Leukemia Mortality and Occupational Ionizing Radiation Exposure

A dissertation submitted to the Graduate School of the University of Cincinnati in partial fulfillment of the requirements for the degree of

Doctor of Philosophy

in the Department of Environmental Health of the College of Medicine October 5, 2011

By Robert D. Daniels

B.S. Thomas Edison State College, June 1996

Committee Chair: S. Pinney, Ph.D.

ABSTRACT Context: Nearly four million U.S. workers are potentially exposed to ionizing radiation in the course of their employment each year. Although, evidence exists that unequivocally establishes ionizing radiation as a human carcinogen, little is known about the effects of protracted low-dose exposures.

Objective: Conduct a case-control analysis examining the relation between protracted low-dose ionizing radiation exposure and leukemia in a cohort of U.S. nuclear workers.

Methods: Each case was matched to four controls on attained age. Ionizing radiation exposures were assessed using measurements and employment histories. Conditional logistic regression analyses were conducted using general relative risk models to estimate the excess relative risk (ERR) of all leukemia, leukemia excluding CLL, AML, CML, and CLL while controlling for potential confounding by race, sex, benzene exposure, social economic status, and either hire date or birth date. Results were tested under

differing exposure lag assumptions and time windows of exposure.

Results: There were 369 leukemia cases identified in a population of 105,245 U.S. nuclear workers.

Positive, but imprecise risk estimates were observed for all outcomes excluding CLL, although the observed dose response for non-CLL leukemia, AML, and CML showed attenuated risk in the low dose

(<10 mGy) and high dose (>100 mGy) regions. The linear ERR per 10 mGy absorbed dose to bone marrow was 0.009 (95% CI:-0.014, 0.051). A three-piece linear spline model best fit the non-CLL data, where slope estimates (ERR∙10 mGy -1) were statistically significant in the first two segments: -0.68 (95%

CI: -0.92, -0.33) for doses ≤8.0 mGy; 0.20 (95% CI: 0.082, 0.35), dose=8<-46 mGy; and -0.016 (95% CI: <-

0.022, 0.018), dose=46+ mGy. Leukemia risks were characterized by a “wave-like” function of time,

where peak risks were observed from exposures occurring from five to ten years prior to the age at

death of the index AML case (ERR∙10 mGy -1 = 0.76; 95% CI: 0.047, 2.7) and from 10-15 years for CML

(ERR∙10 mGy -1 = 0.44; 95% CI: <0, 2.4). Rate ratio modification was not observed in any model, although

risk estimates in most linear models tested were confounded by sex, race, and hire date or birth date.

This confounding disappeared in piecewise models that provided a better fit to the complex dose-

response.

Conclusions: This study of leukemia radiogenicity from protracted occupational radiation exposure is

the largest ever conducted. Estimates from linear ERR models were in fair agreement with other studies;

however, the non-linearity of the dose response contributed to the imprecision of estimates. Leukemia

risks were observed to vary with time since exposure, where large and monotonic increases in myeloid

leukemia risk were evident for exposures accrued in select windows of time prior to the age of the case

at death. Future research should be focused to better evaluate the shape of the dose-response,

particularly in the low-dose range, which is critical for risk assessment purposes. Moreover,

examinations of temporal patterns of risk appear to be crucial in future studies examining risk factors for

leukemia.

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PREFACE This research was funded by the National Institute for Occupational Safety and Health (NIOSH) following competitive award as an intramural research project under the National Occupational

Research Agenda (NORA). The project proposal was reviewed and strongly recommended for funding by both internal NIOSH and external peer reviewers. In addition, tripartite review, involving representatives from labor, management, and the scientific community was conducted in accordance with NIOSH policy.

This research complies with the requirements of the Federal Policy for Protection of Human Subjects

(10CFR745 or, where applicable, 45CFR46), and was reviewed by the NIOSH Human Subjects Review

Board (HSRB) to ensure that the rights and welfare of study subjects are protected (Protocol No.: HSRB

08-DSHEFS-04-XP). Additional reviews were conducted by the respective IRBs of the University of

Cincinnati (Protocol No.: 10-01-14-01) and the U.S. Department of Energy (Protocol No.: CDOE-(11)-

118). These reviews were necessary because:

• The project serves to satisfy the University's requirements for dissertation research for the NIOSH

Principal Investigator. As such, the UC faculty members participating on the dissertation committee

are collaborative partners.

• The project complies with the requirements for IRB review of health research and related studies at

DOE facilities in accordance with guidance in the DOE Access Handbook (DOE EH/0556, 2003).

Disclaimer: Mention of any company or product does not constitute endorsement by the National

Institute for Occupational Safety and Health (NIOSH).Citations to Web sites external to NIOSH do not constitute NIOSH endorsement of the sponsoring organizations or their programs or products. The findings and conclusions in the report are those of the author and do not necessarily represent the views of NIOSH.

ACKNOWLEDGMENTS This study was conducted under the NIOSH Occupational Energy Research Program, which was founded and managed under a long standing agreement between the U.S. DOE and the U.S. DHHS. The research described herein was made possible through the cooperation and support of the DOE and their employees and contractors. Current and former staff at DOE headquarters, including Ms. M. Lawn, Dr.

G. Petersen, and Dr. B. Richter, ensured that NIOSH investigators were allowed access to all essential documents at the various study sites, as well as provided ongoing encouragement for completion of the study. Employees from the Portsmouth Naval Shipyard (PNS), Hanford Site, Savannah River Site (SRS),

Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL) and the Los Alamos National

Laboratory (LANL) provided numerous sources of information that were required to complete this complex study. This study was also made possible through the continued support of NIOSH co- investigators, including Dr. S. Bertke who provided guidance in statistical methods and Ms. K. Waters, who assisted in computer programming. I am also very grateful to my dissertation committee: Dr. S.

Pinney, Dr. M. Schubauer-Berigan, Dr. C.R. Buncher, Dr. R. Hornung, and Dr. H. Spitz, for their encouragement and valuable input throughout all phases of this research project.

Contents

1.0 Introduction ...... 1

2.0 Leukemia ...... 4

2.1 Description ...... 4

2.2 Acute Lymphocytic Leukemia (ALL)...... 5

2.3 Acute Myeloid Leukemia (AML) ...... 6

2.4 Chronic Myeloid Leukemia (CML), ...... 6

2.5 Chronic Lymphocytic Leukemia (CLL) ...... 7

2.6 Leukemia Trends ...... 8

2.7 Risk Factors ...... 10

2.7.1 Lifestyle factors ...... 10 2.7.2 Environmental factors ...... 15 2.7.3 Therapy-related myeloid neoplasms ...... 47 2.7.4 Genetic Factors ...... 47

3.0 Study Cohort ...... 49

3.1 Definition ...... 49

3.1.1 Vital Status Ascertainment ...... 50 3.1.2 Results ...... 52 3.2 Facility Descriptions ...... 54

3.2.1 Hanford ...... 57 3.2.2 INL ...... 60 3.2.3 LANL ...... 63 3.2.4 ORNL ...... 69 3.2.5 PNS ...... 72 3.2.6 SRS ...... 75 3.3 Previous Studies at Primary Sites ...... 78

3.3.1 DOE Health and Mortality Study Overview ...... 79

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3.3.2 Hanford ...... 80 3.3.3 INL ...... 82 3.3.4 LANL ...... 83 3.3.5 ORNL ...... 85 3.3.6 PNS ...... 86 3.3.7 SRS ...... 90 3.3.8 Pooled studies ...... 92

4.0 Study Roster ...... 96

4.1 Case and Control Selection ...... 96

4.2 Demographics ...... 98

4.3 Occupation and Social Class ...... 102

5.0 Exposure Assessment ...... 105

5.1 External low-LET ionizing radiation ...... 106

5.1.1 Facility dosimetry ...... 106 5.1.2 Methods ...... 115 5.1.3 Results ...... 130 5.2 Other occupational exposures ...... 134

5.2.1 Methods ...... 135 5.2.2 Results ...... 137 5.2.3 Limitations ...... 140 5.3 Work-related medical X-ray exposure ...... 141

5.3.1 Introduction ...... 141 5.3.2 Methods ...... 142 5.3.3 Results ...... 149 5.3.4 Limitations ...... 149 5.4 Neutron Exposure Assessment ...... 150

5.4.1 Facility dosimetry ...... 150 5.4.2 Methods ...... 162 5.4.3 Results ...... 168

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5.5 Plutonium Deposition ...... 171

5.5.1 Dose evaluation ...... 171 5.5.2 Results ...... 172 5.5.3 Site characteristics ...... 172 5.5.4 Results ...... 184 5.6 Tritium Exposures ...... 185

5.6.1 Background ...... 185 5.6.2 Methods ...... 186 5.6.3 Results ...... 189 5.6.4 Discussion ...... 202 5.7 Benzene exposures ...... 203

5.7.1 Records collection and abstraction ...... 205 5.7.2 Estimates of exposure level, frequency, and duration ...... 206 5.7.3 Exposure score algorithm ...... 208 5.7.4 General Assignments ...... 209 5.7.5 Descriptive Results ...... 235 5.8 Final exposure assignments for Case and Controls...... 238

6.0 Statistical Analysis ...... 242

6.1 Methods ...... 242

6.2 Results ...... 245

6.3 Univariate analysis ...... 245

6.3.1 Multivariable analyses ...... 252 6.4 Discussion ...... 272

7.0 Conclusions ...... 283

8.0 Bibliography ...... 285

Appendix I: Contributions ...... 325

List of Tables

Table 2-1. U.S. Trends in leukemia by subtype [Altekruse et al. 2010]...... 9 Table 2-2. Studies selected for meta analysis [Daniels and Schubauer-Berigan 2011] ...... 23 Table 2-3. Studies of leukemia and ionizing radiation exposure that did not meet inclusion criteria for meta-analysis...... 25 Table 3-1. Characteristics of the base cohort assembled for case-control selection ...... 54 Table 3-2. ORNL reactors ...... 70 Table 3-3. External whole-body dose limits (mSv) and administrative controls at PNS (1958-1978) ...... 74 Table 4-1. Leukemia subtype classification ...... 96 Table 4-2. Distribution of leukemia cases by subtype...... 99 Table 4-3. Case distribution by site...... 99 Table 4-4. Distribution of the age at death by the year of death...... 100 Table 4-5. Distribution of the age at death by birth cohort quartiles...... 100 Table 4-6. Demographic information on cases and controls ...... 102 Table 4-7. Socioeconomic status assignments among cases and controls ...... 103 Table 5-1. External dosimetry data files for penetrating radiation exposure information...... 117 Table 5-2: Film badge filter characteristics by facility and era used ...... 129 Table 5-3: Bias factors ( B) and uncertainty factors ( K) used to estimate absorbed dose to the active bone

marrow ( Brbm , K rbm ) from recorded exposures...... 130 Table 5-4. Cohort level cumulative dose (mSv) statistics for low-LET penetrating radiation ...... 132 Table 5-5. Combinations of sites with exposure information on 10 or more workers in the base cohort...... 133 Table 5-6. Cumulative doses (mGy) from penetrating low-LET radiation exposures at other facilities ... 139 Table 5-7: Calculated dose coefficient (bone marrow, mGy∙exam -1) for PFG and lumbar spine examinations at study facilities...... 148 Table 5-8. Cumulative dose (mGy) to study participants from work-related x-ray examinations ...... 149 Table 5-9. General characteristic of common neutron dosimeters...... 153 Table 5-10. Other INL Neutron sources ...... 157 Table 5-11. Neutron dosimeter type, period of use, and exchange frequency 1 [Fix et al. 2005] ...... 164 Table 5-12. Workplace neutron spectra for worker exposure groups [Fix et al. 2005] ...... 165 Table 5-13. Group and exposure scenarios [Fix et al. 2005] ...... 166

Table 5-14. Calculated dose conversion coefficients for converting recorded neutron dose to absorbed dose to bone marrow [Fix et al. 2005]...... 167 Table 5-15. Cohort level cumulative dose (mSv) statistics for neutron radiation ...... 170 Table 5-16. Major plutonium in urine bioassay methods in study sites ...... 175 Table 5-17. Sources of plutonium bioassay information from 1945 through 2005...... 176 Table 5-18: Plutonium exposure category assignments for study subjects ( n=1,816) ...... 185 Table 5-19: Dose conversion coefficients used to normalize tritium doses...... 187 Table 5-20: Hanford tritium exposures ...... 191 Table 5-21: LANL tritium releases 1943-1972 ...... 193 Table 5-22: Tritium exposures among LANL cases and controls ...... 194 Table 5-23. Tritium exposures among cases and controls employed at ORNL...... 197 Table 5-24: Primary facilities for tritium exposures at SRS ...... 198 Table 5-25. Tritium exposures among cases and controls employed at SRS ...... 200 Table 5-26. Benzene monitoring information found in Mihlan data [Mihlan 1997] ...... 206 Table 5-27. Recommended benzene exposure levels ...... 207 Table 5-28. Benzene exposure levels ...... 207 Table 5-29. Worker categories ...... 208 Table 5-30. Default assumptions for calculating exposure scores for operations and developmental activities involving known or suspected of benzene exposure...... 210 Table 5-31. Default Time-dependent Weighting Factors (TWF) [2007] ...... 211 Table 5-32. Annual Benzene exposure scores for gasoline dispensing and automotive repair...... 214 Table 5-33. Benzene air samples in painting tasks ...... 216 Table 5-34. Annual benzene exposure scores for painters and painter supervisors ...... 217 Table 5-35. Annual benzene exposure scores for general laboratory workers and supervisors...... 220 Table 5-36. Benzene exposure scores for plutonium bioassay in Building 706 prior to 1954...... 222 Table 5-37. LANL task specific benzene exposure score assignments ...... 227 Table 5-38. ORNL benzene samples ( n=167) by department and job title ...... 229 Table 5-39. ORNL benzene samples ( n=69) by work activity and department ...... 230 Table 5-40. PNS task specific benzene exposure score assignments prior to 1950...... 233 Table 5-41. Summary description of exposure information for cases and controls (two-year lag) ...... 239 Table 5-42. Effects on dose (mGy) from exposure lagging of Low-LET radiation by leukemia subtype ... 240 Table 5-43. Unexposed cases and controls, no lag period ( n=9) ...... 241

Table 6-1. Univariate analysis of covariates in analysis of leukemia cases and age-matched controls ... 247 Table 6-2. Univariate analysis of covariates in analysis of non-CLL cases and age-matched controls ..... 248 Table 6-3. Univariate analysis of covariates in analysis of CLL cases and age-matched controls (excl. indeterminate cases) ...... 249 Table 6-4. Univariate analysis of covariates in analysis of CML cases and age-matched controls ...... 250 Table 6-5. Univariate analysis of covariates in analysis of AML cases and age-matched controls ...... 251 Table 6-6. Effects on low-LET radiation risk estimates from neutrons (n) and/or plutonium (p) exposures in leukemia excluding CLL under a two-year lag in a linear model ( n=264)...... 253 Table 6-7. Tests for interaction between radiation dose and sex, race, birth cohort, hire year, SES, and study group for Leukemia excluding CLL ...... 254 Table 6-8.Tests for confounding of radiation dose effect by sex, race, birth cohort, hire year, benzene exposure, and SES in leukemia excluding CLL...... 255 Table 6-9. The effect of exposure lag on the excess relative risk of non-CLL leukemia from low-LET radiation dose ( n=264 cases) ...... 257 Table 6-10. Analysis of radiation and non-CLL leukemia risk by exposure windows of years before the cutoff date ( n=264 cases) ...... 257 Table 6-11. The effect of exposure lag on the excess relative risk of AML from radiation dose ( n=147 cases) ...... 258 Table 6-12. Analysis of radiation and AML risk by exposure windows of years before the cutoff date (n=147 cases) ...... 258 Table 6-13. The effect of exposure lag on the excess relative risk of CML from radiation dose ( n=52 cases) ...... 259 Table 6-14. Analysis of radiation and CML risk by exposure windows of years before the cutoff date (n=52 cases) ...... 259 Table 6-15. The effect of exposure lag on the excess relative risk of CLL from radiation dose ( n=74 cases) ...... 260 Table 6-16. Analysis of radiation and CLL risk by exposure windows of years before the cutoff date ( n=74 cases) ...... 260 Table 6-17. Summary of final linear modeling results for low-LET radiation and leukemia mortality ..... 262 Table 6-18. Sensitivity analysis of non-CLL leukemia risk including from high and low-LET radiations combined (lagged two-years), adjusted for sex, race, and hire year...... 263 Table 6-19. Model fit characteristics for relation between non-CLL leukemia and low-LET radiation .... 266

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Table 6-20. Leukemia Excess Relative Risk (ERR) per 10 mGy exposure by subtype and by study ...... 278 6-21. Comparison of the results on Non-CLL leukemia risk obtained in the current study to the previous study...... 280

List of Figures Figure 2-1. U. S. benzene exposure control levels for the period 1940-2005 [Capleton and Levy 2005; Infante 1978]...... 38 Figure 3-1. Cohort assembly and vital status followup flowchart. Additional searches against the California Death Index and the IRS database were used to confirm alive status of non-matches to NDI...... 52 Figure 3-2. Annual limits on radiation dose to blood forming organs used by study facilities at time of start-up of radioactive work...... 56 Figure 3-3. Hanford plutonium (solid blue line, in MT) and tritium (dashed red line, in kg) production (1945-1987)...... 59 Figure 3-4. Hanford site employment numbers (1944-2000) ...... 60 Figure 3-5. Employment at INL (1951-1990) ...... 62 Figure 3-6. LANL and Zia Company person-hours worked per year (19450-1985) ...... 64 Figure 3-7. PNS Employment (1950-2002) ...... 73 Figure 3-8. SRS plutonium production reactor power in Mw-d (1955-1987) ...... 77 Figure 3-9. SRS employment, excluding subcontractors (1951-1990)...... 78 Figure 4-1. Histograms of the distribution of leukemia deaths (n=369) by decade (Panel A) and by birth cohort...... 101 Figure 5-1. Average annual dose and number of monitored workers in the base cohort...... 134 Figure 5-2. Hanford plutonium bioassay samples (1950-2000)...... 178 Figure 5-3. INL plutonium bioassay samples (1952-1998) ...... 181 Figure 5-4. Number of plutonium bioassay samples at LANL (1945-2005)...... 182 Figure 5-5. Number of plutonium bioassay samples at ORNL (1950-2000)...... 183 Figure 5-6. The number of tritium urinalysis samples at INL between the years 1974 and 1998...... 192 Figure 5-7. Tritium bioassay samples at ORNL (1950-2000) ...... 195 Figure 5-8. Subjects assigned benzene exposures by exposure year ...... 237 Figure 5-9. Average exposure score among subjects by exposure year ...... 237

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Figure 6-1. Plot of dose-response models for low-LET dose and the relative risk of leukemia, excluding CLL. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=264 cases)...... 264 Figure 6-2. Plot of dose-response models for low-LET dose and the relative risk of leukemia, excluding CLL and restricted to doses < 100 mGy. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=244 cases) ...... 265 Figure 6-3. Plot of dose-response models for low-LET dose and the relative risk of leukemia, excluding CLL, and restricted to doses from 10 mGy to 100 mGy. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=90 cases) ...... 266 Figure 6-4. Plot of dose-response models for low-LET dose and the relative risk of leukemia, excluding CLL. Red= three-piece linear spline model (knots= 8, 46); blue= restricted cubic spline (RCS), 4 knots (n=264 ...... 267 Figure 6-5. Plot of dose-response models for low-LET dose and the relative risk of AML. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=147 cases) ...... 269 Figure 6-6. Plot of dose-response models for low-LET dose and the relative risk of CML. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=52 cases) ...... 270 Figure 6-7. Plot of dose-response models for low-LET dose and the relative risk of CLL. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=74 cases)...... 271 Figure 6-8. Plot of dose-response for low-LET dose in the 5-10 exposure window and the relative risk of non-CLL leukemia. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=264 cases)...... 272

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ABBREVIATIONS

ACGIH American Conference of Government Hygienists AEC U.S. Atomic Energy Commission AES Annual Exposure Score ALARA As Low As Reasonably Achievable ALL Acute Lymphocytic Leukemia AML Acute Myeloid Leukemia ANL Argonne National Laboratory ANP Aircraft Nuclear Propulsion AP anterior -to -posterior ATLAS Automatic Thermoluminescent Analyzer System AWE Atomic Weapons Employers BEIR National Academy of Sciences Committee on the Biological Effects of Ionizing Radiation BMI body mass index BNFL British Nuclear Fuels (UK) BNL Brookhaven National Laboratory Bq becquerel BSR Bulk Shielding Reactor CEA French Alternative Energies and Atomic Energy Commission CEDS Centralized External Dosimetry System CEF Critical Experiments Facility CES cumulative exposure score CI confidence interval CF Central Facilities (INL) CLL Chronic Lymphocytic Leukemia CML Chronic Myeloid Leukemia CMR Chemical and Metallurgical Research (Building) CNDR Canadian National Dose Registry COD cause of death DCAS (NIOSH) Division of Compensation Analysis and Support DHHS U.S. Department of Health and Human Services

DMDTC dimethyldithiocarbamate DNA deoxyribonucleic acid DOB date of birth DOD U.S. Department of Defense DOE U.S. Department of Energy DOELAP U.S. Department of Energy Laboratory Accreditation Program DOL U.S. Department of Labor dpm disintegrations per minute DSB double strand break DSHEFS (NIOSH) Division of Surveillance, Hazard Evaluations, and Field Studies DTPA diethylenetriaminepentaacetic acid DWPF (SRS) Defense Waste Processing Facility EBR Experimental Breeder Reactor ECF Expended Core Facility EEOICPA Energy Employee Occupational Illness Compensation Program Act EPA U.S. Environmental Protection Agency ERR excess relative risk ES effect size ESE entrance skin exposure eV electron volt FID focus -to -image distance FMPC Feed Materials Production Center FSD focus -to -source distance GM geometric mean GSD geometric standard deviation Gy Gray HAM High Activity Moderator HAN Hanford Site HDDRS Hanford Declassified Document Retrieval System HEHF Hanford Environmental Health Foundation HERB Health -related Energy Research Branch HEW Hanford Engineering Works (a.k.a. Hanford Site, Hanford)

