Revision No. 02 Effective Date: 09/30/2016 Page 107 of 189

0.001

0.01

0.11

10

100

1000

-3-2

-1

01

2

3

pCi/L

z - Score

1996 Thoron Alpha Track Measurements

Annual Averages for 68 Locations ln(Data) Pred. GM= .565 GSD= 4.91

Uniform Pred. GM= .435 GSD= 5.83 Finney Pred. GM= .404 GSD= 4.94

Figure B-4. Measurements and lognormal fits, 1996. 
ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Document No. ORAUT-TKBS-0017-5 Revision No. 0

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Figure B-5. WL comparisons, 1986 through 2003. 1990 to 1991 measurements were made before 

189

sealing of the floor in Building 71; 1996 measurements were made after sealing. 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

B.5.6 1954 Thoron/Thoron Daughter Results 

Twenty-six (26) Plant 9 thoron/thoron daughter results for 1954 were located in NLO (1954). Samples 

were collected at six locations, as shown in Table B-10. 

Table B-10. Locations and numbers of samples. 

Location Number of samples

GA wet area–main floor 5

GA wet area–top deck 5

GA TNT & sump cake storage area 5

GA north end of rod storage area 5

GA center of rod storage area 5

GA south end of rod storage area 1

The samples were taken with the ventilation turned off to determine the thoron build-up. The time that 

the ventilation was turned off was noted as March 8, 1954 at 10:37 a.m. (NLO 1954, p. 13). The 

heating unit apparently remained on, recirculating room air. The samples were collected on March 8, 

1954 between 9:25 a.m. and 13:25 a.m.

The samples were counted either on March 8 or March 9 (“first count”). They were recounted 

(“second count”) on March 15. Since the first counts were taken between 3.75 and 19.58 hours after 

the sample was collected, the 220Rn and 216Po would have decayed, leaving only 212Pb and its 

daughters, including the alpha-emitters 212Bi and 212Po. Considering the branching ratios, the average 

alpha energy per emission is 7.8 MeV. This leads to the definition of a thoron WL as 7.5 pCi/L of 

thoron-daughter alpha activity. See Table B-3. 

The calculations for reproducing the net counts per minute shown on the sample sheets are 

straightforward. This quantity has to be divided by the counting efficiency and sample volume, which 

are shown on the sample sheet, as well as a factor of 0.70, to produce the concentration at the time of 

counting. The factor of 0.70 is not shown on the sample sheet. This factor was likely a correction for 

sample self-absorption of the alpha particles due to dust loading of the filter [see Boback (undated, 

pp. 39–40), although an absorption factor of 0.5 was used in that example]. It is clear that this was 

not a correction for decay from sample collection to counting since it was applied to both the first and 

second counts as well as other counts in NLO (1954) for long-lived activity. Most, but not all, of the 

samples had an additional correction applied to the first count which varied from 0.60 to 0.79. This 

correction was assumed to be a decay correction. Since the correction could not be reproduced and 

was not consistently applied, this analysis uses a recalculated decay correction based on the elapsed 

time between the sample collection and the count times and the 11-hour half-life of 212Pb. The second 

counts occurred between 167 and 169 hours after sample collection, or at least 15 212Pb half-lives 

after sample collection. Therefore, all of the thoron daughter activity had decayed, leaving only longlived isotopes, such as uranium and thorium. The difference between the first and second counts is 

due to the activity of 212Pb and its daughters. The decay correction was applied to the difference 

between the activity in the first count and the long-lived activity in the second count. This subtraction 

did not make a significant difference, which is probably why it was ignored in the site’s analysis. The 

concentrations determined by the above method were converted to pCi/L and then divided by 

7.5 pCi/L to determine the daughter concentration in WLs.

The results and their references are shown in Table B-11. 
ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Document No. ORAUT-TKBS-0017-5 Revision No. 0

2 Effective Date: 

09/30/2016 Page 110 of 189

Table B-11. Thoron and thoron daughter results (pCi/L). 

Sample

Date and time

of sample Location

Recalculated

concentration (dpm/m3) WL

Reported

thoron

Calculated

thoron

Page(s)

(NLO 1954)

002 3/8/1954 9:25 GA wet area–main floor 2,399 0.144 3.69 54.0 14, 69

003 3/8/1954 9:25 GA wet area–top deck 1,988 0.119 3.06 44.8 14, 69

004 3/8/1954 9:26 GA TNT & sump cake storage area 256 0.015 0.4 5.8 14, 69

005 3/8/1954 9:26 GA north end of rod storage area 647 0.039 1.05 14.6 14, 69

006 3/8/1954 9:26 GA center of rod storage area 386 0.023 0.63 8.7 14, 69

007 3/8/1954 9:27 GA south end of rod storage area 683 0.041 1.11 15.4 14, 69

008 3/8/1954 10:02 GA wet area–main floor 1,902 0.114 3.2 42.8 14, 69

009 3/8/1954 10:02 GA wet area–top deck 1,027 0.062 1.73 23.1 14, 69

010 3/8/1954 10:02 GA TNT & sump cake storage area 300 0.018 0.5 6.7 14, 69

011 3/8/1954 10:03 GA north end of rod storage area 499 0.030 0.79 11.2 14, 69

012 3/8/1954 10:03 GA center of rod storage area 557 0.033 0.94 12.6 14, 69

013 3/8/1954 10:40 GA wet area–main floor 1,930 0.116 4.32 43.5 12, 13

014 3/8/1954 10:40 GA wet area 1,153 0.069 2.53 26.0 12, 13

015 3/8/1954 10:40 GA TNT & sump cake storage area 827 0.050 1.81 18.6 12, 13

016 3/8/1954 10:39 GA north end of rod storage area 1,931 0.116 4.23 43.5 12, 13

017 3/8/1954 10:39 GA center of rod storage area 1,781 0.107 3.91 40.1 12, 13

018 3/8/1954 11:15 GA wet area–main floor 2,798 0.168 6.36 63.0 12, 13

019 3/8/1954 11:15 GA wet area 2,239 0.134 5.08 50.4 12, 13

020 3/8/1954 11:15 GA TNT & sump cake storage area 2,239 0.134 5.08 50.4 12, 13

021 3/8/1954 11:15 GA north end of rod storage area 2,034 0.122 4.62 45.8 12, 13

022 3/8/1954 11:15 GA center of rod storage area 2,800 0.168 6.36 63.1 12, 13

023 3/8/1954 13:25 GA wet area–main floor 10,123 0.608 20.78 228.0 16

024 3/8/1954 13:25 GA wet area–2nd deck 6,060 0.364 12.43 136.5 16

025 3/8/1954 13:25 GA TNT & sump cake storage area 1,617 0.097 3.44 36.4 16

026 3/8/1954 13:25 GA north end of rod storage area 2,258 0.136 4.74 50.9 16

027 3/8/1954 13:25 GA center of rod storage area 3,948 0.237 8.92 88.9 16
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

The results were plotted by location for the five locations with multiple samples. The plots are in 

Figures B-3 through B-10. The concentrations (in terms of WL) generally increased, as expected. 

The samples were collected from 9:30 a.m. to about 1:30 p.m. In one location (GA TNT and sump 

cake storage area), the concentration decreased between 11:30 a.m. and 1:30 p.m. Also shown on 

the plots is the 0.034 WL estimate from Sections B.1 to B.4. The first 11 sample times were before 

the ventilation was disrupted. These measurements averaged 0.058 ±0.05 WL (±1 standard 

deviation). If only the first measurement in each location is considered, the measurements averaged 

0.064 ±0.05 WL. It is not clear whether any additional measures were taken to maximize the effect of 

the reduced ventilation, such as closing windows and doors that would normally be open. It is also 

not clear whether the sample times were the beginning, middle, or end of the sample collection. For 

the recalculations, it was assumed that they were the end of the sample collection. Given this 

uncertainty, the results appear to be consistent with the Sections B.1 to B.4 value of 0.034 WL.

The sample sheets also reported concentrations of thoron gas in Ci/L. Because there were issues 

with the calculation of the concentrations of the daughters, the thoron concentrations based on them 

also had issues. Although the site’s calculated thoron concentrations were consistent with Sections 

B.1 to B.4, the equilibrium factors being assumed to determine the thoron concentrations were much 

higher than are realistic based on later research (see the response to Comment 3 (Equilibrium Factor) 

in this appendix). The thoron concentrations were recalculated from the recalculated WLs and the 

0.02 equilibrium factor used in Sections B.1 to B.4. The thoron concentrations were 21.8 ±17.2 pCi/L 

before the ventilation was turned off and 23.9 ±20.3 pCi/L for just the first six measurements. The 

12.8 pCi/L estimate in Sections B.1 to B.4 lies within one standard deviation of these results.

Figure B-6. GA center of rod storage area. 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Figure B-7. GA north end of rod storage area. 

Figure B-8. GA TNT and sump cake storage area. 
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ATTACHMENT B 

ANALYSIS OF POTENTIAL THORON EXPOSURES (continued)

Figure B-9. GA wet area, main floor. 