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HFEF Hot Fuels Examination Facility HFIR High Flux Isotope Reactor HMS Hanford Health and Mortality Study

Hp(10) Individual equivalent dose, penetrating to a depth of 10 mm HPAREH Health Protection Annual Radiation Exposure History HPD Health Physics Department HPIL Health Physics Instrument Laboratory HPRR Health Physics Research Reactor HRE Homogeneous Reactor Experiment

HT Equivalent dose to tissue. For this study tissue (T) is red bone marrow HTO tritium oxide (tritiated water) HVL half value layer IARC International Agency for Research on Cancer ICD International Classification of Disease ICP Idaho Cleanup Project ICPP Idaho Chemical Processing Plant ICRP International Commission on Radiological Protection ICRU Internati onal Commission on Radiation Units IMBA Integrated Modules for Bioassay Analysis INEEL Idaho National Engineering and Environmental Laboratory INSERM French National Institute of Health and Medical Research ITPF SRS In -Tank Precipitation Facility kBq kilo becquerel (10 3 Bq) keV kilo electron volts (10 3 eV) LACEF Los Alamos Critical Experiments Facility LAMPRE Los Alamos Molten Plutonium Reactor LANL Los Alamos National Laboratory (formerly LASL) LAPRE Los Alamos Power Reactor Experiment LASL Los Alamos Scientific Laboratory LBL Lawrence Berkeley Laboratory LCCS leukemia case control study LET linear energy transfer

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LHSTC Lucas Heights Science and Technology Centre (Australia) LITR Low -intensity Test Reactor LLNL Lawrence Livermore National Laboratory LNT linear -no -threshold LRT Likelihood ratio test LSC liquid scintillation counter LSD least significant difference LSS Life Span Study of Japanese Atomic Bomb Survivors MDC minimum detectable concentration MDS Myelodysplastic Syndrome (MDS) MED Manhattan Engineering District MeV mega electron volt mg ∙m -3 milligrams per cubic meter of air mGy milligray (10 -3 Gy) ml milliliter (10 -3 liters) MLE maximum likelihood estimation MMP Miamisburg Mound Plant MPD Myeloproliferative Disorder (MPD) MPF Moderator Processing Facility MSRE Molten Salt Reactor Experiment mSv milli -sievert (10 -3 Sv) MT metric ton MTR Material Test Reactor MUD INL Master Update Dump M&O management and operations NA not applicable NASA U. S. National Aeronautics and Space Administration NBS National Bureau of Standards NCHS National Center for Health Statistics NCRP National Council on Radiation Protection and Measurements ND not detected NHL non -Hodgkin’s lymphoma

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NIOSH National Institute for Occupational Safety and Health NNRW National Registry for Radiation Workers (UK) NR U. S. Naval Reactors (NAVSEA08) NRAD Neutron Radiography Facility (INL) NRC National Research Council (of the National Academies) NRCL U.S. Nuclear Regulatory Commission Licensees NRF Naval Reactor Facility NRTS U.S. National Reactor Testing Station NTA Kodak Nuclear Track Type A NTS Nevada Test Site NVLAP National Voluntary Laboratory Accreditation Program OBT organically bound tritium compound OD optical density ODU INL Operation Dosimetry Unit OERP NIOSH Occupational Energy Research Program OR odds ratio ORAU Oak Ridge Associated Universities ORGDP Oak Ridge Gaseous Diffusion Plant (a .k.a. K-25) ORISE Oak Ridge Institute for Science and Education ORNL Oak Ridge National Laboratory (a.k.a. X -10, Clinton Laboratory) ORR ORNL Research Reactor OSHA U.S. Occupational Safety and Health Administration OSTI DOE Office of Scientific and Technical Information OWR Omega West Reactor PA posterior -to -anterior PAH polynuclear aromatic hydrocarbons PBPK physiologically -based pharmacokinetic PCA Pool Critical Assembly PCXMC PC program for X -ray Monte Carlo PEL permissible exposure level PFG photofluorographic chest x -ray PFX photofluorographic chest x -ray machine (Westinghouse nomenclature)

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PIC pocket ion -chamber PMR proportionate mortality ratio PNS Portsmouth Naval Shipyard PoBe polonium -beryllium PP probability plot ting ppb parts per billion PPE personal protective equipment ppm parts per million PRTR Plutonium Recycle Test Reactor PUREX plutonium uranium extraction PVC polyvinyl chloride RaLa radiolanthanum ( 140 La) RBE relative biological effectiveness RDS INL Radiation Dosimetry System REIRS USNRC Radiation Exposure Information Reporting System REL recommended exposure level REMS DOE Radiation exposure Monitoring System REX Hanford Site Radiological Exposure database RFP Rocky Flats Plant RR relative risk RTF Replacement Tritium Facility RW Heavy Water Rework Unit RWMC Radioactive Waste Management Complex (INL) SAS Statistical Analysis System SES socioeconomic status SIR standardized incidence ratio SMR standardized mortality ratio SNL Sandia National Laboratory SPEERA Secretarial Panel for the Evaluation of Epidemiologic Activities SRP Savannah River Plant SRR standardized rate ratio SRS Savannah River Site (formerly SRP)

SSA U.S. Social Security Administration SSB single strand break SSN social security number STEL short -term exposure limit SWEPP Stored Waste Examination Pilot Plant (INL) TA Technical Area (LANL) TAN Test Area North (INL) TBP tributyl phosphate TEC Tennessee Eastman Corporation TED track etch dosimeter TEPC tissue equivalent proportional counter TIOA tri -isooctylamine TLD thermoluminescent dosimeter TLND thermoluminescent neutron dosimeters TLV threshold limit value TPF Technical Purification Facility TRA Test Reactor Area TREAT Transient Reactor 2 TRIGA Training Research and Isotope Atomic (INL) TSNA tobacco -specific N -nitrosamines TSR Tower Shielding Reactor TTA thenoyl trifluoroacetone TWA time -weighted average TWF time -dependent weighting factor UCC -ND Union Carbide Corporation, Nuclear Division UCOD underlying cause of death UK United Kingdom UMTRA Uranium Mill Tailings Remedial Action UNSCEAR United Nations Scientific Committee on the Effects of Atomic Radiation USNRC U. S. Nuclear Regulatory Commission WHO World Health Organization WLM working level month

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WRX work -related x-ray examination WVDP West Valley Demonstration Project Y12 Y12 Plant Zia Zia Company ZnS zinc -sulfide ZPPR Zero Power Plutonium Reactor

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1.0 Introduction

According to estimates by the United Nations Scientific Committee on the Effects of Atomic

Radiation (UNSCEAR), over 11 million workers worldwide are exposed to ionizing radiation each year

[UNSCEAR 2000]. In the U.S., approximately 300,000 workers are monitored annually for radiation exposure at U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (USNRC) licensee facilities combined [DOE 2007; USNRC 2007]. Sufficient evidence from a number of previous studies exists that unequivocally establishes ionizing radiation as a human carcinogen [IARC 2000]; thus, worker protection is needed to mitigate cancer risks within radiation-related occupations. Current worker protection is typified by risk education programs, work practices that maintain exposures as low as reasonably achievable (ALARA), and strict adherence to protection standards promulgated by the various federal agencies [e.g., U.S. Department of Labor (DOL), DOE, and the USNRC]. For over fifty years, risk models that form the cornerstone of these standards have not relied on worker studies but are based almost exclusively on the Life Span Study (LSS) cohort of Japanese atomic bomb survivors. The question remains whether current safety benchmarks, extrapolated from risk patterns observed in a population with highly specialized acute exposures, are adequate for mitigating risks from protracted and fractionated low-dose radiation exposures found in the workplace.

In 1998, the National Research Council of the National Academy of Sciences’ Committee on the

Biological Effects of Ionizing Radiation (BEIR) concluded that sufficient data were available to warrant a reexamination of the health effects of low levels of ionizing radiation. In addition to the LSS, the

Committee reviewed data from medically, occupationally, and environmentally exposed populations for information suitable for risk estimation. Although a wide array of information was reviewed, the

Committee again relied mainly on LSS data to form the risk estimates that were published in 2006 [NRC

2005]. They found that most occupational studies lacked the precision necessary for the projection of population-based risks. However, the Committee acknowledged that occupational studies are, in

principle, preferred for estimating the effects of low-dose protracted exposures and recommended the continued study of nuclear industry workers with emphasis on improved epidemiologic methods that minimize uncertainty of study results.

The central purpose of this study is to develop improved, occupationally-based, quantitative risk estimates for leukemia radiogenicity that may buttress future recommendations on radiation protection. To this end, the research consisted of a nested case-control study of leukemia mortality in a pooled cohort of 105,245 U.S. nuclear-industry workers. Leukemia, excluding the chronic lymphocytic subtype, is considered to be among the cancers most susceptible to induction by ionizing radiation exposure. However, the evidence of radiation leukemogenicity in humans stems mainly from moderate- to high- dose and dose rate studies. Most individual studies involving occupational cohorts have been less informative because of poor study power, limited followup, and large uncertainty in exposure assessment, although some recent studies involving large cohorts have suggested increased leukemia risk from protracted low-dose radiation exposure. This study is larger than previous studies and incorporates additional follow-up and improvements in dose reconstruction.

The study’s central purpose was accomplished by testing the null hypothesis of an association between leukemia mortality and the duration and intensity of protracted radiation exposures in a pooled cohort of U.S. nuclear workers. The study has a single aim: To directly estimate the exposure- response relation between ionizing radiation exposure and leukemia mortality in the cohort, thereby enabling the management of cancer risk based on information that is relevant to workplace exposure conditions. This effort, in conjunction with information from other epidemiologic studies, will be used as a sound basis for future recommendations for occupational radiation protection.

The protection of workers exposed to ionizing radiation is largely dependent on our understanding of the risk associated with exposure. The findings of this research study have the

potential to reduce morbidity and mortality among the 1.5 million workers occupationally exposed to ionizing radiation in the U.S. by influencing U.S. protection standards. The findings have the potential to influence international exposure standards as well, providing protection to the estimated 11 million workers exposed to radiation each year.

2.0 Leukemia

2.1 Description

Leukemia is a term used to broadly describe a heterogeneous group of malignant neoplasms of hematopoietic tissue with relatively unknown etiology. Malignant transformation in leukemia typically occurs in pluripotent stem cells, but can involve a committed stem cell with more limited capacity for differentiation. Leukemic cells present with abnormal proliferation and diminished apoptosis, which leads to a suppression of normal blood cell formation and organ infiltration by malignant cells.

Leukemia is generally grouped by progression (acute or chronic) and cell lineage (myeloid, lymphoid). There are four major leukemia subtypes: chronic lymphoid leukemia (CLL), acute lymphoid leukemia (ALL), chronic myeloid leukemia (CML), and acute myeloid leukemia (AML). Within these subtypes, there are several subcategories. Finally, there are a number of rare leukemia subtypes that fall outside of major categories and are otherwise specified.

Acute leukemias (e.g., AML and ALL) are rare diseases of the leukocytes and their precursors, which are characterized by immature, abnormal cells in the bone marrow, peripheral blood, and parenchymatous organs. Patients with acute leukemia present with anemia, which is usually severe, and an absence of functioning granulocytes and thrombocytopenia. The disease rapidly progresses resulting in death soon after diagnosis if left untreated. Acute leukemias are the leading cause of cancer death in children and in persons who are less than 39 years of age [Deschler and Lubbert 2006]. Chronic leukemias (e.g., CML and CLL) are a proliferation of mature appearing cells in marrow, peripheral blood, and various organs. Cell maturation is sufficient to allow normal function in some abnormal cells, thus slowing the clinical course of the disease. Chronic leukemia is typically asymptomatic in its early stages, followed by a slow progression that may persist for months or many years.

Myeloid Leukemias (e.g., AML, CML) are disorders that originate in pluripotential precursor cells that would normally mature to red blood cells, monocytes, polymorphonuclear neutrophils, and platelets [Linet et al. 2006].

Lymphoid leukemias originate in cells of lymphoid lineage and are composed of either hematopoietic stem cells (mostly in childhood ALL) or immature precursor cells (mostly in adult ALL) that are in the process of rearranging their immunoglobulin or T-cell receptor genes (i.e., precursor disease) or functional B-cells or T-cells in which the arrangement of immunoglobulin or T-cell receptor genes is complete (i.e., peripheral disease). Adult and pediatric forms of ALL are precursor diseases, while the category of peripheral diseases includes CLL and non-Hodgkin Lymphoma. The pathogenesis of lymphoid leukemias is marked by significant genetic instability common to the lymphocyte lifecycle

[Linet et al. 2006]. Translocations involving immunoglobulin heavy chain and T-cell receptor genes are relatively common, which can lead to deregulation of oncogenes in some cases.

2.2 Acute Lymphocytic Leukemia (ALL)

ALL is characterized by the malignant proliferation of lymphoid blast cells in bone marrow [Linet et al. 2006]. There are three major subtypes: 1) precursor B-cell ALL, 2) precursor T-cell ALL, and 3)

Burkett-cell leukemia. Among the acute leukemias, ALL is the most common cancer diagnosis in children and accounts for 74% of all leukemia cases in persons less than 19 years of age [Jemal et al. 2010]. In contrast, less than 40% of ALL cases occur in ages 20+ years. Pediatric ALL is slightly elevated in boys compared to girls. The sex difference is more pronounced in adults; the ALL male to female ratio is about 1.4:1. In the U.S., there is about a two-fold increase in white ALL cases compared to non-white cases. The Philadelphia chromosome is the most common genetic abnormality in ALL, occurring in about

3-5% of pediatric patients [Arico et al. 2000] and 20-50% in adult ALL cases [Secker-Walker et al. 1991].

This lesion is the result of a reciprocal translocation between chromosome 9 and 22 [i.e.,

t(9;22)(q34:11)]. The disease course is particularly aggressive in Philadelphia chromosome-positive patients (Ph+). Those who receive only conventional chemotherapy have disease free survival rates of not more than 10% [Stock 2010].

2.3 Acute Myeloid Leukemia (AML)

AML is essentially a myeloid neoplasm comprised of precursor cells with impaired maturation.

Differential diagnosis may be achieved by assessing the maturation of malignant cells, thus the blast percentage is a practical tool for categorizing myeloid neoplasms. In the classification scheme of the

World Health Organization (WHO), AML is a myeloid neoplasm with 20% or more blasts in the peripheral blood or bone marrow [Vardiman et al. 2009]. AML can occur de novo , or can evolve from a myelodysplastic syndrome (MDS) or myeloproliferative disorder (MPD), such as CML. AML diagnosis is uncommon in patients before the age of 40, but has the highest leukemia mortality compared to other leukemia subtypes. Adult AML is slightly higher in males than females and is significantly elevated in U.S. males compared to males in all other countries [Deschler and Lubbert 2006].

2.4 Chronic Myeloid Leukemia (CML),

CML, or chronic myelogenous leukemia, is a clonal, myeloproliferative disease with origin in a pluripotent bone marrow stem cell. CML accounts for 10-15% of all leukemias and is diagnosed mainly in adults (rarely in children). CML incidence is higher in men than in women, but women have a survival advantage. There is little evidence of geographic or ethnic differences in CML risk.

CML is consistently associated with the Philadelphia chromosome. ABL1 (on chromosome 9) and

BCR (on chromosome 22) have been identified as the specific genes involved in the Ph translocation and the fusion gene (BCR-ABL1) that is present in nearly all CML patients [Tefferi et al. 2009]. This fusion gene encodes an abnormal protein that activates signaling pathways leading to abnormal proliferation in marrow that ultimately causes the clinical and morphologic manifestations of CML [Vardiman 2009].

Whether or not BCR-ABL1 is the initiating lesion or is sufficient alone to cause CML in humans is still a

matter of much debate [Cross et al. 2008]. Nevertheless, the identification of this underlying genetic

defect in has led to targeted molecular therapy for CML-patients. The results of these therapies have

been remarkable, with nearly 80% of CML patients achieving durable clinical and cytogenetic remissions

[Vardiman 2009].

2.5 Chronic Lymphocytic Leukemia (CLL)

CLL is a clinically heterogeneous disease with origin in B lymphocytes that can vary in activation, maturation state, or cellular subgroup [Chiorazzi et al. 2005; Ghia et al. 2007]. The disease is characterized by the accumulation of CD5+ B lymphocytes in the peripheral blood, bone marrow, and secondary lymphoid organs. CLL accounts for about one- third of all leukemias in Western populations and is extremely rare in people under the age of 40. CLL incidence is slightly higher in men than in women [Ghia et al. 2007].

Although considered among the leukemias for this study, CLL, along with other peripheral malignances such as B-cell prolymphocytic leukemia and leukemic reticuloendotheliosis (hairy cell leukemia), share many traits (e.g., cytology, immunophenotype, histopathology) with other lymphoproliferative disorders that form a heterogeneous grouping known as the non-Hodgkin lymphomas (NHL). As such, recent classification schemes include CLL among the NHL group, along with small lymphocytic lymphoma (SLL), its non-leukemic equivalent [Harris et al. 1994]. In fact, most now agree that separate classification of lymphoid leukemias and lymphoma is merely artificial [Jaffe et al.

2008]. Nevertheless, CLL was included sparingly in previous studies of radiation-induced leukemia and

CLL radiogenicity has been a matter of recent debate [Silver et al. 2007]. Main analyses for this study will consider both leukemia and leukemia excluding the CLL subtype.

2.6 Leukemia Trends

World-wide leukemia mortality is highest in populations of Western Europe, Oceania, North

America, Israel, and Costa Rica, where age-adjusted mortality in males ranges from 4.8 to 7.4 per

100,000 person-years. The lowest rates are observed in Latin American and Asian populations, where the mortality rates for males range from 3.7 to 4.5 per 100,000 person-years. Leukemia risk is highest in males in all races. Mortality rates have slightly declined in white populations and slightly increased in non-Whites since the 1950s. Mortality rates have declined in children and adolescents (0-19 years) and young adults (20-44 years) since 1960. The international incidence of total leukemia in children, adolescents, young adults, and middle-aged persons (45-64 years) has remained relatively stable since

1960; however, a slight increase was observed in the elderly (65+ years) from 1960 to the early 1970s

[Linet et al. 2006].

The National Cancer Institute estimated the 2007 leukemia prevalence in the U.S. at 244,272

(137,398 men and 106,874 women) and predicted that 43,050 new leukemia cases and 21,840 leukemia deaths will occur in the year 2010 (Table 2-1) [Altekruse et al. 2010]. Approximately one in 77 persons will be diagnosed with leukemia during their lifetime. The U.S. average age-adjusted incidence of leukemia in the period 1975-2007 was 12.97 per 100,000 person-years. The age-adjusted mortality in the same period was 7.79 per 100,000 person-years. Leukemia incidence per 100,000 person-years was higher in males (17.09) than in females (9.98) and Whites (13.56) than in Blacks (10.70) [Altekruse et al.

2010]. Similarly, Asian Americans have lower age-adjusted leukemia incidence than U.S. Whites, particularly for CLL [Pang et al. 2002; Yamamoto and Goodman 2008]. Hispanics generally have lower age-adjusted leukemia incidence than non-Hispanics, although there is some evidence of increased risk of ALL, CML in the very elderly (85+ years), and acute promyelocytic leukemia (a subtype of AML) in

Hispanics [Matasar et al. 2006; Pullarkat et al. 2009; Yamamoto and Goodman 2008].

The age-adjusted incidence per 100,000 person-years was highest for CLL (4.48), followed by

AML (3.43), CML (1.75), and ALL (1.33). However, the highest mortality (2010 estimates) was from AML

(8,950 deaths), followed by CLL (4,390 deaths), ALL (1,420 deaths), and CML (440 deaths). The overall trend in leukemia has remained fairly constant over recent years; the annual percent change in leukemia mortality over the period 1950-2007 was -0.2 percent. However, leukemia incidence among Blacks increased slightly from 1973 to 1979 and from 1980 to 1986 before declining slowly. Temporal risk differences by age group have also been observed; the leukemia mortality risk in persons who were 70 years or older at time of death has steadily increased over the last several decades [Levi et al. 2000]. The increasing trend in the elderly may be due to improvements in diagnosis and death certification in older patients. From 1999 to 2000, the 5-year relative survival among leukemia patients was 56% compared to only 10% survival in the period 1950-1954. The median age at diagnosis and death are 66 years and 74 years, respectively.

Table 2-1. U.S. Trends in leukemia by subtype [Altekruse et al. 2010].

ALL AML CLL CML Median age at diagnosis (years) 13 67 72 65 Percent diagnosed by age group Birth-<20 60.7 6.1 0.1 2.5 20-<35 10.3 6.5 0.2 7.4 35-<45 6.2 6.7 1.5 10.1 45-<55 6.5 11.3 9.0 13.3 55-<65 5.8 15.1 20.0 15.0 65+ 10.5 54.3 69.2 51.7 Median age at death (years) 49 72 79 74 Age adjusted incidence rate (10 -5pyrs) 1.6 3.5 4.2 1.5 Age adjusted mortality rate (10 -5pyrs) 0.5 2.8 1.5 0.4 5-year relative survival (percent) 65.2 23.6 78.4 56.8 Lifetime risk (percent) 0.13 0.38 0.48 0.16

2.7 Risk Factors

2.7.1 Lifestyle factors

Several lifestyle factors, such as smoking habits, diet, obesity, and alcohol consumption, have been studied as potential modifiers of leukemia risk; however, the evidence from most of these studies is equivocal. One notable exception is cigarette smoking, whereby there is sufficient evidence to conclude that AML is moderately associated with increased smoking habits.

Information on several lifestyle factors and their association with leukemia are discussed in this section. In summary, leukemia risk is largely unaffected by lifestyle factors. When associations are noticeable, they appear weak and preferential to acute leukemia, particularly AML.

2.7.1.1 Alcohol consumption

Information on the effects of alcohol consumption on leukemia risk is sparse and study results are equivocal. Nevertheless, there is some evidence of an inverse relation between drinking alcohol and leukemia. In a case control study of Los Angeles residents, Pogoda et al. [2004] observed decreased AML risk in persons drinking >10 grams of alcohol per day compared to non-drinkers, although estimates were imprecise [odds ratio (OR)=0.8; 95% CI: 0.4, 1.6]. In contrast, Wakabayashi et al. [1994] found no evidence of a relation between alcohol consumption and AML in an earlier case-control study of

Japanese patients. Rauscher et al. [2004] conducted a case-control study of acute leukemia risks in persons recruited in the U.S. and Canada and observed protective effects of light (1-<5 drinks per week) to moderate (6-<8 drinks per week) beer drinking compared to infrequent (<1 drink per week) drinking persons [relative risk (RR)=0.58; 95% CI: 0.44, 0.76]. Interestingly, the same study found a positive association for moderate wine drinking (RR=2.1; 95% CI: 1.2, 3.8) The largest study was conducted by

Klatsky et al. [2009], who examined the effects of alcohol consumption on the risk of hematologic malignancies in a multiethnic population ( n=126,293) and found reduced risks of myeloid leukemia

(RR=0.4; 95% CI: 0.2, 0.9; n=170) and lymphocytic leukemia (RR=0.4; 95% CI: 0.2, 0.9; n=149) in persons reporting greater than three drinks per day compared to lifelong abstainers. An inverse dose response was evident ( P for trend= 0.01) for myeloid leukemia only. Models were adjusted for age, sex, ethnicity, body mass index (BMI), marital status, and smoking habits. There was no evidence of risk differences by beverage choice.

2.7.1.2 Diet

Information on dietary risk factors is sparse, but a few large prospective studies have suggested increased risk in acute leukemias from meat consumption and a protective effect from eating fruits and vegetables. Ross et al. [2002] examined adult leukemia incidence in a prospective study of women

(n=35,221) followed from 1986 through 1999 and found an inverse relation between vegetable consumption (medium vs. low) and leukemia risk (RR=0.56; 95% CI: 0.36, 0.99) although there were few cases ( n=138). Liu et al. [2009] found that Han Chinese between the ages of 2 and 20 who consumed cured or smoked meat and fish more than once a week had elevated acute leukemia risks (OR = 1.74;

95% CI: 1.15, 2.64) but higher intake of vegetables resulted in lower risks (OR = 0.55; 95% CI: 0.37, 0.83).

Ma et al. [2010] examined diet as a risk factor for AML in a large prospective study ( n=491,163) that was followed from 1996 through 2003. They found that a higher intake of meat was associated with increased AML risk [hazard ratio (HR) =1.45; 95% CI: 1.02, 2.07 for the fifth vs. first quintile]. However, there was no association found between AML and fruit or vegetable intake.

2.7.1.3 Smoking

Tobacco smoke is estimated to contain nearly 4,000 chemical compounds, including 69 known carcinogens [IARC 2004]. Among the carcinogens, there are species of polynuclear aromatic hydrocarbons (PAHs), heterocyclic hydrocarbons, volatile hydrocarbons, nitrohydrocarbons, aromatic amines, N-heterocyclic amines, N-nitrosamines, and aldehydes. Because of their carcinogenic potency,

recent studies have focused mainly on compounds such as benzo-[a]pyrene, tobacco-specific N- nitrosamines (TSNA), especially N′-nitrosonornicotine (NNN) and 4-(N-nitrosomethylamino)-1-(3-

pyridyl)-1-butanone (NNK) and aromatic amines, especially 4-aminobiphenyl (4-ABP) [IARC 2004].

Furthermore, cigarette smoke contains benzene. It is estimated that smoking accounts for half the

benzene exposure in the U.S. and about 90% of the benzene exposure to smokers [Wallace 1996].

Benzene, a known carcinogen, is estimated to contribute from 8 to 48% of smoking-induced leukemia

deaths and from 12 to 58% of smoking-induced AML deaths [Korte et al. 2000].

The evidence of an association between cigarette smoke and AML is equivocal, but nevertheless

sufficient for the U.S. Surgeon General [CDC 2004] and the International Agency for Research on Cancer

[IARC 2004] to conclude that there is a causal relation. Their findings were based primarily on several

studies that suggested a weak-to-moderate association between smoking and AML, with a relative risk

of about 1.4 in smokers and approaching 2.0 for heavy smokers [Brownson et al. 1993; Garfinkel and

Boffetta 1990; Kasim et al. 2005; Siegel and Faigeles 1996; Thomas and Chelghoum 2004]. Causal

mechanisms are still unknown but there is evidence of associations between smoking and certain

chromosomal aberrations (e.g., t(8;21), (q22;q22) -5/5q-, -7/7q-, and +8) found in AML, although

estimates were imprecise [Bjork et al. 2009; Crane et al. 1996; Davico et al. 1998; Moorman et al. 2002;

Sandler and Collman 1987]. A recent meta-analyses of 10 case-control studies examined the association

between smoking and myelodysplastic syndromes (MDS), a potential precursor to AML, and reported an

aggregate OR= 1.45 (95% CI: 1.21, 1.74), with heterogeneity among studies ( P= 0.05), but no evidence of publication bias [Du et al. 2010]. However, it is not clear if the relation between smoking and AML is

stronger in one AML phenotype, such as a MDS transformation, relative to another, say de novo AML.

The evidence of a relation between smoking and leukemia subtypes other than AML is much less

compelling. Garfinkel and Boffetta [1990] examined leukemia mortality in two large prospective cohort

studies ( n=437,197 men and 588,148 women enrolled in Study 1; n=489,696 men and 622,488 women

enrolled in Study 2). Although there was evidence of an association between myeloid leukemias and smoking in men in both cohorts [standardized mortality ratios (SMRs) = 2.44 and 1.32; P < 0.05], they found no evidence of elevated risk of lymphatic leukemias by smoking status in either cohort (SMRs=

1.02 and 1.24; P > 0.05). In contrast, Linet et al. [1991] observed a three-fold increase in lymphatic leukemia risk (RR= 2.7; 95% CI: 0.9, 8.3) in cigarette smokers compared to non-smokers in a cohort of white males ( n=17,633) followed from 1966 through 1986. However, there were only 29 lymphatic leukemia cases in the cohort. Stagnaro et al. [2001] examined the relation between smoking and leukemia risk by subtype in a population-based case-control study of adults residing in 12 areas of Italy

(n=649 incidence cases). Using non-smokers as referent, significant elevations were not observed for

ALL, CML, or CLL patients categorized by status (ever, former, current), numbers of cigarettes smoked

(1-<10, 10-<10, 19+) or by smoking duration in years. Similar null results were obtained by Kasim et al.

[2005] in a population-based case-control study of Canadian adults ( n=1,069 incidence cases). Recently,

Richardson et al. [2008] conducted a population-based case-control study of German residents (n=470 incidence cases) and also found little evidence that CML and ALL were associated with smoking status

[never(referent), ex-smoker, current smoker], cumulative exposure (pack-decade), or time since exposure (2-<10, 10-<20, and 20+ years prior). However, CLL exhibited a slight positive association in current smokers (OR=1.33; 95% CI: 0.86, 2.06) and in smoking periods 2-<10 (OR=1.29; 95% CI: 0.66,

2.52) and 10-<20 (OR=1.35; 95% CI: 0.71, 2.55) years prior to diagnosis, although estimates were imprecise.

Although few studies have evaluated the joint effects of smoking and ionizing radiation on leukemia risk, previous studies of nuclear workers have not shown evidence of a strong bias in risk estimates that could be attributed to smoking patterns [Schubauer-Berigan et al. 2007a; Schubauer-

Berigan et al. 2007b; Schubauer-Berigan et al. 2005; Yiin et al. 2005]. In particular, Schubauer-Berigan et al. [2007a; 2007b] did not observe confounding or leukemia risk modification by smoking category

(never, former, or current) in the recent study of leukemia risk in Hanford, LANL, ORNL, PNS, and SRS workers. The lack of smoking effects in previous studies is expected considering that radiation exposure and smoking would need to be highly correlated to account for even a modest effect on the dose- response trend given the small association between smoking and leukemia [Axelson and Steenland

1988; Siemiatycki et al. 1988b]. Even in occupational studies examining smoking-related outcomes, risk estimates are unlikely to be strongly confounded by smoking when comparing exposures within the same working population [Axelson 1989; Axelson and Steenland 1988; Kriebel et al. 2004; Siemiatycki et al. 1988a; Siemiatycki et al. 1988b].

2.7.1.4 Socioeconomic Status (SES)

It is well understood that social class is associated with many health outcomes largely because of disparities in environmental pollutants, available medical care, and lifestyle [Kogevinas et al. 1997;

Rothman et al. 2008]. Furthermore, there is strong evidence that the risk of certain cancers in men (e.g.,

respiratory malignancies and cancers of the mouth, pharynx, esophagus and stomach) increases with

decreasing social class. Certain health risk behaviors that are also linked to social class, such as smoking

and alcohol consumption, are known risk factors for these cancers. However, the association between

these behaviors and leukemia are weak in comparison, which may explain why the evidence of social

class trends in leukemia is equivocal [Bhayat et al. 2009; Kogevinas et al. 1997]. It is also important to

note that the distinction in social class within a specific working population (i.e., U.S. nuclear workers) is

far less pronounced compared to the general population and any effect from uncontrolled confounding

of leukemia risk examined in internal comparisons is likely to be small [Checkoway et al. 2004].

Nonetheless, social class (interchangeably SES) has been widely used in occupational epidemiologic

studies of nuclear workers to indirectly measure health risk behaviors (e.g., smoking, alcohol

consumption, diet, and exercise) that may influence risk estimates [Cardis et al. 2007; Schubauer-

Berigan et al. 2007a; Schubauer-Berigan et al. 2005; Yiin et al. 2005].

2.7.2 Environmental factors

Several of the leading environmental factors in leukemia research are discussed in this section.

The term environmental factor refers to risk factors found in the environment in which one works or lives, thus occupational risk factors are discussed herein. Comprehensive information on potential environmental leukemogens is provided in several thorough reviews [Clapp et al. 2008; Descatha et al.

2005; Irigaray et al. 2007; Lamm et al. 2009; Schubauer-Berigan and Wenzl 2001; Smith 2010; Van

Maele-Fabry et al. 2008; Zhang et al. 2009]. In general, the evidence of an association between these factors and leukemia in humans is equivocal, with benzene and low-LET ionizing radiation being notable exceptions. Only benzene and ionizing radiation are universally acknowledged as leukemogens.

Populations are exposed ubiquitously to low levels of both leukemogens due to pervasive contamination

(primarily benzene) and the availability of natural radiation sources. Furthermore, benzene saw widespread use in many industrial processes resulting in significant exposures to workers. These exposures continued into the late 1970s when overwhelming evidence of adverse health effects prompted strict regulation, engineering controls, and substitution for less hazardous substances. Thus unless accounted for, radiation risk estimates from occupational studies may be distorted by benzene exposures in the workplace. Nevertheless, few studies of leukemia radiogenicity have adjusted risk estimates for the effects of concomitant benzene exposures, and most results do not suggest strong confounding at typical occupational exposure levels [Ishimaru et al. 1971; Kubale et al. 2008; Kubale et al. 2005; Schubauer-Berigan et al. 2007a; Yiin et al. 2005].

2.7.2.1 Ionizing Radiation

The most studied risk factor for leukemia is ionizing radiation. Energy deposition in tissue from ionizing radiation results in targeted DNA damage either through direct ionization or indirectly through

interactions with newly-formed reactive oxygen species. Common lesions observed are single strand

breaks (SSB), double strand breaks (DSB), base lesions, and abasic sites [apurinic/apyrimidinic sites].

DNA damage results in increased levels of chromosomal rearrangements or mutation that, in rare

instances, may set in motion neoplastic transformation. Thus, the clastogenic nature of ionizing

radiation is believed necessary but not sufficient to induce cancer. In addition to the classic model of

targeted effects, recent research has shown that non-targeted effects may have an important role in

radiation carcinogenicity, especially at low doses [Barcellos-Hoff and Nguyen 2009; Morgan and Sowa

2005; Morgan and Sowa 2009]. Non-targeted effects include: a bystander effect, whereby a cellular

response to ionizing radiation is observed in non-irradiated cells; and radiation-induced genomic

instability, whereby the rate of genetic alterations increases in the progeny of irradiated cells.

The evidence of an association between ionizing radiation exposure and leukemia is

unequivocal. As early as 1944, significant excesses in leukemia were observed in epidemiologic studies

of physicians and radiologists [Henshaw and Hawkins 1944; March 1944]. These findings were

subsequently confirmed in the 1950s by comprehensive examinations of radiologists [Court-Brown and

Doll 1958; Lewis 1957], atomic-bomb survivors [Folley et al. 1952; Lange et al. 1954; Moloney 1955;

Moloney and Kastenbaum 1955; Moloney and Lange 1954], and radiation therapy recipients, such as

ankylosing spondylitis patients [Court-Brown and Doll 1957], and infants receiving radiation treatments

for thymic enlargement [Murray et al. 1959; Simpson and Hempelmann 1957; Simpson et al. 1955]. By

the 1960s, it was realized that this association persisted regardless of sex, age, race, amount of body

exposed (provided a sufficient amount of hematopoietic tissue is exposed), and whether the dose was

administered at one time or fractionated [Hempelmann 1960]. A clear radiation dose-response had

emerged at doses below a deterministic threshold but above 500 mSv [Court-Brown 1958; Hempelmann

1960]. The observed dose-response was characterized by short latency (<5 years), was more pronounced in acute leukemia subtypes compared to CML, and was not apparent in CLL cases.

Moreover, early genetic studies suggested that the rate of certain somatic mutations following low-LET exposure was a function of the amount (i.e. dose) and the intensity (dose rate) of ionizing radiation without an apparent threshold [Russell et al. 1958]. Based on these early findings, the cause-effect association was believed limited to ALL, AML, and CML subtypes and was characterized by a linear non- threshold dose-response with slope modified by radiation intensity (hereafter referred to as the LNT model). Although extensive research has continued since the 1950s, the LNT model (or more accurately a linear-quadratic form of the LNT model) prevails as the preferred hypothesis for radiation-induced leukemia [NRC 2005].

2.7.2.1.1 Low-LET, high dose rate studies

The majority of epidemiologic information on the leukemogenicity of ionizing radiation has stemmed from studies of radiation therapy patients and survivors of the atomic bomb blasts in 1945

[NRC 2005; UNSCEAR 2009]. In particular, the continued followup of the Japanese atomic-bomb survivors in the LSS forms the foundation for stochastic models used to derive current worker protection standards. The LSS has the advantages of large size ( ∼87,000), all ages at exposure, both sexes, long followup period (currently 1950-2000) and relatively precise dose estimates over a range of primary interest (0 to 4 Gy with 60% less than 0.1 Gy). The LSS also has a number of limitations, such as: uncertainty in risk transport from Japanese to other populations (e.g., rarity of CLL in Asian populations); all doses indirectly estimated (i.e., no direct measurement data); and all exposures were at high dose rates, thus risk extrapolation must account for differences in acute versus protracted exposures.

Richardson et al. [2009] examined mortality patterns by leukemia subtype in the LSS cohort

(n=86,611) followed from 1950 through 2000. The dose-response for leukemia was best fit to a linear-

quadratic model modified by age at exposure, time since exposure, and city of exposure, but not sex.

The preferred model for leukemia of all types followed the form:

v w w ͦ ƳR ́ͮǦu ́/ͮǦv ́/ ͮǦw ́/ ͮǦy ́ʚ/ͯͧͤʛ~Ʒ ̿͌͌ ʚ͘, ͗, ͙, ͨʛ Ɣ ͘ ƍ ͘ ʚ1 ƍ !͗ ʛ͙ Where d is estimated marrow dose (Gy), c is city and coded -1 for Hiroshima and 1 for Nagasaki, t is time

since exposure (years), and the quantity (e-30)/10 or 0, whichever is smallest, whereby e is age at ͙́ Ɣ exposure (years). The ERR is adjusted by time since exposure, t, using if ͧ ͧ ʚͨ Ǝ 30ʛͮ Ɣ ʚͨ Ǝ 30ʛ and 0, otherwise. Therefore, there was little evidence of time dependency in leukemia risk ʚͨ Ǝ 30 ʛ Ƙ 0 in persons aged greater than 30 years.

The adjusted ERR for leukemia following a 1 Gy dose to bone marrow at age ≥30 years was 2.38

(90% CI: 0.92, 4.47). AML risk was also not modified by sex and was best characterized by a quadratic dose-response function that peaked approximately 10 years after exposure [excess relative risk (ERR) at

1Gy=2.81; 90% CI: 1.63, 4.64; n=124; age=30+]. In contrast, ALL (ERR∙Gy -1 =3.70; 90% CI: 0.81, 12.99; n=19) and CML (ERR∙Gy -1 = 6.39; 90% CI: 3.00, 13.71; 58) risks were best described by a linear dose- response function that did not vary with sex, age at exposure, or time since exposure. These findings were similar to results in a previous study of leukemia incidence, whereby strong associations between

radiation exposure and leukemia incidence were observed for ALL, AML, and CML subtypes [Preston et

al. 1994]. AML incidence was also nonlinear in dose and was modified by age whereby youngest

survivors had the highest effect. As in Richardson et al. [2009], Preston et al. [1994] did not find

evidence of nonlinearity in CML or ALL risk in dose.

2.7.2.1.2 Protracted low-dose and low-dose rate studies

The health effects from radiation exposures in occupational and environmental settings have

been studied extensively and there are several reviews that summarize the epidemiologic evidence on

protracted low-dose leukemogenicity [NRC 2005; Schubauer-Berigan and Wenzl 2001; UNSCEAR 2009].

In principle, the direct estimation of risk in working populations is preferred to extrapolating risk from the LSS and radiation therapy patients. In fact, a major advantage of occupational studies is that most workers have been monitored for radiation exposure throughout their careers, thus individualized exposure records are generally available for periods under study. However, observational studies in environmental and occupational settings have generally lacked the precision necessary to adequately assess risk at low doses because: 1) studies have limited statistical power given ionizing radiation is weakly carcinogenic at low doses, with relative risks below 100 mGy of less than two-fold; 2) risk estimates are subject to distortion by the effects of coexposures or other factors; and 3) dose estimates have been subject to many sources of uncertainty over time, thus increasing the potential for bias.

Nevertheless, several studies have reported relatively precise estimates of increase leukemia risk in populations with average protracted cumulative exposures at or below 100 mGy.

The largest occupational study of leukemia risk and protracted exposure to low-LET, low-dose ionizing radiation was recently conducted by Schubauer-Berigan et al. [2007a; 2007b] who examined leukemia mortality in a cohort (n=94,517) of radiation monitored nuclear workers who were employed for at least 30 days in one of four DOE facilities (Hanford, LANL, ORNL, or SRS) or served as a civilian worker in a nuclear Naval Shipyard (PNS). Vital status followup varied by facility between 1990 and

1996. A nested case-control design was used, whereby each leukemia case ( n=206) was matched to four controls on attained age. This design enabled significant improvements to the exposure assessment, which considered internal (plutonium and tritium deposition) and external radiation exposure pathways from all relevant occupational sources of high- and low-linear energy transfer (LET) radiations [Anderson and Daniels 2006; Daniels et al. 2005; Daniels et al. 2006; Daniels and Schubauer-Berigan 2005].

Cofactors that were considered in analyses included sex, solvent exposure (benzene and carbon tetrachloride), birth cohort, hire year, facility, and smoking habits. Subtype analyses for AML, ALL, and

CML were not conducted; however, analyses considered both leukemia excluding CLL [Schubauer-

Berigan et al. 2007a] and CLL alone [Schubauer-Berigan et al. 2007b].

In adjusted analyses excluding 22 leukemias of ambiguous subtype, Schubauer-Berigan et al.

[2007a] reported an ERR∙100 mSv -1 for leukemia excluding CLL of 0.26 (95% CI: <-0.10, 1.3), which was

comparable to results from the linear region of the LSS for adult males age 20-60 years (ERR∙100 mSv -1=

0.15; 95% CI: -0.11, 0.53). Unlike the LSS, there was no evidence of a quadratic dose-response, i.e., a linear model provided the best fit to the data. When workers with a cumulative bone-marrow dose of at least 100 mSv were excluded, the ERR∙100 mSv -1 rose to 0.68 (95% CI: -0.28, 2.41), suggesting a higher

risk per unit dose among low-dose workers. Similarly, the dose-response for the CLL subtype was not

apparent in the full model (ERR∙100 mSv -1= -0.002; 95% CI: <0, 0.016); however, the risk estimate became positive (ERR∙100 mSv -1= 0.020; 95% CI: 0.004, 0.010) when restricting analysis to persons with

cumulative doses below 100 mSv [Schubauer-Berigan et al. 2007b]. Sensitivity analyses resulted in

preferred exposure lag periods of two- and 10-years for models of leukemia excluding CLL and CLL,

respectively. Separating doses into windows in years prior to the cutoff date resulted in positive ERR

estimates for exposures between 5 and 20 years in analyses of all workers. The highest risk of non-CLL

leukemia mortality per unit dose was observed in for exposures that occurred in the 5-10 year window

(ERR∙10 mSv -1= 0.32; 95% CI: <0, NC). Risk attenuation in dose above 100 mSv was consistent with

observations of strong risk modification by the highly correlated variables of birth cohort and hire year.