Figure B-10. GA wet area, top deck. Assumed to be the same location 

as “GA wet area” and “GA wet area – 2nd deck”. 
Document No. ORAUT-TKBS-0017-5 Revision No. 02 Effective Date: 09/30/2016 Page 114 of 189

 

ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA

TABLE OF CONTENTS

SECTION TITLE PAGE

C.1 Purpose ............................................................................................................................... 116

C.2 Data and Method Overview.................................................................................................. 116

C.2.1 Bioassay Data Selection........................................................................................... 116

C.2.2 Method Description................................................................................................... 116

C.3 Analysis ............................................................................................................................... 119

C.4 Intake Modeling.................................................................................................................... 120

C.4.1 Assumptions............................................................................................................. 120

C.4.2 Bioassay Fitting ........................................................................................................ 120

C.5 Assigning Intakes and Doses ............................................................................................... 120

C.5.1 Intake Rate Summary............................................................................................... 120

C.5.2 Dose Assignment ..................................................................................................... 121

C.6 Coworker Data Figures ........................................................................................................ 122

LIST OF TABLES 

TABLE TITLE PAGE

C-1 Criteria applied to extract for uranium bioassay data............................................................ 116

C-2 Reporting level..................................................................................................................... 117

C-3 Summary of uranium urinary excretion rate analyses, 1952 to 2006 .................................... 118

C-4 Derived uranium intake rates for type F materials, 1952 to 2006.......................................... 121

C-5 Derived uranium intake rates for type M materials, 1952 to 2006......................................... 121

C-6 Derived uranium intake rates for type S materials, 1952 to 2006.......................................... 122

LIST OF FIGURES 

FIGURES TITLE PAGE

C-1 Change in lognormal fit for various CLs................................................................................ 118

C-2 Predicted values versus measured results, 01/01/1952 to 12/31/1993, 50th percentile, 

type F................................................................................................................................... 123

C-3 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 50th percentile, 

type F................................................................................................................................... 123

C-4 Predicted values versus measured results, 01/01/1952 to 12/31/1993, 84th percentile, 

type F................................................................................................................................... 123

C-5 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 84th percentile, 

type F................................................................................................................................... 124

C-6 Predicted values versus measured results, 01/01/1952 to 12/31/1993, 50th percentile, 

type M.................................................................................................................................. 124
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

C-7 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 50th percentile, 

type M.................................................................................................................................. 124

C-8 Predicted values versus measured results, 01/01/1952 to 12/31/1993, 84th percentile, 

type M.................................................................................................................................. 125

C-9 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 84th percentile, 

type M.................................................................................................................................. 125

C-10 Predicted values versus measured results, 01/01/1952 to 12/31/1954, 50th percentile, 

type S................................................................................................................................... 125

C-11 Predicted values versus measured results, 01/01/1955 to 12/31/1955, 50th percentile, 

type S................................................................................................................................... 126

C-12 Predicted values versus measured results, 01/01/1956 to 12/31/1956, 50th percentile, 

type S................................................................................................................................... 126

C-13 Predicted values versus measured results, 01/01/1957 to 12/31/1961, 50th percentile, 

type S................................................................................................................................... 126

C-14 Predicted values versus measured results, 01/01/1962 to 12/31/1966, 50th percentile, 

type S................................................................................................................................... 127

C-15 Predicted values versus measured results, 01/01/1967 to 12/31/1993, 50th percentile, 

type S................................................................................................................................... 127

C-16 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 50th percentile, 

type S................................................................................................................................... 127

C-17 Summarized results of the predicted values versus the excluded results, 01/01/1952 to 

12/31/2006, 50th percentile, type S...................................................................................... 128

C-18 Predicted values versus measured results, 01/01/1952 to 12/31/1956, 84th percentile, 

type S................................................................................................................................... 128

C-19 Predicted values versus measured results, 01/01/1957 to 12/31/1961, 84th percentile, 

type S................................................................................................................................... 128

C-20 Predicted values versus measured results, 01/01/1962 to 12/31/1966, 84th percentile, 

type S................................................................................................................................... 129

C-21 Predicted values versus measured results, 01/01/1967 to 12/31/1980, 84th percentile, 

type S................................................................................................................................... 129

C-22 Predicted values versus measured results, 01/01/1981 to 12/31/1993, 84th percentile, 

type S................................................................................................................................... 129

C-23 Predicted values versus measured results, 01/01/1994 to 12/31/2006, 84th percentile, 

type S................................................................................................................................... 130

C-24 Summarized results of the predicted values versus the excluded results, 01/01/1952 to 

12/31/2006, 84th percentile, type S ...................................................................................... 130
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

C.1 PURPOSE

Some employees at DOE sites were not monitored for internal ionizing radiation exposure, or the 

monitoring records are incomplete or unavailable. In such cases, data from monitored coworkers can 

be used to estimate a worker’s possible exposure. The purpose of this attachment is to provide 

monitored coworker information for calculating and assigning occupational internal doses to 

employees at the Fernald site for whom there are no or insufficient bioassay monitoring records.

C.2 DATA AND METHOD OVERVIEW

ORAUT-OTIB-0019, Analysis of Coworker Bioassay Data for Internal Dose Assignment (ORAUT 

2005b), describes the general process that is used to analyze bioassay data for assigning doses to 

individuals based on coworker results. ORAUT-PLAN-0014, Coworker Data Exposure Profile 

Development (ORAUT 2004), describes the approach and processes to be used to develop 

reasonable exposure profiles based on available dosimetric information for workers at DOE sites.

C.2.1 Bioassay Data Selection

The bioassay results for this analysis were obtained by extracting data from HIS-20_ORAU, a 

database of Fernald bioassay results (ORAUT 2007). One of the database tables, 

HIS20_V_BIOASSAY, contains results for urinalyses, fecal analyses, and in vivo analyses starting in 

1952. The radionuclides in the table include 228Ac, 36Cl, 137Cs, 3

H, 210/211Pb, 210Po, 239Pu, 226/228Ra, 

90Sr, 228/230/232Th, 234/235/238U, and total uranium. The vast majority of data in the table concerns the 

uranium radionuclides. Data were extracted from the HIS20_V_BIOASSAY table using the field 

criterion in Table C-1.

Table C-1. Criteria applied to extract for uranium bioassay data. 

Field name Field value

TYPE_BIOASSAY URINALYSIS

NUC_NAME U-TOTAL or 238U

SAMPLE_TYPE Not equal to 10, 5C, 70, VF, VR, VE

ACT_UNITS_SU μg/L

Samples with code 10 and 70 were collected before employment or reemployment to establish the 

background for the individual. As such, these samples are not appropriate for use in a study to 

establish intake potential. Samples with code 5C were collected to test for possible correlation with 

abnormal clinical laboratory findings. VF, VR, and VE are visitor samples and likely not representative 

of the typical worker (ORAUT 2004). All results in this series were less than the MDA. Of the 

403,159 uranium urinalysis results not excluded as discussed above, 144 had results of “N/A”; these 

were discarded. The remaining 403,015 sample results were evaluated using the time-weighted one 

person–one sample (TWOPOS) methodology (ORAUT 2014c), which yielded 86,964 OPOS results.

C.2.2 Method Description

The reporting methods for bioassay results changed as a function of time. Before 1986 and after 

1997, the data appear to be uncensored. That is, the analytical result is reported regardless of 

magnitude. From 1986 through 1993, the minimum reported value incrementally changed from 3 μg/L 

to a maximum of 14 μg/L. Table C-2 lists the reporting levels as a function of time. Part of this 

change, and especially for 1991 through 1993, appears to be a refinement of the quantification of the 

detection capabilities of the methods and a change from reporting all results to results at or above the 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Table C-2. Reporting level. 

From To CL (μg/L)

Earliest 01/28/86 Uncensored

01/29/86 12/30/88 3

01/01/89 09/17/90 5

09/18/90 10/31/90 8

11/01/90 02/08/91 9

02/09/91 03/04/91 8

03/05/91 04/15/91 7

04/16/91 06/26/91 8

06/27/91 08/03/91 9

08/04/91 12/11/91 10

12/12/91 01/08/92 9

01/09/92 02/12/92 10

02/13/92 04/22/92 11

04/23/92 08/27/92 12

08/28/92 11/30/93 14

12/01/93 06/02/97 0.8

06/03/97 End Uncensored

DL and then to reporting only results above the MDA. This censoring of results below a given level 

skews the results of the calculated fits, which results in values that are biased high. Figure C-1 

depicts how recalculating the results for 1986 with varying censoring levels (CLs) alters the regression 

on order statistics (ROS) fits.3

 The original censoring level (lowest line) was 3 μg/L.

The higher results in 1991 through 1993 are believed to be a result of the higher CLs in those years. 

In 1992 and 1993, over 99% of the results are censored. In 1991, there appear to have been at least 

seven censoring levels that varied with time, which generates uncertainty about what results are 

censored. The results for all three of these years are inconsistent with the results for years before or 

after this period. During this period, Fernald ceased uranium production and began the process of 

site closure. Therefore, the potential for exposure was lower due to the lower production levels and 

the fact that site closure had not begun. Therefore, the results for 1991 through 1993 have been 

excluded from the intake modeling.

The excretion rates for each sample were normalized to represent 24-hour samples assuming 

1,400 mL/d urine excretion, which is the daily volume that is excreted by Reference Man in 

International Commission on Radiological Protection (ICRP) Publication 23 (ICRP 1975). Then the 

intakes that would result in the observed excretion rates were inferred using IMBA as described in 

Section C.4.

The uranium in urine bioassay data from the HIS-20_ORAU database were chronologically grouped 

into annual intervals by calendar year. The data in each year were fit to a lognormal distribution as 

described in Section 3.0. The analysis of the fits resulted in calculated excretion rates at the 

50th- and 84th-percentile values at the midpoint date of the analysis interval as shown in Table C-3.

 

3 Note that this figure was produced using the OPOS method from Revision 01 of ORAUT-RPRT-0053,

Analysis of Stratified Coworker Datasets, analyzed in Revision 02 of this document, whereas the current 

revision of this document used Revision 02 of ORAUT-RPRT-0053 (ORAUT 2014c) to calculate the actual 

intakes. However, the concept being depicted is the same regardless.
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Figure C-1. Change in lognormal fit for various CLs. 

Table C-3. Summary of uranium urinary excretion rate 

analyses, 1952 to 2006. 