This modification disappeared after excluding workers with doses in excess of 100 mSv. Schubauer-

Berigan et al. [2007a] postulated that errors in exposure assessment and/or case ascertainment were

more likely to occur in the early employment years (i.e., pre-1950s) and that this differential may have led to biased results when including earlier person-year contributions.

Daniels and Schubauer-Berigan [2011] conducted a meta analysis that examined the relation between protracted low-dose ionizing radiation exposure and leukemia using information from recent

reviews by the National Academies and the United Nations and articles published prior to the end of

March, 2010. Selection criteria were used to identify 23 epidemiologic studies of occupationally and environmentally exposed populations that: 1) examined the association between protracted exposures to ionizing radiation and leukemia mortality or incidence excluding chronic lymphocytic subtype; 2) were a cohort or nested case-control design absent of any major bias; 3) reported quantitative estimates of exposure; and 4) conducted exposure-response analyses using relative or excess relative risk per unit exposure. These studies are listed in Table 2-2 and comprise the most recent information on leukemia risk in populations with chronic low-dose exposures.

Several combinations of studies were examined using random effects models to summarize between-study variance, assess publication bias, and obtain an aggregate estimate of the excess relative risk at 100 mGy. The preferred model was adjusted for publication bias and made full use of existing pooled analyses, but did not contain overlapping information. The aggregate estimate of the ERR at 100 mGy was 0.19 (95% CI: 0.07, 0.32) from combining information from 10 studies. Other combinations of studies provided similar results. Between-study variance was not evident ( P=0.99) in any model tested.

The results suggested that protracted exposures to low-LET penetrating radiation is associated with leukemia and that the dose-response was comparable to that observed in adults exposed in the LSS.

Three studies were updated since the publication of the meta-analysis. Krestinina et al. [2010] reexamined leukemia risk in Techa River villagers after extending followup through 2005. The risk of leukemias other than CLL was comparable to, although less than, earlier reports (ERR∙100 mGy -1=0.49;

95% CI: 0.16, 1.43). Laurent et al. [2010] examined mortality patterns in the French Electricity Company workers ( n=22,393) after extending followup through 2003. They observed a reduction in the leukemia risk excluding CLL (RR at 100 mGy= 0.52; 90% CI: 0.07, 1.78; 15 deaths) compared to the previous study by Rogel et al. [2005] (RR at 100 mGy= 1.68; 90% CI: 0.016, 7.22; 4 deaths) with followup through 1994.

The new followup increased the number of leukemia cases by three-fold; however, all new cases fell in

the low dose category. As recognized by Laurent et al. [2010], the net effect of increased followup was a substantial reduction in the observed risk, although confidence intervals overlapped. Boice et al. [2011] updated the analysis of mortality in Rocketdyne workers through 2008. The relative risk at 100 mSv was

1.06 (95% CI: 0.50, 2.23) for leukemia excluding CLL, which was lower than the estimate provided in the

earlier study (1.34; 95% CI: 0.73, 2.45), although confidence intervals overlap.

Studies reporting on leukemia risk but excluded from the meta analysis are summarized in Table

2-3. Most excluded studies lacked sufficient information on dose response but did provide additional

information on leukemia risks in radiation-exposed populations. Although most of these studies were

highly imprecise, they collectively suggested that leukemia risk is slightly elevated in exposed groups,

with typically less than a two-fold increase in risk observed in occupational studies with ten or more

cases.

Table 2-2. Studies selected for meta analysis [Daniels and Schubauer-Berigan 2011]

Study Average Related follow-up Study exposure studies Effect size ID Study Population period Design (mGy) (Study ID) Cases 1 (ERR at 100 mGy) Reference 1 Taiwan bldg. residents exposed to 1983-2005 cohort 47.7 none 6 0.19 [Hwang et al. 2008] Co 60 source incidence (90% CI: 0.01, 0.31) 2 Russian nuclear workers at Mayak 1948-1997 cohort 810 none 66 0.10 [Shilnikova et al. mortality (90% CI: 0.05, 0.21) 2003] 3 Techa River residents 1950-1999 cohort 500 none 49 0.65 [Krestinina et al. mortality (95% CI: 0.18, 2.40) 2005] 4 Chernobyl-Ukraine cleanup 1986-2000 case-control 76.4 8 32 0.27 [Romanenko et al. workers incidence (95% CI: <0, 1.35) 2008] 5 U.S nuclear workers from four varied case-control 30 mSv 7, 11, 22, 184 5a:- 0.26 [Schubauer-Berigan DOE facilities and a nuclear mortality 23 (95% CI: -0.10, 1.03) et al. 2007a] shipyard 2 5b: 0.10 (95% CI: <0, 0.65) 6 Rocketdyne workers 1948-1999 cohort 14 mSv none 18 0.34 [Boice et al. 2006] mortality (95% CI: -0.27, 1.45) 7 Pooled cohort of nuclear workers 1943-1997 4 pooled 19 mSv 5, 9, 13, 14, 7a: 7a: 0.19 [Cardis et al. 2007] from 15-countries 3 cohort 15, 16, 17, 196 (90:CI: <0, 0.71 mortality 19, 20 7b: 7b: 0.27 104 (90% CI: <0, 1.4) 8 Chernobyl cleanup workers from Russia: case-control 14.7 4 19 0.50 [Kesminiene et al. Russian Belarus, and Baltic 1993-1998 incidence (90% CI: -0.38, 5.70) 2008] countries Belarus: 1993-2000 Baltic: 1990-1998 9 UK nuclear workers within the 1955-2001 cohort 24.9 mSv 7 198 0.17 [Muirhead et al. National Registry for Radiation mortality (90% CI: 0.01, 0.43) 2009] Workers (NRRW) 10 Karunagappally, India, 1990-2005 cohort 161 none 20 0.37 [Nair et al. 2009] environmentally exposed cohort incidence (95% CI: <0, 33.68)

Study Average Related follow-up Study exposure studies Effect size ID Study Population period Design (mGy) (Study ID) Cases 1 (ERR at 100 mGy) Reference 11 US nuclear workers from the 1952-2002 cohort 43.7 mSv 5 62 0.77 [Richardson and Savannah River Site (SRS) mortality (90% CI: 0.14, 1.98) Wing 2007] 12 French Atomic Energy 1968-1994 cohort 16.9 mSv 7 20 3.10 [Telle-Lamberton et Commission and Cogema workers mortality (90% CI: 0.40, 11.44) al. 2007] 13 US nuclear workers at the Idaho 1949-1999 cohort 13.1 mSv 7 52 0.54 [Schubauer-Berigan National Laboratory (INL) mortality (95% CI: -0.11, 2.38) et al. 2005] 14 Canadian commercial nuclear 1957-1994 cohort 13.5 mSv 7, 16, 21 18 5.25 [Zablotska et al. power workers mortality (95% CI: 0.02, 29.10) 2004] 15 U.S. commercial nuclear power 1979-1997 cohort 25.7 mSv 7 26 0.57 [Howe et al. 2004] workers mortality (95% CI: -0.26, 3.04) 16 Workers in the Canadian National 1951-1987 cohort 6.3 mSv 7, 14, 21 23 0.04 [Ashmore et al. Dose Registry (CNDR) mortality (90% CI: -0.49, 0.57) 1998] 17 Korean nuclear workers 1984-2004 cohort 6.1 mSv 7 7 1.68 [Ahn et al. 2008] mortality (90% CI: -3.40, 14.90) 18 US workers at the Rocky Flats 1957-1983 cohort 41.0 mSv none 6 -0.72 [Gilbert et al. 1993] nuclear weapons facility (RFP) mortality (95% CI: <0, 4.20) 19 Japanese nuclear workers 1986-1997 cohort 12.0 mSv 7 60 0.00 [Iwasaki et al. 2003] mortality (90% CI: -1.00, 1.00) 20 French National Electric Company 1961-1994 cohort 5.6 mSv 5 7 5 0.68 [Rogel et al. 2005] workers mortality (90% CI: -0.84, 6.22) 21 Workers in the Canadian National 1969-1988 cohort 6.6 mSv 7, 14, 16 45 0.27 [Sont et al. 2001] Dose Registry (CNDR) incidence (90% CI: <0, 1.88) 22 Civilian workers at the 1952-1996 cohort 20.6 mSv 5, 23 34 1.09 [Yiin et al. 2005] Portsmouth Naval Shipyard mortality (95% CI: -0.09, 3.88) 23 Civilian workers at the 1952-1996 case control 23.2 mSv 5, 22 34 2.30 [Kubale et al. 2005] Portsmouth Naval Shipyard mortality (95% CI: 0.30, 8.90) 1Leukemia case without CLL subtype 25a are results from the full cohort. 5b results exclude SRS. 37a results are from the full cohort. 7b results exclude the U.S. cohorts. 4Follow-up period shown varied among countries. Period shown refers to all countries reporting. 5Median value

Table 2-3. Studies of leukemia and ionizing radiation exposure that did not meet inclusion criteria for meta-analysis.

Study Population Study Study Design Cases Effect size 1 Exclusion reason Ref. period Belgium nuclear 1969-1994 Cohort mortality 1 SMR=5.55 No risk per unit dose. [Engels et al. 2005] workers (n=7,229) (95% CI: 0.22, 16.50) Included in 15- country IARC study. Australian nuclear 1972-1996 Cohort incidence 8 SIR=1.00 No risk per unit dose. [Habib et al. 2006] workers (n=7,076) (95% CI: 0.50-1.99) Included in 15- country IARC study. Finnish physicians 1970-2001 Cohort incidence 2 Unexposed is referent. RR = 4.4 Leukemia endpoint [Jartti et al. 2006] (n=1,312 exposed; (95% CI: 0.5, 20) not reported. 15,821) Norwegian nurses 1953-2002 Cohort incidence 55 Unexposed is referent. Leukemia endpoint [Lie et al. 2008] (n=1,312 exposed; 5 <30 years since first exposure not reported. 15,821) RR=1.13 (95% CI: 0.44,2.94) 8 +30 years since first exposure RR=0.77 (95% CI: 0.35,1.69 Danish nurses 1980-2003 Cancer incidence 128 SIR=0.9 (95% CI: 0.8, 1.1) No risk per unit dose. [Kjaer and Hansen 2009] (n=92,140) Serbian medical 1992-2002 Cohort incidence 1 4% of cancers in females No risk per unit dose. [Milacic 2008] workers (n=1,560 exposed) Australian nuclear 1972-1998 Mortality in two 6 Monitored workers SMR=1.46 No risk per unit dose. [Habib et al. 2005] workers at Lucas inception cohorts: (95% CI: 0.66, 3.25) Heights Science and Cohort 1 ( n=4,717) Technology Centre Cohort 2 ( n=3,543) (LHSTC) US radiologic 1926-1998 Cohort incidence 41 Holding patients 50 or more times No risk per unit dose. [Linet et al. 2005] technologists (n=71,894) for x ray examination Included in 15- RR = 2.6 country IARC study. (95%: 1.3, 5.4)

Study Population Study Study Design Cases Effect size 1 Exclusion reason Ref. period US nuclear shipyard -1982 Cohort mortality NR Workers +50 mSv compared to No risk per unit dose. [Matanoski et al. 2008] workers (Sample: n=28,000 workers 5-10 mSv ≥5 mSv; 10,462 < 5 RR=2.41 (0.5-23.8) mSv; 33,353 non monitored) Nuclear workers at 1955-1995 Cohort mortality 4 Radiation workers: SMR=1.00, No risk per unit dose. [McGeoghegan and Binks the British Nuclear (n=2,628) P>0.05 Partially overlaps 2001] Fuels (BNFL) with more recent Chapelcross facility study. Nuclear workers at 1947-1992 Cohort mortality 9 Trend test (dose-response); no lag, No risk per unit dose. [Omar et al. 1999] the Sellafield plant of (n=14,319) P=0.006; 2-year lag, P=0.02). Included in 15- British Nuclear Fuels country IARC study. Chernobyl cleanup 1986-1998 Cohort incidence 7 SIR=1.53 No risk per unit dose. [Rahu et al. 2006] workers from (n=4,786 men) and (95% CI: 0.62, 3.17) Included in 15- Estonia Latvia ( n=5,546 country IARC study. men) US nuclear workers 1974-1997 Cohort incidence 9 SIR=0.63 No risk per unit dose. [Whorton et al. 2004] at the Lawrence (n=17,785) (95% CI: 0.29,1.19) Included in IARC Livermore National Chernobyl study. Laboratory (LLNL) Australian 1952 - Cohort mortality 47 SIR=1.61 No risk per unit dose. [Gun et al. 2008] participants in the 2001 and incidence (1.18 to 2.14) British nuclear tests (mortality) (n=10,785) in Australia 1982-2001 40 SMR=1.25 (incidence (95% CI: 0.89, 1.70) ) Dental workers in 1951-1987 Cohort ( n=42,175) 15 SMR=1.26 No risk per unit dose. [Zielinski et al. 2005] the National Dose mortality (90%CI: 0.80, 1.89) Registry cohort

Study Population Study Study Design Cases Effect size 1 Exclusion reason Ref. period UK atomic test 1952-1998 Cohort ( n=21,357 49 incidence RR=1.41 No risk per unit dose. [Muirhead et al. 2003] participants test subjects; n=22, (90%CI: 0.96, 2.09) Included in 15- 333 men who did country study. not participate in 40 mortality RR=1.83 the tests) mortality (90%CI: 1.15, 2.93) and incidence Northern Germany 1986-1998 Population-based 249 4% of men and 8% of women were No risk per unit dose. [Hoffmann et al. 2008] population study 1,430 cases, occupationally exposed to ionizing 3,041 controls radiation for ≥1 year over their lifetime US residents of 1950-2000 Population-based 492 RR=0.95 No risk per unit dose. [Boice et al. 2006] counties near the study (95% CI: 0.85, 1.06) Hanford nuclear facility U.S. residents living 1984-2004 Population-based 59 SIR= 0.96 No exposure [Boice et al. 2009a] near the former study (95% CI: 0.73-1.24) information. No risk Apollo and Parks per unit dose. nuclear facilities in western Pennsylvania U.S. residents in 1950-2004 Population-based 330 mortality RR= 1.00 No exposure [Boice et al. 2009b] Westmoreland and study (95% CI: 0.87,1.14) information. No risk Armstrong counties 1990-2004 per unit dose. in Pennsylvania (incidence 629 incidence RR=1.00 compared to ) (95% CI: 0.90, 1.10) residents in six demographically similar counties in western Pennsylvania Chinese medical x- 1950-1995 Cohort incidence 44 RR=2.17, P<0.05 No exposure [Wang et al. 2002] ray technologists (n=27,011exposed, information. No risk 25,782 unexposed) per unit dose.

Study Population Study Study Design Cases Effect size 1 Exclusion reason Ref. period Lithuanian medical 1948-2004 Cohort ( n=2,250) 3 men SIR= 3.3 (95% CI: 0.68, 9.63) No risk per unit dose. [Samerdokiene et al. radiation workers incidence is workers women SIR=2.67 (95% CI: 0.92, 2009] employed for at 4.2) least one year and alive in 1978 U.S. Submariners 1969-1982 Cohort ( n=76,160 12 SMR=0.91 (95% CI:0.47, 1.60) No risk per unit dose. [Charpentier et al. 1993] men) mortality French workers in 1968-1994 Cohort CEA 1 SMR=1.17 No risk per unit dose. [Guseva Canu et al. 2008] the Atomic Energy (n=3,509) mortality (95% CI: 0.01,1.49) Commission (CEA) and the National 1980-1993 Cohort INSERM 2 SMR=1.12 (95% CI: 0.19, 3.52) Institute of Health (n=4,966) mortality and Medical Research (INSERM). Danish military 1992-2002 Cohort incidence 4 SIR=1.4 No risk per unit dose. [Storm et al. 2006] personnel deployed (n=14,012) (95% CI:0.4, 3.5) CEA workers included to the Balkans. in 15-country IARC Potentially exposed study. to depleted uranium between 1992-2001 U.S. Chemical 1943-1998 Cohort mortality 13 OR (multiplicative model)=1.02 Not relevant [Kubale et al. 2008] laboratory workers (n=6,157) (95% CI: 0.99, 1.06) exposure. employed at Department of Energy facilities. Canadian medical 1951-1987 Cohort mortality 15 SMR= 0.60 (95% CI: 0.37, 0.92) Not relevant [Zielinski et al. 2009] workers (n=67,562) exposure.

1969-1987 Cancer Incidence 19 SIR = 0.58 (95% CI: 0.38, 0.85) Workers of the 1954-1992 Cohort mortality 4 SMR=0.70 (95% CI: 0.19, 1.80) No risk per unit dose. [Rodriguez Artalejo et al. former (n=5,657) 1997] Spanish Nuclear Energy Board

Study Population Study Study Design Cases Effect size 1 Exclusion reason Ref. period U.S. construction 1998-2004 Cohort mortality 5 SMR=0.66 (95% CI: 0.22, 1.55) Leukemia endpoint [Dement et al. 2009] workers employed at (n=8,976) not reported. nuclear sites Workers included in 15-country IARC study. French nuclear 1967-2002 Cohort mortality 1 SMR=0.26 (95% CI: 0.04, 1.84) No risk per unit dose. [Guerin et al. 2009] contract workers (n=6,962) Workers from two nuclear fuel cycle facilities were workers included in 15-country IARC study. French nuclear 1977-2004 Cohort mortality 9 SMR=0.70 (90%CI: 0.37, 1.23) No risk per unit dose. [Metz-Flamant et al. workers (n=9,285) 2009] Korean nuclear 1992-2005 Cancer incidence 3 SIR=1.34 (95% CI: 0.27, 3.92) No risk per unit dose. [Jeong et al. 2010] industry workers (n=16,236; 8,429 (radiation workers) radwkrs, 7,807 6 RR=0.68 (95% CI: 0.12-3.79) nonradwkrs) Finnish reindeer 1971-2005 Cancer incidence 61 SIR=0.83 (95% CI: 0.64, 1.07) No risk per unit dose. [Kurttio et al. 2010] herders exposed to (n=34,563) fallout 1SIR=Standardized Incidence Ratio, SMR=Standardized Mortality Ratio, RR=Rate Ratio, NR=not reported; NA=not applicable

2.7.2.1.3 High-LET studies

Strong elevations (5-20 fold) in leukemia excluding CLL were observed in patients injected with

232 Thorotrast (i.e., Th in a ∼25% ThO 2 colloid suspension) contrast medium that was used mostly in the

period from the 1930s to mid-1950s [dos Santos Silva et al. 2003; Martling et al. 1999; Mori et al. 1999;

Nyberg et al. 2002; Travis et al. 2003; van Kaick et al. 1999]. Similarly, Wick et al. [2009] observed a

three-fold leukemia excess (primarily myeloid) in ankylosing spondylitis patients ( n=1,471) treated with

radium injections ( 224 Ra), compared to a control group of patients without radiation therapies. These

studies and complementary data from animal models are sufficient to conclude that alpha-emitting

radiation is leukemogenic. Nevertheless, there is little evidence of an association between leukemia risk

and radiation from occupational exposures to internal alpha-emitters. A leukemia excess was not

observed in the female radium dial painters; however, slight elevations were observed in male radium

dial painters, although there were few cases [Stebbings 1998]. Leukemia excesses attributable to

internal doses have not been observed in several plutonium worker studies, although numbers were

small and internal dosimetry methods were crude [Omar et al. 1999; Shilnikova et al. 2003; Wiggs et al.

1994]. Studies of leukemia in uranium miners have been inconsistent. Darby et al. [1995] observed

increased leukemia mortality (SMR=1.93; 95% CI: 1.19, 2.95; 22 cases) in a pooled study of underground

miners ( n=64,209) although this risk was limited to the period less than 10 years since first employed

and a dose-response trend was not evident ( P=0.91). A recent study of Czech uranium miners ( n=23,043)

suggested that leukemia, in particular CLL, was associated with radon exposures [Rericha et al. 2006].

Rericha et al. [2006] compared high radon exposed group [110 working level months (WLM); 80th

percentile] to the low radon exposed group (3 WLM; 20th percentile) and found a relative risk of 1.75

(95% CI: 1.10, 2.78) for all leukemia incidence and 1.98 (95% CI: 1.10, 3.59) for CLL incidence. Other

studies of uranium miners have not provided similar results [Lane et al. 2010; Laurier et al. 2004;

Mohner et al. 2006; Schubauer-Berigan et al. 2009].

There is a notable absence of bone marrow neoplasms in human and animal studies involving alpha-emitting chemical compounds other than Thorotrast. This disparity may result from both biokinetic and radiolytic differences. Thorotrast (a colloidal suspension) is phagocytized and uniformly distributed throughout organs of high phagocytotic function (i.e., liver and bone marrow). Once deposited, it is rarely excreted. The isotope 232 Th is long-lived and irradiates surrounding tissue directly by alpha decay and indirectly via its beta-gamma emitting progeny. Most other alpha-emitting compounds encountered in typical nuclear work have decay chains that are less complex and less energetic on an activity basis, are retained less in the body, and tend to deposit on bone surfaces causing only partial irradiation of hematopoietic tissue. It has been suggested that partial irradiation of the target organ is less likely to induce cancer [Breckon and Cox 1990; Spiers and Vaughan 1989]. It has also been proposed that cancer induction results from the indirect effect of cellular responses to chronic tissue damage caused by alpha radiation rather than the direct bombardment by alpha-particles.

Autopsy information from decedents who received Thorotrast injections shows that tumors rarely formed at the center of thorium deposition. Instead, the deposition sites were marked by dense fibrosis

and cancer sites were located at a distance from the fibrosis that was beyond the range of the alpha-

particle [Yamasaki et al. 2004]. Because Thorotrast is highly energetic and is retained in the body

indefinitely, the rate of damage to surrounding tissue is continuous.

The fraction of total dose from high-LET exposures is usually small compared to low-LET

exposures in studies involving mixed radiations [Boice et al. 2006; Howe et al. 2004; Iwasaki et al. 2003;

Muirhead et al. 2009; Richardson and Wing 2007; Rogel et al. 2005; Schubauer-Berigan et al. 2007a;

Zablotska et al. 2004]. Therefore, finding effects on leukemia risk that are attributable to high-LET

exposure is unlikely in epidemiologic studies unless a strong effect exists. To date, studies examining

these effects have been inconclusive, thus suggesting that high-LET exposures in the workplace have

little (if any) effect on the leukemia risk [Boice et al. 2006; Schubauer-Berigan et al. 2007a; Shilnikova et al. 2003].

2.7.2.1.4 Other incorporated radionuclides

Information on leukemia risk from other incorporated radionuclides is sparse. Little et al. [2007] conducted a systematic literature review to compare the cancer risk in persons primarily exposed via incorporated radionuclides to the risks derived from the LSS cohort. They found three studies to be informative on leukemia risk from internal high-LET radiation [dos Santos Silva et al. 2003; Travis et al.

2003; van Kaick et al. 1999] and four studies that were informative on leukemia risk from low-LET internal radiation [Davis et al. 2006; Krestinina et al. 2005; Ostroumova et al. 2006; Ron et al. 1998]. All three high-LET studies were of Thorotrast patients as previously discussed. The low-LET studies were of

populations exposed to widespread fission-product contamination from the former Soviet Mayak facility

[Krestinina et al. 2005; Ostroumova et al. 2006] or Chernobyl reactor accident cleanup activities [Davis

et al. 2006] and from high-dose 131 I treatment of adult hyperthyroid patients [Ron et al. 1998]. Studies examining the health effects from incorporated fission products (e.g., 137 Cs, and 90 Sr) from chronic

exposure to low-level contamination were considered to be the most relevant to exposure potentials in

the workplace and thus are described in more detail below.

Recent studies of persons chronically exposed at low-dose rates from incorporated radioactive

strontium have suggested a linear dose-response relation. The most informative studies involved a

population ( ∼30,000) who lived on the banks of the Techa River in the Southern Urals region of the

Russian Federation between 1950 and 1960 [Krestinina et al. 2010; Krestinina et al. 2005; Ostroumova

et al. 2006]. These people were exposed to fission product contamination in the soil and water from

releases at the Mayak nuclear weapons facility, which was located upstream of the river villages.

Although other fission product and transuranic radionuclides (e.g., 137 Cs, 239 Pu) were present, the primary source of ionizing radiation exposure to Techa River residents is 90 Sr contamination in food and

drinking water. Upon uptake of soluble strontium-bearing compounds, biokinetic transport is similar to that of calcium, which is readily deposited in bone. 90 Sr and its immediate progeny, 90 Y, are beta-

emitters with beta-particle mean energies of 0.2 MeV and 1.0 MeV, respectively, and ranges in soft

tissue of about 2 mm and 10 mm, respectively. Given its long physical half-life ( ∼29 y), and incorporation in bone, 90 Sr retention time is relatively long (15-25 y; Tolstykh et al. [1997]), resulting in chronic

exposure to hematopoietic tissue in those exposed. The estimated mean and maximum cumulative

absorbed doses to bone marrow in Techa residents were 0.3 Gy and 2 Gy, respectively [Krestinina et al.

2010]. Most ( ∼80%) of the dose was accrued during the first decade of exposure. The mean dose rate

decreased from about 40.0 mGy∙y -1 between 1950 and 1951 to 8.0 mGy∙y -1 beginning in 1960.

In the most recent follow-up of the Techa River cohort (1953-2005), Krestinina et al. [2010] reported a significant linear dose-response relation for leukemia incidence excluding CLL (ERR∙100 mGy-1

= 0.49; 95% CI: 0.016, 1.43; n=70 cases). No evidence of a dose-response was found for CLL ( n=23 cases).

Chronic leukemias comprised more than half of the cases ( n=48). There were only three cases in persons

aged 10-20 years; no cases were reported in persons aged less than 10 years. The incidence results were

comparable to earlier studies of leukemia mortality in Techa River residents [Krestinina et al. 2005],

whereby the ERR∙100 mGy -1 for leukemia excluding CLL was 0.65 (95% CI: 0.018, 2.4; n=61 deaths).

Davis et al. [2006] conducted a population-based study of childhood leukemia in persons living

in Ukraine, Belarus, and Russia following the Chernobyl reactor accident in 1986. Study participants

were ≤ 6 years old and were exposed in utero or after birth to mixed fission-products, resulting in

median cumulative doses to bone marrow that were less than 10 mGy in all regions examined. The

source of prolonged radiation exposure is primarily radioisotopes of cesium (e.g., 134 Cs, 137 Cs) ingested by consumption of contaminated food and drinking water. It appears that internal exposure from the ingestion of strontium radionuclides ( 89 Sr, 90 Sr) and transuranic isotopes ( 238 Pu, 240 Pu, 241 Am) is negligible

because significant ground deposition of these radionuclides was not evident [Likhtarev et al. 1996]. The

study considered 421 confirmed leukemia cases, which included 311 (74%) ALL cases, 86 (21%) AML cases, and 24 (5.0%) cases of undefined subtype. A significant increase in leukemia risk with increasing

radiation dose was observed in the combined population (ERR∙100 mGy -1 =3.24; 95% CI: 0.80, 8.4;

n=421), although the association was driven primarily by data from Ukraine (ERR∙100 mGy -1 =7.88; 95%

CI: 2.21, 21.3; n=268). A positive dose-response was apparent, but not statically significant, in Belarus

children (ERR∙100 mGy -1 =0.41; 95% CI: <0, 3.7; n=114) and was not evident in Russian children (n=39

cases). The authors concluded that the heterogeneity in leukemia risk across regions may suggest

evidence of a strong bias in Ukraine results potentially from a tendency to choose controls from less

contaminated areas compared to the cases. Thus, the authors did not consider their results to be

compelling evidence of increased risk of childhood leukemia from exposures to Chernobyl fission

product contamination. Nevertheless, one cannot dismiss the possibility of increased leukemia risks in

the Russian and Belarus children that were not detected because of low statistical power. Noshchenko

et al. [2010] also studied leukemia risk in Ukrainian children of ages between 0 and 5 years in a case-

control study of 492 randomly matched controls and 246 cases diagnosed between 1987 and 1997.

Controls were matched by administrative region (oblast) of permanent residence at the time of the

Chernobyl accident. Although the risk estimate was somewhat diminished from previous results,

leukemia risk was significantly increased (OR=2.4; 95% CI: 1.4, 4.0) among those with doses higher than

10 mGy ( P <0.01).

2.7.2.1.5 Tritium

Tritium ( 3H) is a radioisotope of hydrogen that undergoes radioactive decay to stable helium by emitting a low-energy beta particle (5.7 keV average energy; 18.6 keV maximum) with a physical half-life of approximately 12.4 years and a maximum track length of approximately 6.0 µm in water. Tritium is produced naturally by interaction with cosmic radiation; however, it is principally an occupational exposure hazard via large scale production for nuclear weapons and as a radioactive tracer used in

medicine, research, and industry. Tritium exposures have also occurred, albeit to a lesser extent, through incidental releases as a by-product of nuclear reactor operations. The most likely uptake results from tritiated water or tritium oxide (HTO), which chemically behaves the same as ordinary water.

Common routes of intake include respiration of tritiated water vapor, ingestion via contaminated food and drinking water, and skin absorption. Once introduced into the body, tritium is quickly (within one to two hours) distributed uniformly among soft tissues. Simple single-compartment biokinetic models estimate that 97% of tritium (HTO) is eliminated from the body as water with a biological half-life of approximately 10 days and the remainder is metabolized as organically bound tritium (OBT) that is eliminated with a half-life of 40 days [Okada and Momoshima 1993]. Recent multi-compartment models suggest that the red-bone marrow (with relatively high metabolic rate) has a biokinetic half-life for OBT of about 8 days [Galeriu and Melintescu 2010].

Tritium toxicity and leukemogenicity have been well established by in vitro studies and in several animal models [Balonov et al. 1993; Straume and Carsten 1993]. Nevertheless, there is continued debate regarding the selection of dose coefficients for estimating cancer risks from tritium intake given uncertainties in metabolic models and effectiveness of tritium beta relative to other low-LET radiations

[Bridges 2008; Cox et al. 2008; Galeriu and Melintescu 2010; Goodhead 2009; Little and Lambert 2008;

Paquet and Metivier 2009]. Moreover, the epidemiologic information regarding tritium-induced health effects in humans is sparse and generally uninformative on leukemia risk [Little and Wakeford 2008].

Little and Wakeford [2008] recently conducted a systematic review of epidemiological studies involving occupational and environmental tritium exposures and found that potentially informative studies often lacked tritium-specific dose data, or suffer from low statistical power because of low doses and small numbers of cases. A few occupational studies have used tritium-specific individual dose estimates in combination with dose from other radiation sources; however, information was insufficient to examine

health effects from tritium separately [Beral et al. 1988; Cragle et al. 1999; Richardson and Wing 2007;

Schubauer-Berigan et al. 2007a; Zablotska et al. 2004].

2.7.2.1.6 Studies of radiation and CLL

It has been customary to restrict radio-epidemiologic analyses to all leukemia subtypes

excluding CLL based on a long standing consensus that CLL is not radiogenic. This conclusion was drawn

largely from observations of risk in the LSS, which is considered by most as the gold standard, and sparse

information from observational studies of medically and occupationally exposed cohorts [Silver et al.

2007]. Nevertheless, some researchers have recently questioned the exclusion of CLL from the list of

potentially radiogenic cancers [Linet et al. 2007; Richardson et al. 2005; Schubauer-Berigan et al. 2007b;

Silver et al. 2007]. Their concerns are based largely on limitations in previous observational studies to observe an effect, if examined at all, and the uncertainty in risk transport given that CLL is extremely

rare in Asian populations relative to Western populations. As a consequence, recent observational

studies of nuclear workers have attempted to examine CLL radiogenicity [Romanenko et al. 2008;

Schubauer-Berigan et al. 2007b; Vrijheid et al. 2008]. However, most of these studies have failed to

elucidate an association between ionizing radiation exposure and CLL. A noteworthy exception is the

study of by Romanenko et al. (2008), who reported elevated CLL risk in Chernobyl cleanup

workers([ERR∙Gy -1=2.73; 95% CI: <0, 13.5), although these findings were based on few CLL cases ( n=39).

2.7.2.2 Benzene

Benzene (sometimes referred to as “benzol”) is a natural constituent of crude oil and is the

parent hydrocarbon of the aromatic group. Benzene is among the top 20 most produced chemicals in

the U. S. with nearly 2,500 industrial facilities producing or processing the chemical annually [ATSDR

2007]. It has been used extensively as a solvent, in the synthesis of numerous chemicals (e.g., styrene),

and as a gasoline additive. Due to widespread production and environmental release, benzene is

ubiquitous in air, water, and food supplies. The dominant exposure pathway is inhalation, although exposure routes by ingestion and skin absorption are likely. The primary sources of environmental exposures are tobacco smoke and inhalation of contaminated air primarily in heavy vehicle traffic areas and refueling stations. Environmental benzene exposures are typically below 15 ppb (50 μg∙m -3)

[Capleton and Levy 2005].

In its 1990 National Occupational Exposure Survey, NIOSH estimated that nearly 300,000

workers were exposed to benzene from 1981 to 1983 [NIOSH 1990]. Approximately half of these

workers were employed in general medical and surgical hospitals. Occupational benzene exposure over

time can be characterized by historical changes in control levels. The first U.S. exposure control level

appeared as a maximum allowable concentration of 100 ppm as recommended by the American

Conference of Government Hygienists (ACGIH) in the 1940s [Capleton and Levy 2005]. As health effects

became more apparent, control levels were decreased. The first regulatory limit was 10 ppm set in 1971 by the Occupational Safety and Health Administration (OSHA). The current OSHA Permissible Exposure

Level (PEL) for benzene is 1.0 ppm for an eight-hour time-weighted-average (TWA) set in 1987 (Figure 2-

1). Historically, benzene exposure in U.S. nuclear industry workers occurs as the results of its widespread use as a powerful organic solvent. Bulk use of benzene typically involved closed processes, thus exposures were kept minimal. In addition, changes in work practices beginning in the 1950s, such as switching to non-benzene containing substitutes, is likely to have reduced benzene exposures in nuclear facilities. Nevertheless, benzene continued as a contaminant in petroleum distillates (2-5%), such as gasoline, and in some solvents (11-15%) that remain in general use even today. It is estimated that current long-term average occupational benzene exposures in the U. S. are at or below 1.0 ppm ( ∼3.2 mg∙m -3) or about 50 ppm-years in a working lifetime [Capleton and Levy 2005].

120

100

40 Exposure limit (ppm) 20

0 1940 1950 1960 1970 1980 1990 2000 Year

Figure 2-1. U. S. benzene exposure control levels for the period 1940-2005 [Capleton and Levy 2005; Infante 1978].

Benzene is arguably the oldest known and most studied chemical carcinogen. Concerns about benzene leukemogenicity date back to the late 1920s [Linet et al. 2006] and were widespread in the medical community by the 1960s based on evidence from a large number of case reports [Degowin

1963; Mallory et al. 1939; Vigliani and Forni 1976; Vigliani and Saita 1964]. By the 1970s, observational studies had emerged that suggested increased leukemia in benzene workers [Aksoy et al. 1974; Infante

1978; Infante et al. 1977; Thorpe 1974]. By the early 1980s, there was sufficient evidence for IARC (and subsequently the U.S.) to conclude that benzene is leukemogenic in humans. In its recent review of human carcinogens (IARC Monograph Volume 100F, in prep.), IARC states that there is sufficient evidence that benzene causes AML in humans and limited evidence of a causal association between benzene and lymphocytic leukemias (ALL and CLL). Most of the information stems from early epidemiologic studies of benzene-exposed occupations involving relatively high daily exposures (>10 ppm), such as painters and printers [Lindquist et al. 1987; Paganinihill et al. 1980], chemical workers [Ott et al. 1978; Thorpe 1974; Yin et al. 1987] rubber workers [Arp et al. 1983; Checkoway et al. 1984; Delzell and Monson 1981; Rinsky et al. 1981a], shoe manufacturing [Aksoy et al. 1974], and petroleum industry

workers [Austin and Schnatter 1983]. Overall, these studies suggested that the risk of benzene-induced leukemia in some occupational settings was increased by 1.9- to 10-fold and was mostly attributable to elevations in AML. Later studies suggested that leukemia risks were observed at much lower benzene doses (> 2 ppm-years) and were absent a threshold [Glass et al. 2003; Guenel et al. 2002; Hunting et al.

1995]. The temporal effects in leukemia risk from benzene exposure appear similar to that observed for ionizing radiation, whereby risks are higher within the first 10 years following exposure and there is little evidence of risk 20 years after exposure [Finkelstein 2000; Triebig 2010].

The early recognition of benzene leukemogenicity has resulted in a considerable reduction in its industrial use. Nevertheless, benzene exposures continue to occur in many industrial and environmental settings; thus, research examining the health effects of these exposures is ongoing. For example,

Khalade et al. [2010] recently conducted a meta-analysis of adult leukemia risk and benzene exposure by combining information in 15 occupational studies published prior to 2009. A random-effects model yielded a positive and significant estimate of effect size (ES), adjusted for study heterogeneity (ES=1.72;

95% CI: 1.37, 2.17). Aggregate estimates were also determined by low (<40 ppm-years), medium (40-

<100 ppm-years) and high (100+ ppm-years) exposure categories. Stratifying by exposure eliminated study heterogeneity and provided summary effect estimates for low (ES=1.64; 95% CI: 1.13, 2.39), medium (ES=1.90; 95% CI: 1.26, 2.89), and high exposures (ES=2.62; 95% CI: 1.57, 4.39) that indicated a clear dose-response pattern (trend test P=0.015). Analyses were also conducted by leukemia subtype.

The aggregate estimate for AML was 1.38 (95% CI: 1.15, 1.64) with evidence of a dose-response (low:

1.94; 95% CI: 0.95, 3.95; medium: 2.32; 95% CI: 0.90, 5.94; high: 3.20; 95% CI: 1.09, 9.45). The aggregate

estimate for CLL was 1.31 (95% CI: 1.09, 1.57), also with evidence of a dose-response (low: 1.83; 95% CI:

0.75, 4.48; medium: 1.67; 95% CI: 0.86, 3.24; high: 3.50; 95% CI: 0.90, 13.2), although estimates were

imprecise. There was no indication of an association between benzene and CML. The information on ALL

was insufficient to conduct analyses.

To explore the dose-response characteristics between benzene and leukemia, Vlaanderen et al.

[2010] fit meta-regression models to aggregated risk estimates extracted from observational studies

(n=9) of benzene-exposed workers. Although previous models assumed leukemia was linear in benzene exposure, the best fit was achieved to a curve that was supralinear below exposures of 100 ppm-years.

Their estimate of leukemia relative risk at 10 ppm-years was 1.14 (95% CI: 1.04, 1.26).

2.7.2.2.1 Low-dose and low-dose rate benzene exposures

The evidence of benzene-induced leukemia in humans stems from exposures in occupations involving relatively large quantities of benzene (e.g., rubber manufacturing and petroleum industries).

However, participants in the current study were likely to receive comparably smaller exposures as a result of limited quantities of benzene in facilities under study. Large-scale applications of benzene were not conducted in any study facility. Instead, benzene source terms typically involved laboratory work or incidental exposures to contaminated materials. For example, solvents containing trace amounts of benzene, such as those associated with dispensing gasoline, automotive repair, and painting were available in all study facilities. There is sufficient evidence of human carcinogenicity in exposures related to the paint industry; however, there is little evidence of increased leukemia risk in painters [IARC 2010].

Similarly, results from epidemiologic studies examining the relation between low-level benzene exposure and hematologic malignancies in occupations involving exposures to gasoline vapors provide little evidence of elevated leukemia risk [Hansen 1989; Hunting et al. 1995; Lagorio et al. 1994; Lynge et al. 1997]. Even so, it should be recognized that most of these studies were imprecise and highly susceptible to bias given their small size and limited ability to control for confounding [Hotz and

Lauwerys 1997]. Following an exhaustive review of studies examining hematopoietic and lymphatic cancer in automotive services, Hotz and Lauwerys [1997] concluded that benzene in gasoline cannot be excluded as a leukemia risk factor in vehicle mechanics, although the association was not evident by epidemiologic means.

Laboratory researchers and technicians are potentially exposed to a wide variety of cytotoxic agents, radioactive materials, and organic solvents. Benzene was among the solvents found in most study facility laboratories, albeit its use and storage has been mostly limited to small (<500 ml) quantities. Slight elevations in leukemia risk in laboratory workers have been reported in previous studies. In most cases, study results were imprecise, and comparisons lacked detailed exposure information [Brown et al. 1996; Cordier et al. 1995; Kubale et al. 2008; Rachet et al. 2000]. For example,

Brown et al. [1996] reported elevated leukemia risk in a cohort of researchers ( n=12,703) employed in biological research institutes in the UK and followed through 1994 (SMR=1.21; 95% CI: 0.49, 2.49; 7 deaths). No specific exposures were addressed in this study.

2.7.2.2.2 Benzene and radiation

Studies of the combined effects of radiation and benzene exposure on leukemia risk are sparse.

Most studies lack detailed information on exposures other than ionizing radiation. Nevertheless, a few studies have attempted to account for chemical leukemogens as potential effect modifiers or confounders in models of leukemia risk and ionizing radiation. Ishimaru et al. [1971] examined leukemia in atomic bomb survivors with potential coexposures to benzene and medical X-rays and observed a 2.5- fold increase in leukemia risk after adjusting for the atomic bomb exposures. However, information was insufficient to separate risks further. Yiin et al. [2005] examined leukemia mortality patterns in civilian shipyard workers ( n=13,468) exposed to solvents (primarily benzene and carbon tetrachloride) and/or ionizing radiation. There was no evidence of confounding by solvent exposure (ever/never). In a subsequent case-control study, Kubale et al. [2005] adjusted for solvent exposure (as a continuous variable), which was estimated using employment duration in certain job/shop combinations. They did not find evidence of effect modification or confounding by solvent exposure; however, it was noted that solvent exposures in most workers occurred many years prior to their radiation exposure. Schubauer-

Berigan et al. [2007a] adjusted for solvent exposure (benzene and carbon tetrachloride) in a case-

control study of leukemia mortality in a pooled study of nuclear workers. This study involved simultaneous radiation and benzene exposures at most facilities and showed evidence of potential confounding by benzene exposure, although the effect was weak. Kubale et al. [2008] examined mortality patterns in a cohort of laboratory workers ( n=6,157) ever employed at the Savannah River Site

(SRS) or the Oak Ridge sites (X-10, Y-12, and K-25 facilities) from 1943 to 1998. They found evidence of potential confounding of chemical risk estimates (primarily solvents) by ionizing radiation exposure, although estimates were imprecise. Actual benzene exposures were not assessed by Kubale et al.

[2008], although benzene was listed among the “typical chemicals” used by these workers.

Nakayama et al. [2009] used a novel approach to estimate benzene lifetime risks from a physiologically-based pharmacokinetic (PBPK) model that was derived, in part, from radiation equivalent effects. The model used information on the dose-response for chromosome aberrations from low-LET ionizing radiation and primary benzene metabolites from laboratory studies and the BEIR VII leukemia risk models. Based on this information, Nakayama et al. [2009] predicted that the leukemia risk from continuous benzene inhalation in males of 1.0 ppm over a lifetime (i.e., approx. 50 ppm-years) was approximately equivalent to the risk from an absorbed dose to bone marrow of 84 mGy (90%CI: 61.6,

95.2), using a radiation risk coefficient for leukemia of 7.0 x 10 -3Gy -1 [NRC 2005].