Year

Effective 

bioassay date

Number of 

TWOPOS 

results

50th 

percentile 

(μg/d)

84th 

percentile 

(μg/d)

1952 07/01/1952 71 28.81 107.07

1953 07/01/1953 701 19.60 57.07

1954 07/01/1954 1,376 26.11 71.47

1955 07/01/1955 1,973 45.78 117.20

1956 07/01/1956 2,497 26.11 66.50

1957 07/01/1957 2,937 17.88 46.03

1958 07/01/1958 2,485 12.86 29.79

1959 07/01/1959 2,540 13.97 28.64

1960 07/01/1960 2,630 18.89 33.88

1961 07/01/1961 2,395 18.18 32.30

1962 07/01/1962 2,131 12.54 23.93

1963 07/01/1963 1,983 13.27 26.42

1964 07/01/1964 1,900 10.92 24.05

1965 07/01/1965 1,663 9.01 20.22

1966 07/01/1966 1,484 8.24 18.46

1967 07/01/1967 1,602 6.96 14.94

1968 07/01/1968 1,398 6.14 13.35

1969 07/01/1969 1,281 5.48 12.30
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Year

Effective 

bioassay date

Number of 

TWOPOS 

results

50th 

percentile 

(μg/d)

84th 

percentile 

(μg/d)

1970 07/01/1970 1,119 4.86 10.15

1971 07/01/1971 881 5.27 11.67

1972 07/01/1972 634 5.01 10.41

1973 07/01/1973 735 5.23 11.07

1974 07/01/1974 678 5.36 11.07

1975 07/01/1975 697 5.23 10.37

1976 07/01/1976 697 5.22 10.24

1977 07/01/1977 664 5.15 9.62

1978 07/01/1978 644 5.55 10.03

1979 07/01/1979 599 5.13 9.75

1980 07/01/1980 893 5.04 11.37

1981 07/01/1981 623 2.93 7.26

1982 07/01/1982 262 5.70 10.44

1983 07/01/1983 785 3.86 8.71

1984 07/01/1984 696 5.01 11.00

1985 07/01/1985 858 4.70 8.50

1986 07/01/1986 1,565 4.01 6.16

1987 07/01/1987 1,611 2.91 5.50

1988 07/01/1988 1,609 3.53 5.24

1989 07/01/1989 2,029 1.79 3.64

1990 07/01/1990 2,044 5.12 7.36

1991 07/01/1991 2,163 12.39 12.95

1992 07/01/1992 2,803 11.79 13.64

1993 07/01/1993 3,166 10.81 12.79

1994 07/01/1994 2,817 0.097 0.293

1995 07/01/1995 2,901 0.105 0.317

1996 07/01/1996 2,298 0.582 0.951

1997 07/01/1997 2,159 0.187 0.570

1998 07/01/1998 2,382 0.025 0.070

1999 07/01/1999 2,351 0.029 0.071

2000 07/01/2000 2,076 0.039 0.100

2001 07/01/2001 1,809 0.070 0.147

2002 07/01/2002 1,835 0.063 0.128

2003 07/01/2003 1,787 0.063 0.135

2004 07/01/2004 1,217 0.065 0.123

2005 07/01/2005 1,040 0.089 0.170

2006 07/01/2006 790 0.082 0.162

C.3 ANALYSIS

Bioassay data statistics were generated for each analysis interval using the methods in ORAUT 

(2014c). A lognormal distribution was assumed [9]. After log-transforming the data, the 50th- and 

84th-percentile values were determined for each period using the methods in ORAUT (2014c). For 

years with uncensored data, values less than or equal to zero were treated as being censored at the 

lowest positive value in that year for TWOPOS implementation. However, values equal to the lowest 

positive value are still considered to be uncensored. Table C-3 shows the statistical analysis results 

for uranium urinary excretion parameters.
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ATTACHMENT C

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

C.4 INTAKE MODELING

This section discusses intake modeling assumptions and intake fitting for three different material types 

of uranium compounds.

C.4.1 Assumptions

Each result in the intake calculations was assumed to have normal distribution [10]. A uniform 

absolute error of 1 was applied to all results to assign the same weight to each result. Because of the 

nature of work at Fernald, it is possible that intakes could have been either chronic or acute. 

However, a series of acute intakes can be approximated as a chronic intake. Therefore, intakes were 

assumed to be chronic and to occur through inhalation using a default breathing rate of 1.2 m3

/hr and 

a particle size distribution of 5-μm activity median aerodynamic diameter (ICRP 1995).

C.4.2 Bioassay Fitting

IMBA was used to fit the bioassay results to a series of inhalation intakes. Data from 1952 through 

2006 were fit as a series of chronic intakes. The intake assumptions were based on observed 

patterns in the bioassay data. Periods with constant chronic intake rates were chosen by selecting 

periods during which the bioassay results were similar. A new chronic intake period was started if the 

data indicated a significant and sustained change in the bioassay results. By this method, the period 

from 1952 through 2006 was divided into multiple chronic intake periods.

Because the uranium isotopes at Fernald have long radiological half-lives and the material is retained 

in the body for long periods, excretion results are not independent. For example, an intake in the 

1950s could contribute to urinary excretion in the 1980s and later. To avoid potential underestimation 

of intakes for people who worked at Fernald for relatively short periods, each chronic intake was fit 

independently using only the bioassay results from the single intake period for type S solubility. For 

solubility types M and F, this approach was used where it was determined that earlier intake rates 

significantly biased later intake rates (i.e., the period from 1994 through 2006 was evaluated 

separately from earlier periods). This method results in a potential overestimate of intakes for 

exposures that extend through multiple assumed intake periods. Uranium urinalysis results were 

analyzed with IMBA to derive intake rates for 1952 to 2006. Section C.6 contains the plots that 

compare predicted uranium bioassay results based on IMBA-derived uranium intake rates with the 

measured urine results.

C.5 ASSIGNING INTAKES AND DOSES

This section describes the derived intake rates and provides guidance for assigning doses. For each 

intake period below, the geometric standard deviations (GSDs) were determined by dividing the 

84th-percentile intake rates by the 50th-percentile rates. For the calculation of doses to individuals 

from bioassay data, a GSD of 3 is used to account for biological variation and uncertainty in the 

models (ORAUT 2014a). The same models are used for fitting the coworker data, so the same 

uncertainty applies. Therefore, a minimum GSD of 3 was assigned for each of the intake periods.

C.5.1 Intake Rate Summary

Multiple intake periods were fit to the derived 50th- and 84th-percentile uranium excretion data. 

Table C-4 summarizes the 50th- and 95th-percentile uranium intake rates that correspond to an intake 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Table C-4. Derived uranium intake rates for type F materials, 1952 to 2006. 

Dates

50th 

percentile 

(μg/d)

50th 

percentile 

(pCi/d)a GSD

95th 

percentile 

(μg/d)

95th 

percentile 

(pCi/d)a

01/01/52–12/31/54 91.5 89.3 3.36 673 656

01/01/55–12/31/55 169 164.5 3.00 1,027 1,002

01/01/56–12/31/56 94.3 92.0 3.26 659 644

01/01/57–12/31/61 58.6 57.2 3.00 357 349

01/01/62–12/31/64 38.2 37.3 3.00 233 227

01/01/65–12/31/66 38.2 61.7 3.00 233 376

01/01/67–12/31/93 16.7 27.0 3.00 102 164

01/01/94–12/31/06 0.418 0.675 3.00 2.55 4.12

a. The specific activities used to compute this column were 0.976 pCi/μg (1% EU) for 1952 through 

1964, and 1.616 pCi/μg (2% EU) for 1965 to the present.

of type F materials that was inferred from the excretion rates. The 95th-percentile intake rates were 

calculated using the following equation:

1 645 - - . 95th percentile intake 50th percentile intake = × GSD (C-1)

Table C-5 lists the same information for type M materials, and Table C-6 lists the information for 

type S materials. Periods with the same intake rate and GSD were combined for clarity. For periods 

after 2006 during which intakes were feasible, dose reconstructors should assume the 2006 intake 

rates.

Table C-5. Derived uranium intake rates for type M materials, 1952 to 2006. 

Dates

50th 

percentile 

(μg/d)

50th

percentile

(pCi/d)a GSD

95th 

percentile 

(μg/d)

95th 

percentile 

(pCi/d)a

01/01/52–12/31/54 392 382 3.29 2,779 2,712

01/01/55–12/31/55 744 726 3.00 4,531 4,422

01/01/56–12/31/56 328 320 3.93 3,117 3,042

01/01/57–12/31/61 236 230 3.00 1,435 1,401

01/01/62–12/31/64 153 149 3.00 933 911

01/01/65–12/31/66 153 247 3.00 933 1,508

01/01/67–12/31/93 67.5 109 3.00 411 665

01/01/94–12/31/06 1.73 2.79 3.00 10.5 17.0

a. The specific activities used to compute this column were 0.976 pCi/μg (1% EU) for 1952 through 

1964, and 1.616 pCi/μg (2% EU) for 1965 to the present.

C.5.2 Dose Assignment

For most cases, individual doses should be calculated from the 50th-percentile intake rates. For 

cases where there is justification that the individual might have had larger intakes than the 

50th-percentile intake rates, dose reconstructors should use the 95th-percentile intake rates 

applicable to the solubility class of the material from Table C-4, C-5, or C-6 as appropriate. Dose 

reconstructors should select the material type that is the most favorable to claimants and apply other 

RU components based on information in Section 5.3.2.5. 

Select the lognormal distribution in IREP with the calculated dose entered as Parameter 1 and the 

associated GSD as Parameter 2. The GSD relates to the intake, so apply it to all annual doses that 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Table C-6. Derived uranium intake rates for type S materials, 1952 to 2006. 

Dates

50th 

percentile 

(μg/d)

50th

percentile

(pCi/d)a GSD

95th 

percentile 

(μg/d)

95th 

percentile 

(pCi/d)a

01/01/52–12/31/54 8,611 8,404 3.00 52,471 51,212

01/01/55–12/31/55 25,755 25,137 3.00 156,937 153,171

01/01/56–12/31/56 14,660 14,308 3.00 89,330 87,186

01/01/57–12/31/61 4,691 4,578 3.00 28,584 27,898

01/01/62–12/31/64 2,925 2,855 3.00 17,823 17,396

01/01/65–12/31/66 2,925 4,727 3.00 17,823 28,803

01/01/67–12/31/93 827.1 1,337 3.00 5,040 8,144

01/01/94–12/31/06 21.2 34.3 3.00 129 209

a. The specific activities used to compute this column were 0.976 pCi/μg (1% EU) for 1952 through 1964, and 1.616 

pCi/μg (2% EU) for 1965 to the present.

are determined from the intake period. If used, assign the 95th-percentile intakes as a constant 

distribution.

C.6 COWORKER DATA FIGURES 

This section shows comparisons of measured uranium urine bioassay results with predicted results 

that were calculated using IMBA-derived uranium intake rates. Blue dots represent the measured 

values that were retained for the fit. Red dots represent results that were excluded because they 

were outside the intake period being fit. The green lines represent the predicted values. The figures 

provide the fits as follows:

• Figures C-2 and C-3 show the individual fits to the 50th-percentile excretion rates for type F 

material.

• Figures C-4 and C-5 show the individual fits to the 84th-percentile excretion rates for type F 

material.

• Figures C-6 and C-7 show the individual fits to the 50th-percentile excretion rates for type M 

material.

• Figures C-8 and C-9 show the individual fits to the 84th-percentile excretion rates for type M 

material.

• Figures C-10 through C-16 show the individual fits to the 50th-percentile excretion rates for 

type S material. Figure C-17 summarizes the results for the period from 1952 through 2006.

• Figures C-18 through C-23 show the individual fits to the 84th-percentile excretion rates for 

type S material. Figure C-24 summarizes the type S results for the period from 1952 through 

2006.
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Figure C-2. Predicted values (line) versus measured results (blue dots), 01/01/1952 to 12/31/1993, 

50th percentile, type F. 

Figure C-3. Predicted values (line) versus measured results (blue dots), 01/01/1994 to 12/31/2006, 

50th percentile, type F. 

Figure C-4. Predicted values (line) versus measured results (blue dots), 01/01/1952 to 12/31/1993, 

84th percentile, type F. 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Figure C-5. Predicted values (line) versus measured results (blue dots), 01/01/1994 to 12/31/2006, 

84th percentile, type F. 

Figure C-6. Predicted values (line) versus measured results (blue dots), 01/01/1952 to 12/31/1993, 

50th percentile, type M. 

Figure C-7. Predicted values (line) versus measured results (blue dots), 01/01/1994 to 12/31/2006, 

50th percentile, type M. 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Figure C-8. Predicted values (line) versus measured results (blue dots), 01/01/1952 to 12/31/1993, 

84th percentile, type M. 

Figure C-9. Predicted values (line) versus measured results (blue dots), 01/01/1994 to 12/31/2006, 

84th percentile, type M. 

Figure C-10. Predicted values (line) versus measured results (blue dots), 01/01/1952 to 12/31/1954, 

50th percentile, type S. 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Figure C-11. Predicted values (line) versus measured results (blue dots), 01/01/1955 to 12/31/1955, 

50th percentile, type S. 

Figure C-12. Predicted values (line) versus measured results (blue dots), 01/01/1956 to 12/31/1956, 

50th percentile, type S. 

Figure C-13. Predicted values (line) versus measured results (blue dots), 01/01/1957 to 12/31/1961, 

50th percentile, type S. 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Figure C-14. Predicted values (line) versus measured results (blue dots), 01/01/1962 to 12/31/1966, 

50th percentile, type S. 

Figure C-15. Predicted values (line) versus measured results (blue dots), 01/01/1967 to 12/31/1993, 

50th percentile, type S. 

Figure C-16. Predicted values (line) versus measured results (blue dots), 01/01/1994 to 12/31/2006, 

50th percentile, type S. 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Figure C-17. Summarized results of the predicted values (line) versus the excluded results (red dots), 

01/01/1952 to 12/31/2006, 50th percentile, type S. 

Figure C-18. Predicted values (line) versus measured results (blue dots), 01/01/1952 to 12/31/1956, 

84th percentile, type S. 

Figure C-19. Predicted values (line) versus measured results (blue dots), 01/01/1957 to 12/31/1961, 

84th percentile, type S. 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Figure C-20. Predicted values (line) versus measured results (blue dots), 01/01/1962 to 12/31/1966, 

84th percentile, type S. 

Figure C-21. Predicted values (line) versus measured results (blue dots), 01/01/1967 to 12/31/1980, 

84th percentile, type S. 

Figure C-22. Predicted values (line) versus measured results (blue dots), 01/01/1981 to 12/31/1993, 

84th percentile, type S. 
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ATTACHMENT C 

INTERNAL DOSIMETRY URANIUM COWORKER DATA (continued)

Figure C-23. Predicted values (line) versus measured results (blue dots), 01/01/1994 to 12/31/2006, 

84th percentile, type S. 

Figure C-24. Summarized results of the predicted values (line) versus the excluded results (red dots), 

01/01/1952 to 12/31/2006, 84th percentile, type S. 
Document No. ORAUT-TKBS-0017-5 Revision No. 02 Effective Date: 09/30/2016 Page 131 of 189

 

ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA

TABLE OF CONTENTS

SECTION TITLE PAGE

D.1 Manipulation of Fernald Chest Count Data........................................................................... 133

D.2 Discussion of Actinium-228 and Lead-212 in Chest Counts ................................................. 134

D.3 Bias and Decision Level Plots .............................................................................................. 135

D.4 Chest Burden Plots .............................................................................................................. 135

LIST OF TABLES 

TABLE TITLE PAGE

D-1 Summary of bias and DL for 228Ac and 212Pb for each year .................................................. 134

LIST OF FIGURES 

FIGURES TITLE PAGE

D-1 Normal probability plot of 212Pb chest count results for 1978 ................................................ 133

D-2 Normal probability plot of 212Pb chest count results for 1978 after adjusting for bias............. 133

D-3 Normal probability plot of 212Pb chest count results for 1978 after adjusting for bias............. 134

D-4 Plot of 228Ac versus 212Pb for 1978 where the levels of both nuclides are above the DL ..... 135

D-5 Normal probability plots of 212Pb and 228Ac, 1978 ................................................................. 136

D-6 Normal probability plots of 212Pb and 228Ac, 1979 ................................................................. 137

D-7 Normal probability plots of 212Pb and 228Ac, 1980 ................................................................. 138

D-8 Normal probability plots of 212Pb and 228Ac, 1981 ................................................................. 139

D-9 Normal probability plots of 212Pb and 228Ac, 1982 ................................................................. 140

D-10 Normal probability plots of 212Pb and 228Ac, 1983 ................................................................. 141

D-11 Normal probability plots of 212Pb and 228Ac, 1984 ................................................................. 142

D-12 Normal probability plots of 212Pb and 228Ac, 1985 ................................................................. 143

D-13 Normal probability plots of 212Pb and 228Ac, 1986 ................................................................. 144

D-14 Normal probability plots of 212Pb and 228Ac, 1987 ................................................................. 145

D-15 Normal probability plots of 212Pb and 228Ac, 1988 ................................................................. 146

D-16 Both radionuclides above the DL, 1978................................................................................ 147

D-17 Either radionuclides above the DL, 1978.............................................................................. 148

D-18 Both radionuclides above the DL, 1979................................................................................ 149

D-19 Either radionuclides above the DL, 1979.............................................................................. 150

D-20 Both radionuclides above the DL, 1980................................................................................ 151

D-21 Either radionuclides above the DL, 1980.............................................................................. 152

D-22 Both radionuclides above the DL, 1981................................................................................ 153

D-23 Either radionuclides above the DL, 1981.............................................................................. 154

D-24 Both radionuclides above the DL, 1982................................................................................ 155

D-25 Either radionuclides above the DL, 1982.............................................................................. 156

D-26 Both radionuclides above the DL, 1983................................................................................ 157
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

D-27 Either radionuclides above the DL, 1983.............................................................................. 158

D-28 Both radionuclides above the DL, 1984................................................................................ 159

D-29 Either radionuclides above the DL, 1984.............................................................................. 160

D-30 Both radionuclides above the DL, 1985................................................................................ 161

D-31 Either radionuclides above the DL, 1985.............................................................................. 162

D-32 Both radionuclides above the DL, 1986................................................................................ 163

D-33 Either radionuclides above the DL, 1986.............................................................................. 164

D-34 Both radionuclides above the DL, 1987................................................................................ 165

D-35 Either radionuclides above the DL, 1987.............................................................................. 166

D-36 Both radionuclides above the DL, 1988................................................................................ 167

D-37 Either radionuclides above the DL, 1988.............................................................................. 168
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

D.1 MANIPULATION OF FERNALD CHEST COUNT DATA

The 228Ac and 212Pb results from chest counts at Fernald from 1978 to 1988 were analyzed separately 

by year. Based on previous experience, the distributional model for these annual data is that they are 

a combination of a normal analytical background distribution with a mean of zero and a lognormal 

exposure distribution (ORAUT 2009). The normal component of the data is estimated by fitting a line 

through the negative data on a normal probability plot. The intercept of the line is the estimate of the 

mean or median of the normal distribution, and the slope is the estimate of the standard deviation. If 

the median of the distribution is not equal to zero, the data are assumed to be biased. For example, 

the normal probability plot for the 212Pb chest count data from 1978 are shown below in Figure D-1. 

The median of the data is –0.019 nCi, which means that the data are assumed to be biased by that 

much. The data are adjusted for any bias and then refit to force the line through zero (i.e., make the 

median equal to zero). For example, the adjusted 1978 212Pb data are shown below in Figure D-2.

-3 -2 -1 0 1 2 3

-0.5 0.0 0.5 1.0

Standard Normal Quantiles

Pb-212 (nCi)

Bias = -0.019 nCi

Figure D-1. Normal probability plot of 212Pb chest count 

results for 1978. The line is a robust regression of the data 

that are less than zero nCi.

-3 -2 -1 0 1 2 3

-0.5 0.0 0.5 1.0

Standard Normal Quantiles

Pb-212 (nCi)

Bias = 0 nCi

Figure D-2. Normal probability plot of 212Pb chest count 

results for 1978 after adjusting for bias. The line is a robust 

regression of the data that are less than zero nCi.