2.7.2.3 Other Chemical exposures

2.7.2.3.1 Formaldehyde

Formaldehyde (CH 2O) is a common chemical used in resin production, textile industry, health-

care, laboratories, and by morticians and embalmers. More than 11 billion pounds of formaldehyde is

produced in the U.S. each year, making it the 25 th in overall chemical production. It is estimated that 2.1 million U.S. workers are exposed to formaldehyde each year [Zhang et al. 2009]. Formaldehyde is a known human carcinogen (IARC Group 1) based primarily on the epidemiologic evidence of elevations in

nasopharyngeal cancer following exposure. Nevertheless, the evidence of formaldehyde leukemogenicity exists and is increasing. Collins et al. [2004] found slight elevations in leukemia in meta- analyses of embalmers (RR=1.6; 95% CI: 1.2, 6.0), and pathologists/anatomists (RR=1.4; 95% CI: 1.0, 1.9).

Zhang et al. [2009] conducted a meta-analysis of occupational studies (n=15) that indicated elevated leukemia risk in exposed workers (RR=1.54; 95% CI: 1.18-2.00). The highest risks were observed in studies ( n=6) of myeloid leukemias (RR=1.90; 95% CI: 1.31.-2.76). Widespread formaldehyde use by study participants was not likely, although there was some evidence of its use in small quantities in study facility laboratories [Mihlan 1997].

2.7.2.3.2 Other aromatics and organic solvents

Exposures to some substitutes for benzene, such as xylene and toluene, and other organic solvents are suspected to cause AML, although epidemiologic data are sparse. Studies of petrochemical workers exposed to benzene and other aromatic hydrocarbons have reported conflicting results [Divine et al. 1999; Huebner et al. 1997; Sathiakumar et al. 1995] and none controlled for benzene coexposures.

Elevated AML risk from exposure to solvents, but not to benzene, was reported in one case-control

study. Albin et al. [2000] found OR= 1.2 (95% CI: 0.69, 2.0) for "low" exposed group and OR=2.7 (95% CI:

1.0, 7.3) for ''moderate-high" exposed group compared to the unexposed group. However, recent large

studies have not confirmed these results [Costantini et al. 2008].

Trichloroethylene and tetrachloroethylene (perchloroethylene) solvents were used for metal

cleaning and degreasing activities in many of the study facilities over most operational years [Mihlan

1997]. Both of these halogenated hydrocarbon solvents are probably carcinogenic in humans (IARC

group 2A), based on limited evidence of increased risk of cancers of the liver and biliary tract from

trichloroethylene exposures, esophageal and cervical cancers from perchloroethylene exposures, and

non-Hodgkin’s lymphoma (NHL) from exposures to either solvent [IARC 1995]. The results from cohort

studies of leukemia risks in workers exposed to these solvents were inconsistent. One study reported

slightly elevated leukemia incidence [standardized incidence ratio (SIR)=1.1; 95% CI: 0.35-2.5; 5 cases] in a cohort of workers ( n=3,089) exposed to trichloroethylene, although results were imprecise [Anttila et al. 1995]. Similarly, another study found excess leukemia mortality (SMR= 4.9; 95% CI: 1.0-14.4; 3 deaths) in a cohort of white men ( n=2,610) exposed to perchloroethylene while employed in a U.S.

chemical company from 1956 to 1980 [Olsen et al. 1989]. Limited power was observed in all studies

reviewed, whereby no study reported more than 16 leukemia cases [IARC 1995].

Lastly, carbon tetrachloride was also widely used as a degreaser in several study facilities

[Mihlan 1997]. Although carbon tetrachloride-induced hepatocarcinogenesis has been observed in

several animal studies, the existing evidence is insufficient to conclude it is carcinogenic in humans (IARC

Group 2B). Few observational studies have examined cancer risks associated with carbon tetrachloride

exposure and results on leukemia risk are equivocal [Blair et al. 1998; Blair et al. 2003; Bond et al. 1987;

Ott et al. 1985; Wilcosky et al. 1984]. Of the five occupational studies recently reviewed by IARC [1999],

one nested case-control study reported elevated lymphocytic leukemia risk in rubber workers exposed

to carbon tetrachloride for more than one year (OR=15.3; p<0.01). However, individual exposure

measurements were not available (i.e., exposure surrogates were used), study participants were

potentially exposed to several (n=24) other solvents during the period under observation, and the

leukemia risk estimate was based on just eight cases [Wilcosky et al. 1984].

Studies specifically examining the combined effects of ionizing radiation and carbon

tetrachloride are sparse. Schubauer-Berigan et al. [2007a; 2007b] examined leukemia risks from carbon

tetrachloride in their recent study of Hanford, ORNL, PNS, and SRS nuclear workers. Carbon

tetrachloride exposures were assessed using a job-exposure matrix and algorithm to calculate

cumulative exposures scores for each study subject. Cumulative scores were stratified into three

exposure categories (none, low, and high) using a median-based cutpoints. About 25% of the study

subjects were believed to have been exposed to carbon tetrachloride. A slight elevation in leukemia

excluding CLL was observed among workers with the highest exposure compared with those with no exposure, although risk estimates were imprecise (univariate rate ratio =1.20; 95% CI: 0.73, 1.91)

[Schubauer-Berigan et al. 2007a]. The incorporation of a carbon tetrachloride variable in multivariable risk models did not appreciably change the estimate of risk from external radiation (i.e., <15% change on a relative scale). Rate ratio modification by carbon tetrachloride was not evident by likelihood ratio test

(LRT). Similar results were observed for CLL risks in this study population [Schubauer-Berigan et al.

2007b]

2.7.2.3.3 Others

There is a paucity of evidence on other potential chemical leukemogens. In particular, styrene,

1,3-butadiene, and ethylene oxide have been considered by some to be potentially leukemogenic

[Descatha et al. 2005]. Styrene (ethenylbenzene) is a monomer that is used in the manufacture of plastics, paints, synthetic rubber, and polyesters that has been linked to cancer (IARC group 2B).

Although some studies of styrene exposures have suggested leukemia elevations, there is no consistent epidemiologic evidence of an association between styrene and human leukemia [Boffetta et al. 2009].

1,3-butadiene (IARC group 2A), styrene, and dimethyldithiocarbamate (DMDTC), among others, are chemicals that are used extensively in the synthetic rubber-producing industry. Like styrene, metabolites of 1,3-butadiene are genotoxic. Several large studies have reported excesses in leukemia in

men working in the synthetic rubber industry [Cheng et al. 2007; Delzell et al. 1996; Graff et al. 2005;

Sathiakumar et al. 2005], but not in women [Sathiakumar and Delzell 2009]. Some studies suggested

associations were stronger for chronic leukemias (mostly CML and to a lesser extent CLL) than for AML

or ALL [Graff et al. 2005; Sathiakumar et al. 2005]. Alder et al. [2006] conducted a meta-analysis of

cancer risk in synthetic rubber workers and found elevated leukemia risk (RR=1.21; 95% CI: 1.03, 1.43)

by combining information in 16 occupational cohorts. This risk increased slightly when restricting

analysis to persons working exclusively in non-tire manufacturing cohorts (RR=1.70; 95% CI: 1.14, 2.54;

n=4). However, caution is needed when interpreting results because of mixed exposures to 1,3-

butadiene (IARC group 2A), styrene, DMDTC and others (such as benzene in some cases). Even when

attempts were made to disentangle exposure effects, the authors noted that chemical exposures were

highly correlated [Graff et al. 2005].

Ethylene oxide (IARC group 1) is used mostly in the chemical industry as a raw material in the

manufacture of chemicals for major consumer goods. Over half of the bulk ethylene oxide that is

produced world-wide is used to in the manufacture of monoethylene glycol, which is widely used in the

manufacture of polyester resins, films, and fibers and is a major ingredient of automobile antifreeze.

U.S. production in 2004 was about 4,000 MT [IARC 2008]. It is also used in industrial sterilization, as a

sterilant in hospitals, and in waste water treatment processes. NIOSH estimates that approximately

270,000 workers were potentially exposed to ethylene oxide from 1981 to 1983 [NIOSH 1990]. Ethylene

oxide is a direct-acting alkylating agent that is genotoxic and mutagenic and has been shown to be a

low-potency carcinogen in animal studies. IARC considers ethylene oxide to be carcinogenetic in

humans, although their decision was based mainly on findings from invitro and animal studies [IARC

2008]. There is limited information on ethylene oxide leukemogenicity from observational studies. A

large cohort study of workers ( n=18,235) employed at 14 industrial plants and followed through 1998

did not observe excesses in leukemia (SMR= 0.99; 95% CI: 0.71, 1.36), although there were few cases

(n=29) [Steenland et al. 2004]. Similarly, leukemia risks were not elevated (SMR=0.93; 95% CI: 0.47, 1.67;

n=11) in a cohort study by Swaen et al. [2009], who examined mortality patterns in men ( n=2,063)

employed at an ethylene oxide production facility in the period 1940-1988 and followed through 2003.

Pooling of these two cohorts was conducted by Valdez-Flores et al. [2010], who reported a combined

leukemia SMR of 0.97 (95% CI: 0.70, 1.33; n=40). Dose-response analyses by Cox proportional hazards

models of the pooled information did not reveal a positive trend of leukemia in dose.

Facility process records did not provide evidence of large-scale use of styrene, 1,3-butadiene, or ethylene oxide in operations conducted at the study facilities. Nevertheless, limited sampling data from industrial hygiene records was found that suggested exposures to styrene and ethylene oxide were possible in certain situations. For example, ORNL pipefitters were monitored for styrene exposures in

1988 while “sealing a swimming pool” [Mihlan 1997]. Likewise, LANL and Zia Company (Zia) painters assigned to the P-2 Group were monitored for styrene exposures while applying a surface coating known as “Ev-R-Shield liquid tile” in the early 1950s. Personal air monitoring conducted over the five- day operation indicated a maximum styrene airborne concentration of 750 ppm. Supplied air respiratory protection and local ventilation were required [Mihlan 1997].

2.7.3 Therapy-related myeloid neoplasms

Chemotherapy, radiation therapy, and therapeutic exposures to certain immunosuppressive agents (e.g., azathioprine) have been linked to AML. Chromosome lesions induced by alkylating agents

(primarily chromosome 5 and/or 7 monosomies or deletions) or DNA-topoisomerase II inhibitors

(primarily balanced translocations) that are used in chemotherapy are typical in secondary AML cases.

Risks are modified by primary cancer site and by therapeutic regimen. Moreover, given that only a small fraction of patients treated with these therapies present with AML, it is likely that the risk is influenced by a number of susceptibility factors such as polymorphisms of detoxification and DNA-repair enzymes

[Leone et al. 2010].

2.7.4 Genetic Factors

Heritable forms of leukemia are rare [Horwitz 1997] but some genetic conditions (e.g., Down syndrome, neurofibromatosis, ataxia telangiectasia) have been associated with increased acute leukemia risks. In particular, studies have shown a 20-fold increase in ALL among patients with Down syndrome [Bassan et al. 2004]. Recurrent somatic chromosomal translocations and inversions are

lesions most frequently associated with leukemia [Michaud et al. 2002]. Often these lesions involve chromosome 21q22.1 [i.e., t(8;21), t(3;21), and t(12;21)], which also contains the RUNX1 gene, one of only two loci linked to a familial predisposition for hematologic malignancies. Germline mutations in

RUNX1 cause an autosomal dominant familial platelet disorder in which affected individuals also have a

35% increased risk of developing AML [Jongmans et al. 2010]. Overall, it is estimated that approximately

5% of acute leukemia cases are associated with an inherited genetic syndrome that most likely involves

genes linked to genomic stability [Linet et al. 2006].

Genetic and familial predisposition appears to be strong in CLL cases [Goldin et al. 2009; Goldin

et al. 2010]. In fact, familial tendency toward CLL has been acknowledged for over 50 years as one of the

most important risk factors for developing the disease [Linet et al. 2006]. Deletions are observed

frequently in chromosomal regions 13q14 (50-60%), 11q22-q23 (10-20%), 17p13 (5-10%) and trisomy of

chromosome 12 (15-25%) in most (>80%) CLL cases, although few genes have been linked to the disease

[Coll-Mulet and Gil 2009; Dohner et al. 2000]. However, a germ-line mutation found in familial CLL has

been linked to the silencing of gene ARLTS1, which has a possible role in apoptosis [Calin et al. 2005].

Interestingly, a small subset of CLL patients has been found with this gene silenced by promoter

methylation, which may suggest that CLL is a disease involving both epigenetic and genetic

abnormalities.

In contrast, evidence of familial CML is sparse and is limited to a few clinical reports [Kapsali et

al. 2000; Lessen et al. 2005]; thus, it appears that familial CML is extremely rare [Linet et al. 2006].

Nevertheless, there is some evidence of familial clustering of CML in first degree relatives of patients

afflicted with myeloproliferative neoplasms (NPN). Landgren et al. [2008] conducted a population study

of first-degree relative cases ( n=24,577) and controls ( n=99,542) in Swedish residents and found that the

relatives of MPN patients had a two-fold increase risk of CML, although estimates were imprecise (RR=

1.9; 95% CI: 0.9, 3.8).

3.0 Study Cohort

3.1 Definition

The study design is intended to maximize the use of information from previous observational studies. To that end, nuclear workers were drawn from the NIOSH Occupational Energy Research

Program (OERP) master roster, which contains demographic and vital status information on about

500,000 potential study participants ever employed in at least one of several DOE facilities or a nuclear shipyard. Using criteria previously specified by Schubauer-Berigan et al. [Schubauer-Berigan et al.

2007a], the study population was restricted to studied workforces characterized by: 1) radiation exposures that were predominantly from external whole-body irradiation by photons; 2) adequate person-time to contribute a substantial (>10) number of leukemia cases; and 3) sufficient availability of dosimetry, employment, and operational records to conduct the exposure assessment. Facilities meeting these criteria were: 1) the Hanford Site (Hanford) near Richland, Washington [Gilbert et al.

1993]; the Savannah River Site (SRS) near Aiken, South Carolina [Cragle et al. 1999]; the Oak Ridge

National Laboratory (ORNL) in Oak Ridge, Tennessee [Frome et al. 1997]; the Los Alamos National

Laboratory (LANL), including the Zia Company employees, in Los Alamos, New Mexico [Galke et al. 1992;

Wiggs et al. 1994]; the Idaho National Laboratory (INL) near Idaho Falls, Idaho [Schubauer-Berigan et al.

2005]; and the Portsmouth Naval Shipyard (PNS) in Kittery, Maine [Silver et al. 2004].

To be eligible, each study participant must have been monitored for radiation exposure and

employed at one or more of the primary facilities for at least 30 days. Prior to the 1950s, nuclear

weapons plant operations, exposure monitoring and control practices, worker demography, and worker

lifestyles appeared to be very different than that found in the Cold War years and later [Frome et al.

1990]; therefore, the selection of workers employed at Hanford, LANL, or ORNL was further restricted to

workers hired on or after January 1, 1951.

3.1.1 Vital Status Ascertainment

Considerable efforts were made to uncover additional information on study subjects to improve the likelihood of matches to various databases for vital status information. The study roster was uploaded into the NIOSH Integrated Data Management System (IDMS) and linked to the Death Master

File (DMF) maintained by the Social Security Administration (SSA) as a preliminary step to gauge the sufficiency of matching information. Questionable matches were identified by IDMS algorithms and reconciled by hand by two independent researchers. Reconciliation included searches of facility records, compensation claims, and public records to gather sufficient information to accurately identify study subjects. Subsequent updates to the roster were made using these data (Figure 3-1).

Initial vital status information was obtained under the previous studies and was available through: 1990 for LANL, Zia, and ORNL; 1994 for Hanford; 2002 for SRS; 1996 for PNS; and 1999 for INL.

In the previous studies, death certificates were coded to the 9 th revision of the International

Classification of Diseases (ICD-9) for INL; ICD-8 for LANL, Zia, and ORNL; to a mixture of ICD-8 and ICD-9

for Hanford; and to the ICD revision in effect at the time of death for PNS and SRS. Only the underlying

cause of death (UCOD) was available for early followup of the combined LANL cohort (i.e., LANL and Zia

workers), thus only the UCOD was used to identify leukemia cases for this study.

Followup was extended through December 31, 2005, by linkage to the National Death Index-

Plus (NDI Plus ) maintained by the National Center for Health Statistics (NCHS), which provides

information on the causes of death beginning in 1979 ( http://www.cdc.gov/nchs/ndi.htm ). The roster

was also matched to the California Death Index, which was a useful means to uncover cause of death

information for some subjects who had died prior to 1979. Study subjects not matched to decedent

records and lost to follow-up were searched against records maintained by the Internal Revenue Service

(IRS) to confirm alive status. This step was especially important to extend the date last observed (DLO)

for those workers who left employment prior to 1979. In general, it was assumed that followup information from the major mortality databases was essentially complete for deaths after 1978; therefore, workers who were employed after 1978 but not found in NDI or SSA-DMF were presumed living. Workers terminating employment prior to 1979 and not found in SSA-DMF or IRS searches were assigned the date last employed (DLE) as the DLO.

It is noteworthy that considerable efforts were expended by current and previous researchers to uncover vital status information from a wide array of available resources (e.g., NDI, SSA, IRS, pension

boards, health insurers, state death indices, and death certificate searches). In particular, NDI and SSA

were primary sources of death information in this study, whereby most causes of death were identified

in NDI Plus . The accuracy of these databases has been studied extensively, and a wealth of information

exists on their wherewithal to provide highly accurate data for epidemiologic research [Boyle and

Decoufle 1990; Cowper et al. 2002; Doody et al. 2001; Wojcik et al. 2010]. Nevertheless, complete and

infallible ascertainment is unlikely in most studies because of existing imperfections in the information

sources used, poor linkage to the study roster, or both. Therefore, false- or underreporting of some

deaths, especially in the years prior to NDI (i.e., <1979), is an inevitable limitation of this study.

LCCS cohort INL cohort (n=94,517) (n=63,560)

exclusions: ( n=51,841)

duplicates: ( n=991)

Base cohort (n=105,245)

Reconcile Upload to questionable NIOSH IDMS matches

? Match to SSA DMF Match

N Y NDI Those without UCOD

Reconcile ? N questionable Match Assumed alive matches

VS Followup complete

Figure 3-1. Cohort assembly and vital status followup flowchart. Additional searches against the California Death Index and the IRS database were used to confirm alive status of non-matches to NDI.

3.1.2 Results

The previous study cohorts consisted of 94,517 workers from Hanford, LANL, ORNL, PNS and SRS

(i.e., LCCS cohort) and 63,560 INL workers (i.e., INL cohort). Applying the inclusion criteria and then combining the two cohorts resulted in a base cohort of 105,245 that was followed through 2005. There were 22,786 persons excluded from the LCCS cohort because of employment prior to 1951. The remaining LCCS exclusions ( n=32) were because of not meeting other inclusion criteria or a lack of demographic (e.g., missing date of birth) or exposure information. Nearly all (n=27,258) of the INL cohort exclusions (n=29,022) were non-monitored workers. Finally, there were 991 duplicates removed after combining the LCCS and INL cohorts.

There were 30,266 deaths (28%) in the base cohort, which included 8,721 cancer deaths and

369 known leukemia cases (Table 3-1). Cause of death information was available for 29,843 decedents

(98.6%), although 236 of these deaths (<1%) were coded as ill-defined, unspecified, or unknown. Most deaths occurred after 1978 (n=25,269, 83.5%); however, slightly more than half the decedents who lacked a valid UCOD (n=423) were verified as deceased prior to 1979 ( n=262). Of decedents lacking

cause of death information, 161 were post-1978 deaths that were linked to the SSA DMF but were not

found in subsequent NDI searches. A small number of decedents lacked a valid date of death (n=41), and a disproportionate number of these deaths were among women ( n=16, 39.0%).

Most cohort members were male (81.5%) and White (80.8%). Information on gender was nearly

complete (>99%) while race information was absent for about 25% o f the cohort. The completeness of

information on gender and race varied among sites, although SRS and LANL/Zia subjects were attributed

most of the missing gender information (98.9%) and over half of the missing race information (59.5%).

No cohort exclusions were made based on missing gender or race information because additional

information to resolve these data gaps were likely available at the level of cases and/or controls .

Table 3-1. Characteristics of the base cohort assembled for case-control selection

Main sites All Hanford INL LANL/ Zia PNS SRS X10 Subjects n 105,24 5 23,177 33,47 3 12,296 9,657 12,585 14,057 % 100 22.0 31.8 11.7 9.1 12.0 13.4 Sex male 85, 726 16,672 27,80 2 10,343 9,619 10, 868 10,422 male% 81. 5 71.9 83.1 84.1 99.6 86.3 74.1 female 19,158 6,503 5,667 1,68 8 38 1, 62 7 3,635 unknown 361 2 4 265 0 90 0 Race White 85,041 21,666 28 ,619 6, 953 9,607 5,284 1,291 2 White% 80. 8 93.5 85. 5 56.5 99.5 42 .0 91. 9 other 4, 214 1,491 1,02 7 16 8 22 599 90 6 unknown 15,991 20 3,8 27 5, 175 28 6,702 239 Median birth 1936 1938 1942 1937 1926 1928 1937 year Median year 1965 1967 1974 1967 1959 1953 1965 first employed Median Years 8.4 6.3 7.3 9.9 23.8 11. 5 4.0 employment Employment 1951 - 1949 - 1951 - 1952 - 1951 - 1951 - eligibility 1978 1999 1978 1977 1974 1978 Deceased n 30,26 6 5,684 7,33 2 3,041 4,995 5,932 3,2 82 Site pop. % 28. 4 24. 5 21. 4 24. 4 51. 3 46.6 23.0 Tot. deaths % 18. 8 24. 2 10.1 16. 5 19. 6 10.8 Cancer deaths n 8,72 1 1,70 5 2,05 0 794 1,503 1,681 98 8 Leukemia n 369 58 90 46 48 78 49 deaths Missing UCOD n 423 55 176 37 39 67 49

3.2 Facility Descriptions

This study uses information from five nuclear facilities operated by the DOE and its predecessor agencies and a U.S Navy operated nuclear shipyard. Descriptions of these facilities are provided in the following sections.