The process decision level (DL) above which a measured chest burden is deemed to be “positive” is 

taken to be the 95th percentile of the (unbiased) normally distributed analytical background. For the 

1978 212Pb data, the DL is 0.12 nCi as shown in Figure D-3. Measured 212Pb chest burdens below 

this DL are deemed to not indicate the presence of 212Pb.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

-3 -2 -1 0 1 2 3

-0.5 0.0 0.5 1.0

Standard Normal Quantiles

Pb-212 (nCi)

DL = 0.12 nCi

Figure D-3. Normal probability plot of 212Pb chest count 

results for 1978 after adjusting for bias. The line is a robust 

regression of the data that are less than 0 nCi, and the red 

point is the 95th percentile of the background normal 

distribution (i.e., the DL).

A summary of the bias and DL for 228Ac and 212Pb for 1978 through 1988 are given in Table D-1, and 

the corresponding plots are given in Section D.3. The DLs for 212Pb ranged from 0.11 to 0.18 nCi with 

a mean of 0.13 nCi, whereas the DL for 228Ac ranged from 0.08 to 0.17 nCi with a mean of 0.11 nCi. 

For simplicity, a single DL of 0.12 nCi has been adopted for both nuclides for all years. If a minimum 

detectable amount is needed for dose reconstruction purposes, a value of 2 times the DL, 0.24 nCi, 

should be used for both nuclides.

Table D-1. Summary of bias and DL for 228Ac and 212Pb for each year.

Year

Pb-212 bias 

(nCi)

Pb-212

DL (nCi)

Ac-228 bias 

(nCi)

Ac-228

DL (nCi)

1978 –0.019 0.120 0.008 0.153

1979 –0.014 0.156 0.031 0.169

1980 –0.045 0.123 0.025 0.091

1981 –0.056 0.108 0.022 0.127

1982 –0.058 0.121 0.010 0.117

1983 –0.101 0.122 –0.008 0.097

1984 –0.089 0.111 0.001 0.079

1985 –0.086 0.117 0.017 0.102

1986 –0.085 0.128 0.020 0.114

1987 –0.047 0.123 0.008 0.106

1988 0.002 0.181 0.058 0.107

D.2 DISCUSSION OF ACTINIUM-228 AND LEAD-212 IN CHEST COUNTS

In Figure D-4, chest burdens from 1978, where both 228Ac and 212Pb are greater than the DL, are 

presented (228Ac/

212Pb plots for all years are presented in Section D.4). These data are considered to 

be above the level of analytical noise and represent actual material in the chest.

4

 The diagonal blue 

line has a slope of 1 and an intercept of 0. Data falling exactly on this line have equal quantities of 

 

4 If the 228Ac to 212Pb ratio included the results that are basically noise, there would be quite a range of ratios 

beyond the anticipated 1:1 that would be very difficult to interpret in a meaningful way.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

228Ac and 212Pb. All of these measurements, except one, are considered to indicate equal quantities 

of 228Ac and 212Pb in the chest (to within measurement uncertainty).

-1 0 1 2 3 4 5

-1 0 1 2 3 4 5

 

Pb212

Ac228

Figure D-4. Plot of 228Ac versus 212Pb for 1978 where the 

levels of both nuclides are above the DL. The blue line has a 

slope of 1 and an intercept of 0.

The lone outlier has an 228

Ac/

212Pb ratio of about 18. The only plausible explanation for such a large 

ratio is the presence of unsupported 228Ra. This suggests that the 228Ac activity in excess of the 212Pb

activity should be interpreted as a 228Ra intake and assessed separately from the 232Th/228Th intake 

estimated from the 212Pb. Acute and chronic intakes of 228Ra measured with 228Ac are readily 

evaluated with IMBA and do not require custom treatment.

D.3 BIAS AND DECISION LEVEL PLOTS

Figures D-5 to D-15 plot bias and decision level. Each figure contains two plots for each year from 

1978 to 1988. The first is a normal probability plot of the 212Pb data, and the second is that for 228Ac. 

The median of the 212Pb data was estimated from a robust regression to the negative (less than zero) 

data. The median, given as Bias on the plots, was then subtracted from each result. The data on the 

plots are bias-corrected. The regression was repeated on the bias-corrected data to obtain the 

standard deviation and mean of the normal background distribution. These parameters were then 

used to calculate the DL, which is assigned the 95th percentile of the normal background data.

D.4 CHEST BURDEN PLOTS

Figures D-16 to D-26 plot chest burden data by year. Each figure contains two plots, the first on a 

scale of 0 to 10, the second on a scale of 0 to 1: 

• The first figure for each year contains two plots of 228Ac versus the 212Pb where both nuclides 

are above the DL.

• The second for each year contains two plots of 228Ac versus the 212Pb where either nuclide is 

above the DL.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-5. Normal probability plots of 212Pb and 228Ac, 1978.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-6. Normal probability plots of 212Pb and 228Ac, 1979.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-7. Normal probability plots of 212Pb and 228Ac, 1980.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-8. Normal probability plots of 212Pb and 228Ac, 1981.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-9. Normal probability plots of 212Pb and 228Ac, 1982.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-10. Normal probability plots of 212Pb and 228Ac, 1983.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-11. Normal probability plots of 212Pb and 228Ac, 1984.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-12. Normal probability plots of 212Pb and 228Ac, 1985.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-13. Normal probability plots of 212Pb and 228Ac, 1986.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-14. Normal probability plots of 212Pb and 228Ac, 

1987.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-15. Normal probability plots of 212Pb and 228Ac, 1988.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

1978 (Ac and Pb Results > 0.12 nCi)

Figure D-16. Both radionuclides above the DL, 1978.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-17. Either radionuclide above the DL, 1978.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-18. Both radionuclides above the DL, 1979.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-19. Either radionuclide above the DL, 1979.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-20. Both radionuclides above the DL, 1980. 
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-21. Either radionuclide above the DL, 1980.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-22. Both radionuclides above the DL, 1981.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-23. Either radionuclide above the DL, 1981.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-24. Both radionuclides above the DL, 1982.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-25. Either radionuclide above the DL, 1982.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-26. Both radionuclides above the DL, 1983.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-27. Either radionuclide above the DL, 1983.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-28. Both radionuclides above the DL, 1984.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-29. Either radionuclide above the DL, 1984.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-30. Both radionuclides above the DL, 1985.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-31. Either radionuclide above the DL, 1985.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-32. Both radionuclides above the DL, 1986.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-33. Either radionuclide above the DL, 1986. 
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-34. Both radionuclides above the DL, 1987.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-35. Either radionuclide above the DL, 1987.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-36. Both radionuclides above the DL, 1988.
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ATTACHMENT D 

EVALUATION OF ACTINIUM-228/LEAD-212 CHEST COUNT DATA (continued)

Figure D-37. Either radionuclide above the DL, 1988. 
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ATTACHMENT E 

TECHNICAL BASIS FOR ASSIGNING DOSES FROM

EFFECTIVE DERIVED AIR CONCENTRATION BREATHING ZONE RESULTS

TABLE OF CONTENTS

SECTION TITLE PAGE

E.1 Background.......................................................................................................................... 170

E.2 Affected Locations................................................................................................................ 170

E.2.1 BL-13........................................................................................................................ 170

E.2.2 BL-65........................................................................................................................ 171

E.2.3 KS-65 ....................................................................................................................... 171

E.2.4 CELL-8 ..................................................................................................................... 173

E.2.5 RT-210 ..................................................................................................................... 173

LIST OF TABLES 

TABLE TITLE PAGE

E-1 Count of EDAC results in HIS-20 ......................................................................................... 170

E-2 Activity fractions for the BL-13 EDAC................................................................................... 171

E-3 Isotopic composition of K-65 silo 1....................................................................................... 172

E-4 Isotopic composition of K-65 silo 2....................................................................................... 172

E-5 Activity fractions for the RT-210 EDAC................................................................................. 174

LIST OF FIGURES 

FIGURES TITLE PAGE

E-1 222Rn decay chain showing half-lives.................................................................................... 174
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ATTACHMENT E 

TECHNICAL BASIS FOR ASSIGNING DOSES FROM

EFFECTIVE DERIVED AIR CONCENTRATION BREATHING ZONE RESULTS (continued)

E.1 BACKGROUND

The Fernald site historically used the most restrictive DAC to limit exposures to airborne thorium and 

its progeny. The standard procedure was to count the air samples in a low-background proportional 

counter. This instrument was capable of measuring total alpha activity (beta-gamma activity was 

ignored by the appropriate setting on a pulse-height discriminator). By 1995, it was recognized that 

this method, while conservative, did not provide accurate results (Allen 1995c). The solution was to 

develop an effective DAC (EDAC) for each specific mixture of radionuclides that would apply to 

specific locations and operations on site. The equation used was:

(E-1) 1

i

i

EDAC f

DAC

=

∑

where

fi = activity fraction of isotope i in the airborne mixture

DACi = DAC of isotope i (Toler 1999)

The HIS-20 database was implemented at Fernald in the mid-1990s. Some of the data it contains 

was migrated from legacy health and safety databases. The HIS-20 database is the source of the 

DOE-provided BZ data. DOE has also provided data from HIS-20 to NIOSH. The data in the BZ 

table indicate thorium BZ results first appear in 1993. The first EDAC, designated BL-65 

(Building 65), was recorded in 1996. A technical basis for this EDAC had been published in 1995 

(Allen 1995). The four other EDACs in HIS-20 are not formally documented in any of the documents 

captured so far. This attachment provides a method of calculating intakes from the EDAC data for 

dose reconstruction purposes.

Table E-1 summarizes the EDAC results in the Fernald HIS-20 database DOE provided to NIOSH.

Table E-1. Count of EDAC results in HIS-20.

Radionuclide

Count of 

results

Year of 

minimum

Year of 

maximum

BL-13 924 2003 2004

BL-65 5,260 1996 2003

CELL 8 2 2005 2005

KS-65 23,034 2004 2006

RT-210 489 2001 2006

E.2 AFFECTED LOCATIONS

The likely definitions for BL-13, BL-65, and KS-65 have been located. These and potential definitions 

for CELL-8 and RT-210 are discussed in the following sections.

E.2.1 BL-13

The BL-13 EDAC is documented in Connell (2003). The calculated EDAC for a mixture of 232Th, 

230Th, 228Th, and total uranium was 3.28 × 10-12 μCi/mL. The author recommended that it be rounded 

down to 3.0 × 10-12 μCi/mL to be consistent with another project. The mean activity fractions based 

on alpha spectroscopy measurements appear in the Table E-2.
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ATTACHMENT E 

TECHNICAL BASIS FOR ASSIGNING DOSES FROM

EFFECTIVE DERIVED AIR CONCENTRATION BREATHING ZONE RESULTS (continued)

Table E-2. Activity fractions for the 

BL-13 EDAC.

Isotope/element Activity fraction

Th-232 0.0504

Th-230 0.248

Th-228 0.0565

U-total 0.645

Dose reconstructors should multiply the reported number of BL-13 DAC-hours by 3.0 × 10-12

μCi/mL/DAC and the breathing rate (1.2 × 106 mL/hr) to get the total intake. They then should multiply 

the total intake by the respective activity fractions for 232Th, 230Th, 228Th, and total uranium as follows:

• 232Th intake (μCi) = (reported DAC-hr)(3.0 × 10-12 μCi/mL/DAC)(1.2 × 106 mL/hr)(0.0504)

• 230Th intake (μCi) = (reported DAC-hr)(3.0 × 10-12 μCi/mL/DAC)(1.2 × 106 mL/hr)(0.248)

• 228Th intake (μCi) = (reported DAC-hr)(3.0 × 10-12 μCi/mL/DAC)(1.2 × 106 mL/hr)(0.0565)

• U-total intake (μCi) = (reported DAC-hr)(3.0 × 10-12 μCi/mL/DAC)(1.2 × 106 mL/hr)(0.645)

E.2.2 BL-65

The BL-65 EDAC is documented in Technical Basis for the Effective DAC for Th232 Stored in 

Building 65 (Allen 1995). The technical basis shows that the activity of 230Th is insignificant (less than 

3%) in comparison with the other isotopes in this particular mixture (see Attachment B of Allen 1995, 

p. 8). In addition, its DAC is higher than the calculated EDAC. Therefore, the 230Th is ignored. The

other long-lived isotopes are assumed to be in equilibrium. Therefore, 232Th, 228Ra, 228Ac, 228Th, and 

224Ra are each assumed to represent 20% of the activity. The resulting EDAC based on total activity 

is 2.22 × 10-12 μCi/mL. Dose reconstructors should multiply the reported number of BL-65 DAC-hours 

by 2.22 × 10-12 μCi/mL/DAC and the breathing rate (1.2 × 106 mL/hr) to get the total intake. They then 

should multiply the total intake by 0.2 to get the intakes of 232Th, 228Ra, 228Ac, 228Th, and 224Ra, as 

follows:

• 232Th intake (μCi) = (reported DAC-hr)(2.22 × 10-12 μCi/mL/DAC)(1.2 × 106 mL/hr × 0.20)

• 228Ra intake (μCi) = (reported DAC-hr)(2.22 × 10-12 μCi/mL/DAC)(1.2 × 106 mL/hr × 0.20)

• 228Ac intake (μCi) = (reported DAC-hr)(2.22 × 10-12 μCi/mL/DAC)(1.2 × 106 mL/hr × 0.20)

• 228Th intake (μCi) = (reported DAC-hr)(2.22 × 10-12 μCi/mL/DAC)(1.2 × 106 mL/hr × 0.20)

• 224Ra intake (μCi) = (reported DAC-hr)(2.22 × 10-12 μCi/mL/DAC)(1.2 × 106 mL/hr × 0.20)

E.2.3 KS-65

The K-65 silos contained high concentrations of radium-bearing wastes from former production 

activities. The activity fractions for the two K-65 silos are listed in Tables E-3 and E-4.

The EDAC that was designated “KS-65” was not located in the documentation. Calculations were 

made for each of the silos using the Equation E-1, above.

The resulting EDACs for total activity were 3.2 × 10-11 μCi/mL and 2.0× 10-11 μCi/mL for silos 1 and 2, 

respectively. Using the higher of the two would produce the most favorable intakes. Therefore, dose 

reconstructors should multiply the reported number of KS-65 DAC-hours by 3.2× 10-11 μCi/mL/DAC 

and the breathing rate (1.2 × 106 mL/hr) to get the total intake. They then should multiply the total 

intake by the respective activity fractions in the above tables. Two calculations will be necessary 
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ATTACHMENT E 

TECHNICAL BASIS FOR ASSIGNING DOSES FROM

EFFECTIVE DERIVED AIR CONCENTRATION BREATHING ZONE RESULTS (continued)

Table E-3. Isotopic composition of K-65 silo 1.

Isotope

Activity

(nCi/g) Activity fraction

U-total 1.68 1.61E–03

Ac-227 7.67 7.36E–03

Pb-210 202 1.94E–01

Po-210 281 2.70E–01

Ra-226 477 4.58E–01

Th-228 2.28 2.19E–03

Th-230 68.9 6.62E–02

Th-232 1.11 1.07E–03

Table E-4. Isotopic composition of K-65 silo 2.

Isotope

Activity

(nCi/g) Activity fraction

U-total 2.37 3.04E–03

Ac-227 6.64 8.50E–03

Pa-231 4.04 5.17E–03

Pb-210 190 2.43E–01

Po-210 231 2.96E–01

Ra-226 263 3.36E–01

Th-228 7.36 9.42E–03

Th-230 76.2 9.75E–02

Th-232 0.99 1.26E–03

because not only are the activity fractions different, but also Table E-4 contains 231Pa in addition to the 

isotopes in Table E-3. The calculations are as follows:

Silo 1

• U-total intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2E × 106 mL/hr)(1.61 × 10-3)

• 227Ac intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(7.36 × 10-3)

• 210Pb intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(1.94 × 10-1)

• 210Po intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(2.70 × 10-1)

• 226Ra intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(4.58 × 10-1)

• 228Th intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(2.19 × 10-3)

• 230Th intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(6.62 × 10-2)

• 232Th intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(1.07 × 10-3)

Silo 2

• U-total intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(3.04 × 10-3)

• 227Ac intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(8.50 × 10-3)

• 231Pa intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(5.17 × 10-3)

• 210Pb intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(2.43 × 10-1)

• 210Po intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(2.96 × 10-1)

• 226Ra intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(3.36 × 10-1)

• 228Th intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(9.42 × 10-3)

• 230Th intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(9.75 × 10-2)

• 232Th intake (μCi) = (reported DAC-hr)(3.2 × 10-11 μCi/mL/DAC)(1.2 × 106 mL/hr)(1.26 × 10-3)
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ATTACHMENT E

TECHNICAL BASIS FOR ASSIGNING DOSES FROM

EFFECTIVE DERIVED AIR CONCENTRATION BREATHING ZONE RESULTS (continued)

E.2.4 CELL-8

The Onsite Disposal Facility (OSDF) was used to dispose of wastes such as contaminated soil and 

building rubble that were too low-level to justify sending off site. The OSDF was filled in sections, 

called cells. The waste would have been characterized before being placed in the cells. Cell 8 was 

the last cell filled. The final cap of cell 8 was installed in 2006. There are only two results in HIS-20 

with this EDAC designation, making up less than 0.01% of the results. The results are for two 

individuals on September 6, 2005. Both wore respirators and both results were below the minimum 

detectable DAC-hours shown in the HIS-20 records. Neither individual is a claimant. The method for 

addressing this EDAC has been deferred because it may never be needed.

E.2.5 RT-210

The remaining EDAC, RT-210, makes up only 1.7% of the results. There were two radon treatment 

(RT) systems. Because of the relatively small number of results, it is assumed that the RT-210 results 

apply to workers at the RT systems. The first system was a temporary arrangement constructed 

around 2001 that was used to lower radon emissions from the K-65 silos while the material was in 

storage. Little information has been found for this system. Construction of the new Radon Control 

System (RCS) for silos 1 and 2 (as well as some other facilities) was completed and operations began 

in 2003. The RCS was designed to draw radon gas from the headspace inside silos 1 and 2 and 

reduce the concentration of gas by at least 95% to protect workers during removal of the K-65 

material. The silo 1 and 2 wastes were removed by slurry, solidified, and placed in casks. The casks 

were shipped to an off-site facility for disposal.

The date range for RT-210 is 2001 to 2006, which covers both RT systems. In 2006, the Silos 1 and 

2 Remediation Facility, Transfer Tank Area (TTA), and RCS demolitions were completed.

The following description of the Fernald RCS was located in USACE (2011):

Centrifugal fans pulled radon-laden gas from the sources through the roughing filters for 

initial particulate daughter removal. The air stream was chilled and dried to enhance the 

dynamic adsorption capacity of the activated carbon. Condensed liquids from the gas 

stream were transferred to shielded holdup tanks until transfer and disposition in the 

Remediation Facility could be completed. There were four carbon beds, each containing 

20,400 kg (45,000 lbs) of carbon. These beds were configured so that any two of the 

four beds were in use at any given time. This allowed for decay time of the alternate two 

beds, whereby no carbon changes were required over the life of the project. The RCS 

reduced radon concentration to less than 2% of the inlet concentration and the carbon 

bed outlet air was either recycled to the silos or exhausted through the 46-m (150-ft) tall 

stack. Approximately four inches of carbon steel shielding was designed and installed 

adjacent to the carbon beds to reduce general area dose rates ... ventilation 

requirements were considered for all tasks where radon concentrations greater than 

0.01 WL are expected in the air in occupied spaces, such as the TTA and RCS buildings.

The 222Ra decay chain is shown in Figure E-1. During the operation of the facility, the three long-lived 

progeny, 210Pb, 210Bi, and 210Po, built up in the system’s piping and filters. These isotopes would have 

been the concern for worker exposures during shutdowns and maintenance (e.g., filter changes). 