Of the selected DOE facilities, three (i.e., ORNL, LANL, and Hanford) were commissioned during

World War II under the Manhattan Engineering District (MED) with the primary directive to develop the world's first nuclear weapons. During MED operations (1942-1946), radiation exposure standards were based on the prevention of deterministic effects and used the concept of a tolerance dose that limited

daily exposures to 0.1 roentgen, i.e., approximately 1.0 mGy∙d -1 or 250 mGy∙y -1 based on a 250-day

working year [Edwards 1991]. The popular concept of keeping radiation exposures ALARA would not be

fully adopted until many years later [Auxier and Dickson 1983].

Two DOE study facilities (INL and SRS) began operations under the Atomic Energy Commission

(AEC), which superseded the MED in January, 1947. Both facilities began handling radioactive materials

in the early 1950s at a time when the leukemogenicity of ionizing radiation was becoming more

apparent and health physics practices had been refined from lessons learned in previous years. By 1954,

most AEC facilities had adopted controls that prevented external exposures to any individual from

exceeding 3.0 mSv in a week (150 mSv in a working year) [Edwards 1991]. Growing concerns over

radiation-induced hereditable effects and cancer prompted further reductions in exposure limits from

1950 to 1960. By 1958, most U.S. nuclear facilities (including PNS) limited allowable lifetime

accumulated doses (mSv) to blood-forming organs to no more than 50 times the worker’s age in years.

This resulted in a prospective annual limit of 50 mSv∙y -1, although the AEC continued to use a dose limit of 30 mSv∙qtr -1 (i.e., 120 mSv∙y -1), provided that the lifetime limit was not violated. By the early 1970s,

scientific concern was clearly greatest for somatic effects of ionizing radiation, which prompted the AEC

to adopt the current 50 mSv∙y -1 annual limit in 1974 [Horan and Braun 1993]. Subsequently, the notion of LNT and acceptable risk were forefront in radiation protection, which prompted the decline in worker exposures without a concomitant reduction in regulatory limits.

300

SRS (1952) and 250 INL (1951)

200

PNS (1958) 150 ORNL (1943) LANL (1943) 100 Hanford (1944) Whole-bodydose (mSv)

0 1925 1935 1945 1955 1965 1975 1985 1995 Year

Figure 3-2. Annual limits on radiation dose to blood forming organs used by study facilities at time of start-up of radioactive work.

Poor characterization of workplace exposures and increased occurrences of overexposures were likely during the MED era because of the urgency of the program mission and the limited knowledge of health physics and radiation protection methods. Moreover, stiff competition from the Armed Services and other employers that were mobilized for the war effort had largely affected the population available for hire. Thus, some studies have shown differences in the mortality experienced in workers recruited from 1940 to 1945 compared to workers hired prior or after [Bond et al. 1989; Frome et al. 1990; Wen et al. 1986]. For example, more than half of the MED-era Hanford employees were hired at age 38 or older and three-quarters of those from 18 to 26 years old were medically disqualified from military service [Marceau et al. 2002]. Also, Frome et al. [1990] did not find evidence of a healthy worker effect

(e.g., all-cause SMR=1.11, P<0.01) in a mortality study of World War II Oak Ridge nuclear workers, although this selection bias is routinely observed in studies that included nuclear workers hired after the

war [Fry et al. 1995]. Similar results were observed in a cohort ( n=995) of white male War II-era uranium workers employed from 1943 to 1949, whereby mortality from all causes (SMR=1.18; 95% CI: 1.07-1.30) was significantly elevated compared to the U.S. population [Dupree et al. 1987]. To control for the effects of differences between MED-era workers and later workers, the study cohort was limited to workers who were hired at a primary facility on or after January 1, 1951. However, contributions to lifetime doses that occurred prior to the hire date and any exposure-lag period were included in the reconstructed dose for all study subjects.

3.2.1 Hanford

In 1943, the MED selected a 670 mi 2 area in the remote southeastern region of Washington

State along the Columbia River for the construction of “Site W,” its large-scale plutonium production

facility. The Hanford site was divided into three major operational areas: the 100 Areas (100-B, 100-C,

etc.) were designated for reactor operations, the 200 Areas were for reactor fuel reprocessing and

plutonium separations, and reactor fuel fabrication was performed in the 300 Area. Construction of the site, known as the Hanford Engineering Works (HEW) during the war years, began in March, 1943, by E.I.

Du Pont de Nemours and Company (hereafter referred to as DuPont). By the end of the war, the HEW was comprised of 554 buildings including three operating water-cooled, graphite moderated, plutonium production reactors [Reactors B (1944-1968), D (1944-1967), and F (1945-1965)], three chemical separations plants (221-B, 221-T, and 221-U), and 64 single-shell underground storage tanks in four

“tank farms” (241-B, 241-C, 241-T, 241-U) used for high-level radioactive waste storage. Large-scale plutonium production was underway by the end of 1944 and the first shipment of freshly separated plutonium to LANL was made in February of 1945 [Marceau et al. 2002].

With the creation of the AEC following the war, the HEW was renamed the Hanford Works

(1947-1974) and construction of new facilities was begun to meet the increased plutonium production

demands of the Cold War. Two new reactor complexes [DR Reactor (1950-1964) and H Reactor (1949-

1965)] and the Plutonium Finishing Plant (234-5Z) were placed in operation by 1950. Construction of the

C Reactor (1952-1969) complex began June 6, 1951. The REDOX solvent-extraction plant (202-S) went online in January 1952 for improved plutonium recovery from irradiated fuel. Uranium recovery operations took place in 221-U and 224-U plants. Also during this time (1949-1952), Hanford began its tritium production program in support of the development of the first thermonuclear device [Marceau et al. 2002].

Cold War expansion of Hanford facilities continued during the Korean War era. Two new reactor complexes [KW Reactor (1954-1970) and KE Reactor (1955-1971)] were operational by early 1955, as was the last separations plant constructed, the Plutonium-Uranium Extraction (PUREX) Plant (202-A

Building). Startup of the Plutonium Recycle Test Reactor (309 Building) took place in 1960, followed by the startup of N Reactor (1964-1987). The N Reactor was the last of nine plutonium production reactors constructed at Hanford and the first to have a dual purpose; it also included a steam plant that produced usable electrical energy. Reactor operations ceased at the Hanford Site with the shutdown of the N Reactor complex in 1987, signaling the end of plutonium production and the beginning of environmental restoration [Marceau et al. 2002].

In all, the nine operating production reactors at the Hanford Site produced about 67.6 metric tons of weapons- and fuel-grade plutonium from 1944 to 1987 (Figure 3-3). Production rose sharply beginning in the early 1950s and peaked at an annual production rate of about 4.7 MT in 1964, when all nine reactors were producing. By 1972, all but one reactor (N Reactor) was deactivated. Reactors B, C,

DR, H, KW, and N were also used to produce about 13.0 kg of tritium from 1950 to 1971 [Lini 1993].

5 5

4 4

3 3

2 2 plutonium(MT)

1 1 Tritiumproduction (kg)

0 0 1940 1945 1950 1955 1960 1965 1970 1975 1980 1985 1990 Year

Figure 3-3. Hanford plutonium (solid blue line, in MT) and tritium (dashed red line, in kg) production (1945-1987).

Radiation work at Hanford primarily involved plutonium production operations including: 1) fuel

fabrication; 2) irradiation of fuel elements; 3) chemical separations of uranium and/or plutonium from

fuel elements and waste products; 4) high-level radioactive waste, handling, and storage; 5) plutonium

handling and refinement; and 6) supporting research and development. The most significant exposures

from penetrating low-LET radiations occurred at the reactors and chemical separation facilities. Neutron

radiation exposures were common to the plutonium handling facilities and, to a lesser extent, reactor

operations [Fix 2004; Selby 2003].

Employment increased with production, with an observable plateau in the 1960s (Figure 3-4).

However, employment declined beginning in 1963 following plans to close three reactors. This decline

continued until the early 1970s, when construction on the Fast Flux Test Facility began. An agreement to

continue operating the N reactor for electrical power was reached in 1972, resulting in further increases

in employment through the 1980s. Nevertheless, the shutdown of N Reactor in 1987 marked the end of

production, which is evident by a steep decline in the number of Hanford workers shortly thereafter

[Marceau et al. 2002].

18 16 14 12 10 8 6 4 2 Number of Numberworkers (thousands) 0 1940 1950 1960 1970 1980 1990 2000 Year

Figure 3-4. Hanford site employment numbers (1944-2000)

3.2.2 INL

INL is located about 30 miles west of Idaho Falls and 14 miles southeast of Arco, on a 2,350 kmi 2 plot of land in the Snake River Plain of eastern Idaho. The site began operations under the AEC in 1949 as the National Reactor Testing Station (NRTS) with the primary mission to build, test, and operate nuclear reactors. The NRTS was renamed the Idaho National Engineering Laboratory (INEL) in 1974, which later became the Idaho National Engineering and Environmental Laboratory (INEEL). On February

1, 2005, the site was renamed the Idaho National Laboratory (INL) to combine the research mission of the INEEL and ANL-W facilities with site closure activities under the Idaho Cleanup Project (ICP).

NRTS workers did not begin handling radioactive materials until late 1951, when the first reactor

[Experimental Breeder Reactor 1 (EBR-1)] became operational. Ionizing radiation exposure was primarily a consequence of reactor operations and maintenance activities at test reactor sites. The site brought online about one to two new reactors of varying designs each year through the mid 1950s. The first criticality accident occurred on November 29, 1955 at EBR-1 and the first recorded over-exposure was

recorded on July 23, 1956, when a maintenance worker received approximately 216 mSv whole-body exposure. The highest individual exposures occurred as a result of the SL-1 reactor accident on January

3, 1961, whereby nine immediate responders received whole-body doses in excess of 120 mSv. The SL-1 reactor was a prototype low-power boiling water reactor that was designed to be used in remote locations. The reactor suffered a catastrophic steam explosion following inadvertent prompt criticality, resulting in the deaths of three workers. SL-1 clean-up activities took place for the next two years and accounted for 1.3 person-Sv in 1961 alone, which was nearly one-half of the total exposure for the site in that year [Horan and Braun 1993]. Over the years, 52 reactors have operated at the site, which is more than anywhere else in the world. Of these reactors, 40 were built prior to 1964 and all but two were complete by 1973. However, only 16 reactors remained in operation as of June of 1973.

INL also operated a chemical processing plant that reprocessed spent nuclear fuels to recover

235 U. This plant, called the Idaho Chemical Processing Plant (ICPP or CPP), was operated from 1953 to

1988. During that time, approximately 31,000 kg of enriched uranium was recovered. The primary radiological hazards in the ICPP were external whole-body irradiation from the decay of aged mixed fission products. Internal deposition of fission products, along with enriched uranium isotopes and transuranics in specific process locations, were also of concern at the ICPP. The ICPP experienced a criticality accident on October 16, 1959, during inadvertent transfer of an enriched uranium solution. A second criticality at the ICPP occurred on January 25, 1961, as the result of improper uranyl nitrite transfer. The ICPP experienced its third criticality accident on October 17, 1988. Remarkably, none of these accidents resulted in individual whole-body penetrating exposures in excess of 100 mSv

(maximum was 80 mSv) [Ginkel et al. 1960; McLaughlin et al. 2000].

The Argonne National Laboratory-West (ANL-W) area is near the eastern boundary of INL and its primary mission has been in support of the fast breeder reactor program. ANL-West facilities include the

Experimental Breed Reactor 2 (EBR-2), the Transient Reactor 2 (TREAT), the Zero Power Plutonium

Reactor (ZPPR), and the Hot Fuels Examination Facility (HFEF). In total, nine experimental reactors were under the direction of ANL-W.

The Naval Reactor Facility (NRF) is located on the west-central side of INL and covers approximately 80 developed acres. NRF workers were involved in the design, development, and testing of naval nuclear propulsion systems and the training of U.S. Navy nuclear workers. The facility is comprised of the Submarine Prototype (S1W), operated from 1953 to 1989; the Large Ship Reactor

(A1W), operated from 1958 to January 1994; and the Natural Circulation Submarine Prototype (S5G), operated from 1965 to May 1995. The NRF also housed the Expended Core Facility (ECF), which provided for testing and inspection of naval expended fuel assemblies prior to uranium recovery at the

ICPP [INEL 1973].

4 Number of Numberworkers (thousands) 2

0 1950 1960 1970 1980 1990 Year

Figure 3-5. Employment at INL (1951-1990)

3.2.3 LANL

LANL, [previously the Los Alamos Scientific Laboratory (LASL), 1943-1981], occupies approximately 43 mi 2 (111 km 2) of U.S. government land in northern New Mexico. The laboratory itself

is actually divided into 47 separate technical areas (TAs) collectively containing about 1,800 buildings,

although most laboratory personnel are concentrated in the main technical area (originally TA-1, now

TA-3, circa 1950) [Cruz 1998].

Initially established as “Project Y” in 1943 by the MED, the laboratory’s primary mission was to

design and assemble the world’s first nuclear weapons. Throughout the years, LANL’s responsibilities

have focused on research and development such as nuclear and thermonuclear weapons design, high

explosives and ordnance development and testing, weapons safety, nuclear reactor research, waste

disposal, chemistry, criticality experimentation, tritium handling, biophysics, and radiobiology. In many

cases, operations, facilities, and capabilities to support development and production of the various types

of nuclear devices expanded to support other missions after World War II. Since the end of the Cold

War, LANL scientists and engineers have shifted toward research and development supporting a broad

range of other programs, including non-nuclear defense and nonproliferation; nuclear and non-nuclear

energy; atmospheric, space, and geology sciences; bioscience and biotechnology.

Throughout the period under observation, all laboratory operations were managed by the

University of California. During the war years, the workforce was comprised mainly of military

personnel, civilian technical staff, and some Sandia Laboratory personnel. By 1946, the military

compliment, who was instrumental in the initial construction and maintenance of early facilities at LANL,

had essentially disappeared; therefore, the Zia Company (1946-1986) was formed as the primary

maintenance subcontractor beginning in April, 1946 [Truslow and Thayer 1973]. A subsidiary of the Zia

Company, Los Alamos Constructors, Inc. (LACI), was responsible for construction of the town and

laboratory facilities. Information on annual employment was sparse; however, a number of Group H-3

(Safety) reports provided estimates of person-hours worked for each year between 1945 and 1985

(Figure 3-6). During most of this time, the Zia Company employees comprised from 15 to 25% of the total LANL workforce.

160

140 x 100000 x

120

100

80 Person-hours 60

0 1945 1950 1955 1960 1965 1970 1975 1980 1985 Year

Figure 3-6. LANL and Zia Company person-hours worked per year (19450-1985)

LANL workers were potentially exposed to a wide array of nuclear and chemical hazards as a result of diverse operations at the laboratory throughout its operating history. Radiation exposures were likely from high- and low-LET irradiation by sources both internal and external to the body. These sources include, but are not limited to, primary weapons materials (plutonium, uranium, polonium, and tritium); fission products; and isotopic sources and tracers (e.g., radium, lanthanum). Some of the more important radiation sources are described below.

3.2.3.1 Metal processing

At facility startup in 1943, LANL was responsible for research and development of weapons- grade plutonium, including its final purification, reduction to metal, and weapon component manufacture. Plutonium processes began in the D Building in the original technical area (TA-1) but were moved to the DP West area (TA-21) by the end of August 1945 [Christensen and Maraman 1968]. DP

West consisted of several buildings, each designed for a specific step in plutonium processing [i.e., ether extraction (Building 2), oxalate precipitation (Building 3), fluorination (Building 4), and metal fabrication

(Building 5). Research and development was conducted in Building 22. From 1945 to 1978, plutonium

operations at DP West involved plutonium extraction, dissolution, recovery, and americium waste

storage. Plutonium work was relocated to the Plutonium Facility (TA-55) in 1978. Some plutonium work

was also conducted in the New Core Area (TA-3) following the transfer of the main technical facilities in

1953. In particular, researchers in TA-3’s Chemical and Metallurgical Research (CMR) building (1952 to

current) handle laboratory quantities of uranium, thorium plutonium, and other nuclear materials for

experiments in analytic chemistry, metallurgy and materials science [Garcia et al. 2009].

Working with fissile materials in DP West was extremely hazardous, as evidenced by a number

of serious radiation accidents over the years. Most of these accidents involved containment breaches

resulting in localized plutonium exposures, although more widespread contamination was evident from

some drybox explosions and fires. On December 30, 1958, LANL suffered its third radiation-induced

fatality from a nuclear criticality accident involving plutonium recovery operations at the DP West site.

Estimated doses to surviving workers ranged from <40 mSv to 0.134 Sv [Paxton et al. 1959].

3.2.3.2 RaLa experiments

The Bayo Canyon Site (TA-10) was the location of experiments examining implosion

characteristics using high explosives, radiolanthanum (RaLa), and uranium. From September 1944

through March 1962 there were 254 of these hydrodynamic tests conducted resulting in widespread contamination from fallout [Dummer et al. 1996]. Maximum exposures to workers who cleaned the firing pad after test shots typically approached 3.0 mGy each day of active cleaning. Exposures to personnel observing RaLa tests were comparatively smaller, whereby maximum doses for observers were typically <1.0 mGy per shot.

The test assemblies contained high explosive (HE) charges from 20 to 750 lbs TNT-equivalent and a short-lived tracer (140 La ; 40.0±0.3 hour half-life) from 2.2 to 26.0 TBq. Other radionuclides ( 140 Ba,

89 Sr, and 90 Sr) were available in lesser amounts [Dummer et al. 1996]. The RaLa sources were prepared

in the TA-10 RaLa Chemistry Building from 1944 to 1950 and then in the TA-35 “Ten-Site” facility from

1951 to 1963. The first significant amounts of RaLa arrived in 1944 by truck transport following

manufacture in the Graphite Reactor at ORNL (Clinton Laboratory). Chemists at Bayo prepared each

source by separating (“milking”) the 140 La from 140 Ba and other impurities. Separation activities were initially conducted locally; therefore, overexposures were likely. In fact, there was evidence of overexposures to Chemistry Group workers in the RaLa Chemistry Building (25 to 50 mGy), Ordinance

Group workers who prepared charges for test assemblies (∼10 mGy), and various support personnel who handled these sources [Hempelmann 1944]. These early incidents prompted changes in radiation safety procedures at Bayo and the relocation of the separations process in 1951 to the newly constructed “hot cells” at Ten-Site. The hot-cells were designed for remote operations of high-dose rate sources maintained within a biological shield. Nevertheless, Bayo Canyon work continued to be the source of worker overexposures. For example, a serious exposure event took place on June 21, 1956 as a result of a faulty RaLa source resulting in protective clothing contamination in excess of 100 mGy∙hr -1

and whole-body doses to affected workers in excess of 40 mGy [Buckland 1957]. In the same report,

Buckland [1957] points out that nearly one-half of the individual external radiation overexposures in

1956 (n=57) occurred in Bayo Canyon. During this time period, an “overexposure” likely referred to a

dose greater than the weekly tolerance dose used by LANL during that period (i.e., 30 mGy∙wk -1 from

1948 to mid-1958).

3.2.3.3 Criticality Experiments

The Omega Site (TA-2) was home to LANL nuclear criticality experiments until moved to the

Pajarito Site (TA-18) in 1946 [Paxton 1983]. The move was prompted by the nation’s first criticality accident involving a fatal radiation injury, which took place at TA-2 on August 21, 1945. The accident involved a bare and reflected metal assembly with a total core mass of 6.2 kg of plutonium. Remarkably, a second radiation-induced fatality happened about one year later (May 21, 1946) while performing experiments involving the same critical assembly. Doses to personnel who were nearest the accident sites and survived ranged from 370 mSv to 3.6 Sv [McLaughlin et al. 2000]. These two accidents, and a preceding accident (June 4, 1945) that also resulted in serious radiation exposures (74-660 mSv), were the result of operator error during hands-on manual operations. Thus, a new criticality experiment facility, the Los Alamos Critical Experiments Facility (LACEF), began operations in 1947 that included a concrete-reinforced Integral Assembly Building (Kiva) for containment and remote handling equipment.

This Kiva design enabled remote handling of critical assemblies from a control room that was located about one-quartile mile away. In all, three Kivas would be built and operated at the Pajarito Site [Paxton

1983].

From 1947 to 1955 several early weapons-related critical assemblies (e.g., Topsy, Lady Godiva, and Jezebel) were used for experiments to gather neutronic information. By May 1967, there were at least eight accidental prompt-critical excursions, excluding the May 21, 1946 accident described earlier

[McLaughlin et al. 2000]. Information on dose values or the number of persons potentially exposed as a result of the excursion (e.g., equipment repair and decontamination) was sparse but suggested that significant exposures did occur from these events. For example, three persons received up to 135 mGy

each while working with activated components following a super-critical event involving the Godiva

Assembly on February 12, 1957 [Buckland C. W. 1958]. Moreover, there was evidence that some critical experiments caused increased neutron and gamma radiation fields in occupied areas surrounding the

Pajarito Site and occasionally resulted in releases of activation and fission products from the Kivas

[Paxton 1983].

Weapons-related criticality experiments continued at the Pajarito site after 1955; however, most work in the period from 1955 to 1972 focused on the Rover nuclear rocket propulsion development. During this time, critical experiments were conducted in support of mock-up test of several reactor designs for propulsion [Paxton 1983]. More recent operations at LACEF supported research and development for the National Aeronautics and Space Administration (NASA) and other non-weapons related critical experiments, safety studies, and criticality safety training. The last critical experiment was conducted July 8, 2004 [Loaiza and Gehman 2006].