Information from the site indicates that the RT-210 EDAC was most likely calculated assuming 

equilibrium among the three isotopes (Hinnefeld 2014). The activity fractions are shown in Table E-5.
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ATTACHMENT E 

TECHNICAL BASIS FOR ASSIGNING DOSES FROM

EFFECTIVE DERIVED AIR CONCENTRATION BREATHING ZONE RESULTS (continued)

Rn-222

3.82 d

Beta Decay

Alpha Decay

Pb-214

26.8 min

Po-218

3.05 min

Bi-214

19.7 min

Pb-210

22 y

Po-214

1.6E-04 s

Bi-210

5.0 d

Pb-206

Stable

Po-210

138 d

Figure E-1. 222Rn decay chain showing half-lives.

Table E-5. Activity fractions for the 

RT-210 EDAC.

Isotope Activity fraction 

Pb-210 0.333

Bi-210 0.333

Po-210 0.333

The following DACs were obtained from 10 CFR Part 835, Appendix A:

• 210Pb: 1 × 10-10 μCi/mL,

• 210Bi: 1 × 10-7 μCi/mL, and

• 210Po: 3 × 10-10 μCi/mL.

From the DAC values and an activity fraction of 0.333 for each isotope an EDAC of 2.25 × 10-10

μCi/mL was calculated. Dose reconstructors should multiply the reported number of RT-210 DAChours by 2.25 × 10-10 μCi/mL/DAC and the breathing rate (1.2 × 106 mL/hr) to get the total intake. 

They then should multiply the total intake by 0.333 to get the intakes of 210Pb, 210Bi, and 210Po, as 

follows:

• 210Pb intake (μCi) = (reported DAC-hr)(2.25 × 10-10 μCi/mL/DAC)(1.2 × 106 mL/hr)(0.333)

• 210Bi intake (μCi) = (reported DAC-hr)(2.25 × 10-10 μCi/mL/DAC)(1.2 × 106 mL/hr)(0.333)

• 210Po intake (μCi) = (reported DAC-hr)(2.25 × 10-10 μCi/mL/DAC)(1.2 × 106 mL/hr)(0.333)
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989

TABLE OF CONTENTS

SECTION TITLE PAGE

F.1 Purpose ............................................................................................................................... 177

F.2 Thorium Mixtures ................................................................................................................. 177

F.3 Thorium Used for MIVRML Calibration................................................................................. 179

F.4 Triple-Separation Thorium.................................................................................................... 179

F.5 Data and Method Overview.................................................................................................. 183

F.5.1 Data Selection .......................................................................................................... 184

F.5.2 Statistical Description Of Data .................................................................................. 184

F.6 Intake Modeling.................................................................................................................... 185

LIST OF TABLES 

TABLE TITLE PAGE

F-1 Bias for 212Pb in chest .......................................................................................................... 184

F-2 In vivo lung counting statistics and 232Th lung burdens......................................................... 185

F-3 Derived thorium intakes, 1979 through 1989........................................................................ 186
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

LIST OF FIGURES 

FIGURES TITLE PAGE

F-1 Natural thorium decay chain showing isotopes and half-lives....Error! Bookmark not defined.

F-2 Relative activity of the isotopes in the natural thorium decay chain as a function of time 

after chemical purification..................................................................................................... 179

F-3 228Th activity after a single chemical purification................................................................... 180

F-4 228Ra and 224Ra ingrowth after thorium chemical separation................................................. 181

F-5 228Th activity after a second chemical purification at its lowest equilibration point 

(4.55 years) after the first purification ................................................................................... 181

F-6 224Ra and 228Ra activity after a second chemical purification when 228Th was at its lowest 

equilibration point (4.55 years) after the first purification ...................................................... 182

F-7 228Th activity after a third chemical purification when at its lowest equilibration point 

(7.09 years) after the first purification ................................................................................... 182

F-8 224Ra and 228Ra activity after the third chemical separation .................................................. 183

F-9 50th- and 84th-percentile 232Th chest burdens, 1979 to 1988............................................... 185

F-10 Predicted thorium lung burden calculated using the three-separation model to derive 

thorium intake rates compared with measured chest burden, 1979 to 1989, 

50th percentile, Type M........................................................................................................ 186

F-11 Predicted thorium lung burden calculated using the three-separation model to derive 

thorium intake rates compared with measured chest burden, 1979 to 1989, 

84th percentile, Type M........................................................................................................ 187

F-12 Predicted thorium lung burden calculated using the three separation model to derive 

thorium intake rates compared with measured chest burden, 1979 to 1989, 

50th percentile, Type S ........................................................................................................ 187

F-13 Predicted thorium lung burden calculated using the three separation model to derive 

thorium intake rates compared with measured chest burden, 1979 to 1989, 

84th percentile, Type S ........................................................................................................ 188
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

F.1 PURPOSE

Some employees at DOE sites were not monitored for internal ionizing radiation exposure, or the 

records of such monitoring are incomplete or unavailable. In such cases, data from monitored 

coworkers can be used to estimate an individual’s possible exposure. The purpose of this attachment 

is to provide information about monitored coworker exposures to thorium from 1979 to 1989. This 

information may be used for calculating and assigning occupational internal doses to employees at 

Fernald for whom no or insufficient monitoring records exist.

The thorium intake rates developed in this coworker study are to be used for dose reconstruction for 

the period from 1979 through 1989. During Fernald Working Group discussions, agreement was 

reached that the median of the coworker model would be used as a minimum value for intake rate 

regardless of the data in a worker’s internal dose record (NIOSH 2012, NIOSH 2014b). This ensures 

that no in vivo count underestimates the intake rate due to a count being conducted shortly after 

chemical separation, before equilibration of 228Th and 212Pb can occur.

F.2 THORIUM MIXTURES

Natural Thorium

Natural thorium is composed of equilibrated 232Th and 228Th and their decay products as shown in 

Figure F-1. Because the specific activity of 232Th is small (1.1 × 10-7 Ci/g) in comparison with 228Th 

(820 Ci/g) (Eckerman and Sjoreen 2006) essentially all of the mass in any thorium mixture is 

associated with 232Th. In 1.0 g of natural thorium there would be nearly 1.0 g of 232Th and only 13 ng 

of 228

Th (Eckerman and Sjoreen 2006).
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

Figure F-1. Natural thorium decay chain showing isotopes and half-lives. Alpha decays are 

represented by vertical arrows, beta decays by horizontal arrows (ORNL 2006).

After a single chemical purification of natural thorium, the isotopic ratio of progeny is disrupted and the 

degree of equilibrium changes as a function of time while ingrowth of progeny occurs, as illustrated in 

Figure F-2.

0.64

0.36
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

0 2 4 6 8 10 12 14 16 18 20

0.01

0.1

1

Th-228 total

Th-228 unsupported

Th-228 supported

Ac-228

Time (years)

Activity Relative to Th-232

Figure F-2. Relative activity of the isotopes in the natural thorium decay chain as a 

function of time after chemical purification. All 220Rn and decay products are assumed to 

be retained in the thorium material (ORNL 2006).

One of the bounding scenarios examined in this coworker model assumes that the 232Th, 228Th, and all 

progeny are in equilibrium. This is an extreme example that is unlikely to have occurred because 

natural thorium would have been processed after mining and before arrival at FMPC. However, this 

scenario could have been realized if the thorium had been refined two or three decades before 

processing at FMPC.

F.3 THORIUM USED FOR MIVRML CALIBRATION

The MIVRML was calibrated using a thorium source described in a note published in Health Physics

(Scott 1966). This was confirmed by an interview with a subject matter expert (ORAUT 2012). The 

calibration source had an activity ratio of 232Th to 228Th of 1.27. In other words, the equilibrium 

between 228Th and 232Th was 0.787. Because of the short half-life of 228Th progeny, it is reasonable to 

assume that 228Th and 212Pb exist in equilibrium. Note that for this to occur, 220Rn, a noble gas that is 

an intermediate member of the decay chain, is assumed to remain in place. Because its half-life is 

short (56.6 s), this is a reasonable assumption, even though not rigorously true [11].

Thorium material similar to this calibration source would have been chemically purified approximately 

1 year before analysis and is deemed to be a realistic mixture, likely to be encountered during 

operations.

F.4 TRIPLE-SEPARATION THORIUM

Another bounding scenario considered involves a sequence of thorium purifications that are timed to 

produce the greatest possible reduction of the ratio of 228Th to 232Th (ORAUT 2014b). Because of its 

long half-life the activity of 232Th does not change with time, but the activity of 228Th decreases then 
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

increases as shown in Figure F-3. The activity of both isotopes is assumed to be 1 Bq for illustration 

when chemical purification is completed. This is illustrated in a simpler depiction in Figure F-3.

0 5 10 15 20

0.1

1

Th228

Th-228 Present After Separation

Time (years)

Th-228 (Bq)

Figure F-3. 228Th activity after a single chemical purification. The minimum occurs 4.55 

years after purification.

For all practical applications 224Ra, with its short 3.66-day half-life, is always in secular equilibrium with 

228Th, but 228Ra, with a 5.75-year half-life, is depleted and takes years to grow back to any significant 

degree. The ingrowth curves for 224Ra and 228Ra are shown in Figure F-4. The minimum in the 224Ra 

curve coincides with the minimum in the 228Th curve.

If the thorium source were again stripped of all progeny at 4.55 years, the 228Th ingrowth curve in 

Figure F-5 would result, which has a minimum of 0.26 Bq at 2.54 years after the second separation 

(or 2.54 + 4.55 = 7.09 years after the first separation). The corresponding 224Ra and 228Ra curves 

after the second separation are shown in Figure F-6.

If this thorium source were again stripped of all progeny a third time at 7.09 years after the first 

chemical separation, the 228Th ingrowth curve in Figure F-7 would result, which has a minimum of 

0.19 Bq at 1.75 years after the third separation (or 1.75 + 2.54 + 4.55 = 8.84 years after the first 

separation). The corresponding 224Ra and 228Ra curves after the third separation are shown in 

Figure F-8.
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

0 5 10 15 20

0.1

1

Ra224

Ra228

Ra-228 & 224 Present After Separation

Time (years)

Activity (Bq)

Figure F-4. 228Ra and 224Ra ingrowth after thorium chemical separation.

0 5 10 15 20

0.1

1

Th228

Th-228 Present After Separation

Time (years)

Th-228 (Bq)

Figure F-5. 228Th activity after a second chemical purification at its lowest 

equilibration point (4.55 years) after the first purification. The activity minimum is 

reached 7.09 years after the first purification.
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

0 5 10 15 20

0.1

1

Ra224

Ra228

Ra-228 & 224 Present After Separation

Time (years)

Activity (Bq)

Figure F-6. 224Ra and 228Ra activity after a second chemical purification when 

228Th was at its lowest equilibration point (4.55 years) after the first purification.

0 5 10 15 20

0.1

1

Th228

Th-228 Present After Separation

Time (years)

Th-228 (Bq)

Figure F-7. 228Th activity after a third chemical purification when at its lowest 

equilibration point (7.09 years) after the first purification. The activity minimum, 

0.19 Bq, is reached 8.84 years after the first purification. 
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

0 5 10 15 20

0.1

1

Ra224

Ra228

Ra-228 & 224 Present After Separation

Time (years)

Activity (Bq)

Figure F-8. 224Ra and 228Ra activity after the third chemical separation.

In summary, this scenario results in the equilibrium ratio between 232Th and 212Pb reaching a minimum 

value of 0.19. This chronic intake analysis assumes that all of the activity inhaled always has this 

most pessimistic isotopic makeup. That is, this scenario results in the highest intake rates and is 

therefore most favorable to claimants.

F.5 Data and Method Overview

ORAUT-OTIB-0019, Analysis of Coworker Bioassay Data for Internal Dose Assignment (ORAUT 

2005b), describes the general process that is used to analyze bioassay data for assigning doses to 

individuals based on coworker results. ORAUT-PLAN-0014, Coworker Data Exposure Profile 

Development (ORAUT 2004), describes the approach and processes to be used to develop 

reasonable exposure profiles based on available dosimetric information for workers at DOE sites.

The Y-12 MIVRML was used at Fernald from 1968 through 1989. Before its use at Fernald, the 

MIVRML system was described in a management memorandum to all employees (Heatherton 

undated). The memorandum explains that those workers with the highest uranium and thorium 

exposure potential would be counted most frequently and those with virtually no exposure potential 

would not be routinely counted at all. Workers identified to have potential exposures to thorium were 

given priority for counting during the first use of the MIVRML (ORAUT 2007) [12].

The technology and challenges associated with in vivo counting for thorium were described by 

West (1962, 1965). An interview with [name redacted] was conducted (ORAUT 2012) and provided 

additional specific details on the calibration and operation of the system while at Fernald.
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

F.5.1 Data Selection

Handwritten records of in vivo chest count data for individual Fernald workers were obtained 

(NLO 1968–1988a through NLO 1968–1988v). The great majority of the entries in these records 

represent data obtained between 1968 and 1988 using the MIVRML. Data from the records from 

1979 through 1988 were transcribed into a spreadsheet and validated as described by ORAUT 

(2006a). A total of 2,994 records from this time interval were transcribed. This represented data on 

approximately 1,000 individuals. Most results were reported as “Pb-212” and “Ac-228” activities; only 

the 212Pb data was used since it is most indicative of a thorium intake (ORAUT 2014b).

The ORAU Team PID reviewed the transcribed data. This review yielded a number of modifications 

and exclusions from the statistical analysis. The reasons for exclusion included:

• Results labeled “AEC/ERDA/DOE recount program,”

• Control samples,

• Those indicated to be “bad samples” or “unreliable,” and

• Entries with no results (most of these appear to have been counts that were scheduled but not

performed; comments included “claustrophobic,” “off sick,” and “count was canceled because

of high background problems”).

Samples indicated to be recounts or duplicates were averaged into a single result for the given date or 

sample.

F.5.2 Statistical Description Of Data

The chest count data were statistically evaluated using yearly intervals according to the specifications 

and requirements in ORAUT (2006b) and ORAUT (2014c) using the TWOPOS methodology. In vivo 

chest count data statistics were generated for each analysis interval. The 212Pb chest counts were 

determined to be biased, so a bias correction factor was subtracted from each individual chest count 

as given in Table F-1.

Table F-1. Bias for 212Pb 

(nCi) in chest [13].

Year 212Pb bias

1979 –0.014

1980 –0.045

1981 –0.056

1982 –0.058

1983 –0.101

1984 –0.089

1985 –0.086

1986 –0.085

1987 –0.047

1988 0.002

A lognormal distribution was assumed [14], and the 50th- and 84th-percentile values were calculated. 

All calculations were independently repeated for validation. The values were converted to 232Th lung 

burdens by dividing by 0.19, as discussed above. Table F-2 presents the annualized results of the 

statistical analysis for 232Th in nanocuries. Figure F-9 shows plots of the 50th- and 84th-percentile 

chest burden.
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

Table F-2. In vivo lung counting statistics and 232Th lung burdens.

Year

Effective 

bioassay date

# of TWOPOS 

results

50th percentile 

(nCi)

84th percentile 

(nCi)

1979 6/30/1979 124 0.031 0.126

1980 6/30/1980 149 0.027 0.107

1981 6/30/1981 129 0.034 0.113

1982 6/30/1982 167 0.039 0.152

1983 6/30/1983 167 0.026 0.122

1984 6/30/1984 322 0.032 0.108

1985 6/30/1985 355 0.038 0.104

1986 6/30/1986 433 0.031 0.101

1987 6/30/1987 523 0.037 0.097

1988 6/30/1988 108 0.045 0.104

Figure F-9. 50th- and 84th-percentile 232Th chest burdens, 1979 to 1988.

F.6 INTAKE MODELING

Results from the statistical analyses, adjusted to represent 232Th activity in picocuries as discussed 

above, were used to determine thorium coworker intake rates. The biokinetic model described in 

ORAUT-OTIB-0076 (ORAUT 2014b) was used to fit the in vivo measurement data associated with 

thorium isotopes to a series of chronic inhalation intakes for both types M and S.

A single chest count result for each period was assumed to have occurred at the midpoint of the 

period. A uniform absolute error of 1 was applied to all results, which assigned the same weight to 

each result. The intake assumptions were based on patterns observed in the in vivo data. A single 

constant chronic intake rate was chosen due to the consistent magnitude of the chest counts.

The resulting coworker thorium intake data are shown in Table F-3. GSD values are equal to the

84th-percentile intake rates divided by the 50th-percentile intake rates. To account for the error in 

biokinetic modeling, a minimum GSD value of 3.00 is assigned because that is the assumed GSD for

an individual who was monitored. The data for 1980 through 1988 are extrapolated to include 

1989 [15]. Figures F-10 through F-13 illustrate the result of coworker intake modeling for thorium.
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

Table F-3. Derived thorium intakes, 1979 through 1989. 

Intake solubility 

type

50th 

percentile 

(pCi/d)

84th 

percentile 

(pCi/d) GSD

95th 

percentile 

(pCi/d)

Type M 35.3 117 3.00 215

Type S 5.46 17.0 3.00 33.3

Figure F-10. Predicted thorium lung burden (line) calculated using the threeseparation model to derive thorium intake rates compared with measured chest 

burden (dots), 1979 to 1989, 50th percentile, Type M.
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

Figure F-11. Predicted thorium lung burden (line) calculated using the threeseparation model to derive thorium intake rates compared with measured chest 

burden (dots), 1979 to 1989, 84th percentile, Type M.

Figure F-12. Predicted thorium lung burden (line) calculated using the three 

separation model to derive thorium intake rates compared with measured chest 

burden (dots), 1979 to 1989, 50th percentile, Type S.
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ATTACHMENT F 

THORIUM IN VIVO COWORKER STUDY, 1979 to 1989 (continued)

Figure F-13. Predicted thorium lung burden (line) calculated using the three 

separation model to derive thorium intake rates compared with measured chest 

burden (dots), 1979 to 1989, 84th percentile, Type S.
Document No. ORAUT-TKBS-0017-5 Revision No. 0

2 Effective Date: 09/30/2016 Page 189 of 189

ATTACHMENT G 

CLASS W THORIUM-232 INHALATION INTAKE RATE

BASED ON 10% OF THE DERIVED AIR CONCENTRATION

100% 232Th DAC 5.00 × 10-13 μCi/cm3

10% 232Th DAC 5.00 × 10-14 μCi/cm3 (5.00 × 10-14 μCi/cm3

)(2.22 × 10+6 dpm/μCi) 

= 1.11 × 10-7 dpm/cm3

(1.11 × 10-7 dpm/cm3 is based on 2.22 × 10+6 dpm/μCi)

Work year 2,000 hr/yr (2,000 hr/yr) ÷ (8 hr/d)

= 250 days

Breathing rate (ICRP 1995) 1.2 m3

/hr 1.2 × 106 cm3/hr based on 1.00 × 106 cm3/m3

OCAS-TIB-009 0.2 Intake per workday, not calendar day 

= 250 days

Annual inhalation intake of 232Th 266.400 dpm/yr (1.11 × 10-7 dpm/cm3

)(1.2 × 106 cm3/hr)(2,000 hr/yr) 

= 266.4 dpm/yr

Annual ingestion intake of 232Th 5.550 dpm/yr (1.11 × 10-7 dpm/cm3

)(1.00 × 106 cm3/m3)(0.2)(250 d/yr) 

= 5.55 dpm/yr