3.2.3.4 Reactor operations

There were three homogeneous reactor systems operated at the Omega Site: the “Water

Boilers”, which actually consisted of three different designs (Hypo, Lopo, and Supo) operated from 1943 to 1974 [Montoya 1991]; “Clementine”, a plutonium fast reactor operated from 1946 to 1952 [McCool

1956]; and the Omega West Reactor (OWR), which operated from 1956 to 1992 [Cruz 1998]. Elevated radiation exposures from early reactor testing and operations were documented in several LANL reports. A criticality accident occurred in the water boiler reactor in 1949 resulting in a 25 mGy dose to the operator [McLaughlin et al. 2000].

In 1953, LANL began design and construction of the first of two high temperature uranium- fueled homogeneous test reactors supporting the Los Alamos Power Reactor Experiment (LAPRE I; 1956-

1957) located in TA-35 [Peterson 1959]. LAPRE I was replaced by a second reactor (LAPRE II; 1959),

which also experienced a number of operational problems that necessitated shutdown. The LAPRE program was discontinued in 1960 [Clark 1960]. Also located in Ten-Site was the Los Alamos Molten

Plutonium Reactor (LAMPRE); a one-megawatt gas-cooled reactor which achieved criticality on April 3,

1961. LAMPRE I (1961-1963) was intended to examine the use of molten plutonium as fuel in a fast breeder design incorporating molten sodium as a coolant. Another test reactor, the Ultra-High

Temperature Reactor Experiment (UHTREX), was constructed in the late 1960 in TA-52. This three- megawatt graphite-moderated and helium-cooled reactor was operated for about one-year then shutdown in February 1970 [Salazar and Elder 1993].

3.2.4 ORNL

Construction on the Clinton Engineer Works (MED codename “Site X”) began in early 1942 on a

59,000-acre (146,000-hectare) tract of land in west of Knoxville, Tennessee, between the Black Oak

Ridge to the north and the Clinch River to the south. The early site consisted of two plants that competed in the research, development, and large-scale production of enriched uranium (K-25 Plant and Y-12 Plant); and a pilot plant for plutonium production (X-10). The latter facility, named the Clinton

Laboratories during the war years, was completed in November 1943. The site was renamed ORNL in

1948 shortly after the formation of the AEC.

Much of the early work at ORNL was devoted to the development and operation of the Oak

Ridge Graphite Reactor (OGR; also known as the Oak Ridge Pile, and X-10 Pile), a graphite-moderated nuclear reactor that was the precursor to larger plutonium production reactors at the Hanford Site. The

OGR (1943-1963) was used to produce gram quantities of plutonium, and later, fission products and other radioisotopes for continued research and commercial use. Criticality was achieved in October of

1943 and the sufficient plutonium was chemically separated from spent uranium fuel to make the first shipment to Los Alamos in February, 1944. ORNL provided pilot testing of several processes supporting

nuclear weapons development. These operations included methods to separate plutonium, uranium, and thorium from irradiated fuels [Dickens and Fleming 2003].

ORNL has had an important role in the development of new reactor technologies. Including the

X-10 Pile as the world’s first operational reactor, there were 13 nuclear reactors designed and constructed at ORNL over the time under observation (Table 3-2). In all, there were : two aqueous homogeneous reactors and two molten-salt reactors that supported reactor development programs; seven light-water reactors used mainly for isotope production; one fast-burst reactor used in health physics studies; and the OGR. Twelve of these reactors were operated at ORNL for at least a year. Of these, only the High Flux Isotope Reactor (HFIR) remains in service at ORNL [Rosenthal 2009].

Table 3-2. ORNL reactors

Design Name Years operated Air -cooled graphite moderated Graphite Reactor (OGR) 1943 -1963 Aqueous homogenous Homogenous Reactor Experiment (HRE) 1952 -1954 Homogeneous Reactor Test (HRT) 1957 -1961 Molten salt Aircraft Reactor Experiment (ARE) 1954 -1955 Molten Salt Reactor Experiment (MSRE) 1965 -1969 Fast burst Health Physics Research Reactor (HPRR) 1963 -1987 Water -cooled and moderated Low -Intensity Test Reactor (LITR) 1950 -1968 Bulk Shielding Reactor (BSR) 1950 -1987 Geneva Conference Reactor 1 1955 -1994 Tower Shielding Reactor I (TSR -I) 1954 -1958 Tower Shielding Reactor II (TSR -II) 1958 -1987 Oak Ridge Research Reactor (ORR) 1958 -1987 Hig h Flux Isotope Reactor (HFIR) 1965 -present 1The Geneva Reactor was assembled and tested at ORNL then shipped to Switzerland for operations.

The expanding use of radioisotopes resulted in development of new processes and products in support of the isotope-distribution program beginning in 1946. ORNL workers were tasked with the separation, packaging, and distribution of radioisotopes for government and commercial use. Processes were developed to recover and purify radioactive gases (i.e., 133 Xe, 37 Ar, 85 Kr, etc.), fission products (i.e.,

137 Cs, 99 Tc, 237 Np, and 147 Pm, etc.), and a host of other short-lived and long-lived radioisotopes [Dickens

and Fleming 2003].

Consistent with the experiences at the other National Laboratories, the diverse nuclear

operations at ORNL resulted in the potential for worker exposures to low- and high-LET radiations from

a wide array of source terms. Although protracted low-dose external exposures were most common,

significant internal dose and acute exposures have resulted from a number of radiation incidents. For

example, experiments involving fissile materials were routinely conducted in the ORNL Criticality

Experiments Laboratory (Building 9213). There were at least four inadvertent nuclear excursions in the

laboratory; three of which occurred while working with aqueous solutions of enriched uranium and a

fourth involved a bare uranium metal assembly [McLaughlin et al. 2000]. Only two of these events

resulted in significant radiation exposure. The first involved a cylindrical container which held a UO2F2

solution of approximately 18.3 kg of 93% enriched uranium. The excursion occurred on May 26, 1954 as

a result of a mechanical failure while adding solution to the tank. Exposures to all involved personnel

were below 10 mGy [Thomas and Callihan 1958]. The second excursion also involved cylindrical tank

235 containing a UO2F2 solution of approximately 27 kg of U. An operator error during solution transfer resulted in an inadvertent prompt criticality on February 1, 1956. Resultant exposures ranged from ND to 28.7 mGy [Johnson 1956].

Perhaps the most hazardous work involved the development of several large-scale radiochemical processes for the separation of certain radionuclides (e.g., 239 Pu, 233 U, 140 Ba, 131 I, 137 Cs) from highly radioactive irradiated fuel. Beginning in 1943, ORNL piloted many of the solvent extraction processes adopted by other production facilities for the separation of uranium and plutonium

[Brooksbank et al. 1994]. Much of this work was conducted in the hot cells of the Pilot Plant (Building

3019, formerly Building 205), the By-Product Processing and Chemical Separations Laboratory (Building

3026, formerly 706-C and -D), High-Level Radiochemical Laboratory (Building 4501), and Unit Operations

(Building 4505). ORNL was also responsible for the manufacture of RaLa used in nuclear weapons research at LANL. From 1944 through 1956 ORNL provided over 18,500 TBq of 140 Ba for the Bayo Canyon

implosion experiments following separation in Building 3026. As with LANL Bayo Canyon workers, there

was evidence of overexposures to ORNL RaLa workers, especially during the earliest years of operations,

found in several safety reports.

These radiochemical processes involved the handling of highly radioactive spent fuel capable of

producing dose rates of several Gy∙hr -1. Although the facilities were designed to keep exposures ALARA, unusual operating conditions led to some serious incidents. In 1957, two operators assigned to Building

3019 received 630 mGy and 314 mGy, respectively, after inadvertently entering areas in and around a hot cell containing irradiated thorium slugs. The dose rate just beyond the cell entrance was in excess of

100 Gy∙hr -1 [Winters et al. 1957b]. Remarkably, a similar event occurred in Building 3019 just one month later, resulting in one worker receiving 134 mGy [Winters et al. 1957a]. The exposed worker was a clerk, who, at the request of a supervisor, entered a hot cell to drain a filter. A contributing factor in this

overexposure was the policy to occasionally select non-operators for high-dose work to avoid exposure-

based work restrictions from being placed on operations staff. In this case, the clerk apparently lacked

the skill and training necessary to keep his exposure below established limits.

3.2.5 PNS

PNS, located in Kittery Maine, has a storied past of shipbuilding and repair dating back to

Colonial America. Among its many “firsts”, PNS is the first U.S. Navy shipyard (est. 1800), the first

shipyard to commission a U.S. Navy submarine (1917), the first government-owned and operated

shipyard constructing nuclear powered submarines (1956), and the first Naval shipyard to acquire full

capability for overhaul, repair, and refueling of nuclear-powered submarines (1958) [McDonough 1978].

To date, PNS workers have completed nearly 250 overhauls or other scheduled maintenance activities

on nuclear-powered submarines. This activity is responsible for most occupational radiation exposures at PNS; however, a small group of shipyard workers (i.e., <200) were exposed while performing duties as industrial radiographers, instrument technicians, and medical x-ray technicians [Daniels et al. 2005].

PNS is a large industrial complex that has employed workers from a variety of trades including welding, insulating, pipefitting, painting, machining, and electrical trades. Workers were assigned to specific “shops” and each shop conducted specific tasks. Workers were further characterized by job title.

Individual employment histories were maintained in personnel records that document job and shop assignments throughout a worker’s career. Employment varied with mission; however, during the period from 1950 to 2002 the mean employment was about 7,000 workers and ranged from 3,300 workers in 1998 to 10,500 workers employed in 1952 (Figure 3-7).

5 Number of Numberworkers (thousands)

0 1950 1960 1970 1980 1990 2000

Year

Figure 3-7. PNS Employment (1950-2002)

The first large-scale nuclear work began with the overhaul and refueling of the USS Nautilus on

June 3, 1959. As a direct result, an extensive radiological controls program was established according to

Navy policy at the time, which included provisions for personal monitoring and occupational dose limits

consistent with international and national recommendations [NAVMED 1951; PNS 1980]. The initial standards limited whole-body (deep dose equivalent) exposures to 3.0 mSv-wk -1 not to exceed an accumulated total body dose that equaled 50 times the worker’s age in years over 18 [i.e., 50(age-18) mSv]. Naval policy on exposure limits and PNS administrative controls levels were refined over time until achieving the currently accepted standard of limiting annual whole-body exposures to 50 mSv (Table 3-

3).

Table 3-3. External whole-body dose limits (mSv) and administrative controls at PNS (1958-1978)

Period Dose limits (mSv) Lifetime Yearly Quarterly Monthly

Nearly all occupational doses resulted from work within the shielded reactor compartment onboard nuclear-powered submarines during reactor shut down. The predominant exposure was external, whole-body-penetrating gamma radiation emitted by activation products deposited in reactor systems and components, principally 60 Co (half-life 5.27 y). 60 Co results from neutron activation of 59 Co

and emits two high-energy photons (1.17 MeV and 1.33 MeV) during beta decay transformation to

stable 60 Ni. Neutron exposures, from reactor operations during testing or from small calibration sources,

were infrequent and minimal. Previous estimates indicate that neutron exposures contributed approximately 0.03% to the total shipyard dose [Daniels et al. 2005].

3.2.6 SRS

SRS is located on a 806 km 2 (200,646 acres) tract of land along the Savannah River approximately 25 miles southeast of Augusta, Georgia and 20 miles south of Aiken, South Carolina. The facility was constructed and operated as the Savannah River Plant (SRP) by DuPont until succeeded by the Westinghouse Savannah River Company (WSRC) in April, 1989 who operated the site as SRS through

2005. The site’s primary mission was to produce weapons-grade plutonium ( 239 Pu) and tritium ( 3H).

Large-scale production of these materials continued into the late 1980s to early 1990s, when the site

shifted its focus to clean-up and closure activities following the end of the Cold War.

SRS was originally organized into nine manufacturing areas, a central administration area, and

two “service” building areas. Construction of nuclear facilities began in 1951. Within three years of

groundbreaking, the first of five heavy-water moderated and cooled nuclear reactors [i.e., R (1953-

1964), P (1954-1988), L (1954-1968), K (1964-1988), and C (1955-1985)] and two chemical separations

and processing plants (H-Area and F-Area Canyon Buildings) were operational. Construction began with

the fuel and fabrication area (300 Area), where reactor fuel assemblies and targets were manufactured

for the plutonium producing nuclear reactors. The first uranium slugs were produced October 1952, and

the first fuel rod assemblies were assembled in January 1953. The R reactor went online in December

1953, and the first irradiated fuel was removed from the reactor in June 1954. The first plutonium was chemically extracted in the chemical processing facility, known as the 221-F canyon facility, in November

1954. Tritium production began shortly thereafter, with the first product produced from irradiating lithium-aluminum targets in the 200-F area in October 1955.

As at Hanford, SRS reached its highest production mark in the early 1960s (Figure 3-8). The demand for plutonium and tritium had skyrocketed from 1955 to January 1964, and SRS produced half of the nation’s plutonium and nearly all of the tritium during that time. To meet increasing demand, SRS began its power ascension program (1955-1963), whereby reactor systems were heavily redesigned to support power increases in all five operating reactors to several times higher than the initial rated power. Production rapidly decreased following President Johnson’s State of the Union Address in

January 1964, announcing a reduction in nuclear materials production, which prompted the shutdown of the R Reactor shortly thereafter [Reed et al. 2002].

Reactor operations continued to decline steadily until 1987-1988, when the remaining three operating reactors (K, L, and P) were shut down. During this time, DuPont announced that the company would not continue SRS operations. A new management and operations (M&O) contract was awarded to Westinghouse on April 1, 1989, who promptly changed the name to SRS to set the stage for clean-up operations. Reactor operations had essentially ceased, although K Reactor would be briefly operated in

1992 and then placed in cold standby in 1993.

4000

3000

2000

1000 Combinedreactorpower (Mw-d) 0 1955 1960 1965 1970 1975 1980 1985 1990 Year

Figure 3-8. SRS plutonium production reactor power in Mw-d (1955-1987)

Radioactive materials handling began in the fall of 1951, with the receipt of nuclear materials

(primarily uranium and radium) for research and development purposes in the CMX and TNX facilities.

Once the production reactors were brought online (1953), radiation exposures at the site were typical for reactor operations, whereby accumulated doses were predominately from whole-body low-LET exposures to gamma radiation during operations and maintenance. However, unlike other facilities in the study, there was a significant potential for internal exposure to tritium. Likewise, large-scale operations that focused on all facets of plutonium ( 238 Pu, 239 Pu, and 242 Pu) production were likely to

result in significant internal exposures to some workers. SRS workers were also involved in the large

scale production and separation of other useful radionuclides, such as 60 Co, 210 Po, 147 Pm, and isotopes of

americium, curium, and californium.

2 Number of Numberworkers (thousands) 0 1950 1955 1960 1965 1970 1975 1980 1985 1990 Year

Figure 3-9. SRS employment, excluding subcontractors (1951-1990).

3.3 Previous Studies at Primary Sites

By design, each of the facilities selected for the current study had been considered in one or more previous studies. Looking back at the previous studies provides information on the evolution of methods used in radioepidemiology, which in some cases has elucidated associations between ionizing radiation and leukemia. For example, several studies of leukemia in civilian workers at PNS have been published from 1978 to 2005 [Kubale et al. 2005; Najarian and Colton 1978; Rinsky et al. 1981b; Silver et al. 2004; Stern et al. 1986; Yiin et al. 2005]. Excluding the first PMR study in 1978 [Najarian and Colton

1978], only the most recent study reported a significant elevation in leukemia risk [Kubale et al. 2005].

This finding is testament to the value of continued followup and improvements in exposure assessment and epidemiologic methods that have occurred over the years.

This section provides a brief description of the relevant results from previous studies of workers employed at the primary studies. A noteworthy omission is the large study of leukemia mortality in workers employed at Hanford, LANL, ORNL, PNS, or SRS that forms the foundation of the current study

[Schubauer-Berigan et al. 2007a]. This study is described in detail in a previous section of the report.

3.3.1 DOE Health and Mortality Study Overview

In 1964, the AEC awarded a contract to the University of Pittsburgh to begin a pilot study for a comprehensive epidemiologic examination of its workforce [Fry et al. 1995]. The pilot study was essential for identifying and collecting information necessary for a large study, including the linkage with the AEC and the Social Security Administration (SSA) for vital status and the placement of a moratorium on the destruction of historical records within the AEC. By the early 1970s, the AEC Health and Mortality

Study (HMS) was underway, first as a continuation of work piloted by the University of Pittsburgh (1964-

1977) and then later by other AEC contractor staff (1978-1990). The original concept for the HMS was to

examine complex-wide mortality patterns; thus, data were collected on several study facilities including

Hanford, LANL, ORNL, and SRS. The first published results stemmed from a proportionate mortality

study of Hanford workers that was completed in 1977 [Mancuso et al. 1977].

During the period from 1978 to 1990, HMS studies were designed foremost as “hypothesis-

generating”, thus cause-specific mortality studies with the U.S. population as referent were most

common. Although study cohorts were relatively large (5,000-44,000), few exceeded a followup period

of 25 years or 20% mortality. Few studies considered the effects of other known risk factors (e.g.,

smoking and chemical exposures). Most studies relied on external comparisons using the U.S.

population as baseline, which nearly always showed strong deficits in all-cause mortality because of a

“healthy worker “selection bias. None of the early HMS studies reported a statistically significant

positive association for leukemia, although elevations in leukemia were suggested in early ORNL studies

[Fry et al. 1995].

The year 1990 marked the end of DOE-conducted epidemiologic studies. Based on the

recommendations of the National Academies [NRC 1994] and the SPEERA [DOE 1990], DOE entered into

an agreement with NIOSH to conduct an independent program of analytic epidemiologic research

related to DOE’s activities. Data supporting the HMS studies were turned over to NIOSH shortly thereafter. Thus, a second generation of health studies began with the foundation created by DOE and additional information collected by NIOSH. The NIOSH OERP studies typically involved expanded cohorts that included women and all races. Advances in computing technology and epidemiologic methods had enabled dose-response modeling and improved control of cofactor effects. With these advancements and increased follow-up, positive leukemia associations were more apparent in recent low-dose occupational studies conducted by NIOSH and others [Cardis et al. 1995; Kubale et al. 2005; Muirhead et

al. 2009; Richardson and Wing 2007; Telle-Lamberton et al. 2007; Zablotska et al. 2004].

3.3.2 Hanford

Health effects in Hanford workers have been studied extensively. As the first publication under

the AEC HMS, Mancuso et al. [1977] examined the mortality patterns in 24,939 Hanford workers ever

employed from 1944 to 1971 and found some evidence of elevated myeloid leukemia risk (O/E= 1.9).

However, there were few observed cases ( n=11) and the at-risk experience of survivors was not

considered. Moreover, comparisons were made to the total U.S mortality experienced in 1960 without

any attempt to standardize the age-distribution. The unconventional statistical analyses used by

Mancuso et al. [1977] prompted several critiques and subsequent analyses of the data that largely

refuted their results [Anderson 1978; Gilbert and Marks 1979; Hutchison et al. 1979]. Nevertheless,

Mancuso and colleagues [Kneale et al. 1981; Kneale et al. 1984; Kneale and Stewart 1993; Stewart et al.

1980] continued to publish a series of articles based on these data that suggested elevated cancer

endpoints (primarily multiple myeloma and adenocarcinoma of the pancreas). Detailed information on

the early Hanford studies and the controversy surrounding rival estimates is available elsewhere

[Nussbaum and Kohnlein 1994; Stewart and Kneale 1991].

In the midst of the firestorm following the Mancuso publication, the now newly formed DOE sponsored a reanalysis of the Hanford data. Gilbert et al. [1979] examined mortality in a cohort of white males ever employed at the Hanford Site from 1944 to 1965 ( n=20,842). Their analysis focused on the subcohort of workers employed at least two years ( n=13,075). Of these workers, analyses were limited to those who were monitored for radiation exposure ( n=12,522). Vital status followup was through April

1, 1974. The SMR for leukemia in workers with two or more years of employment was 0.46 (10 observed cases) indicating a strong healthy worker effect. A positive dose-response with increasing dose categories was not observed in trend tests associated with myeloid leukemia ( P=0.59) or other leukemias ( P=0.71). Two addendums to the study immediately followed that updated analysis to include death information though 1977 [Gilbert and Marks 1980] and additional employment information after

1965 [Tolley et al. 1983]. These updates did not result in appreciable changes in leukemia risk estimates.

A subsequent study [Gilbert et al. 1989] added three years of followup to the Hanford cohort, which was expanded to include all workers employed prior to 1978 ( n=44,345). This study showed similar but less depressed leukemia mortality risk compared with the U.S. population (SMR=0.71, n=52) and also lacked a positive association between radiation exposure and leukemia mortality in internal comparisons (ERR∙10 mSv -1= -0.02; 95% CI: -0.02, 0.08). A second followup to this cohort was conducted

by Gilbert et al. [1993], who added mortality information though 1986 (5+ years) but did not include

more recent hires. As before, a strong healthy worker effect was indicated (Leukemia SMR=0.84, n=80)

and a radiation dose-response was not evident for leukemia excluding the CLL subtype (ERR∙10 mSv -1= -

0.01; 95% CI: <0, 0.03).

Wing and Richardson [2005] examined a cohort of 26,389 Hanford workers employed from 1944 to 1978 and followed through 1994. Poisson regression was used to estimate associations between mortality and cumulative external radiation dose adjusted for age, birth cohort, race, sex, socioeconomic status (SES), employment status, in vivo monitoring, and work in plutonium jobs. There

were 2,265 decedents identified with cancer as an underlying or contributory cause. Similar to the results from previous studies on Hanford workers, the estimate of leukemia ERR with a five-year lag was negative (ERR∙Sv -1=-1.16; 90% CI: <0, NC).

3.3.3 INL

Schubauer-Berigan et al. [2005] examined the mortality experience in a cohort ( n=63,561) of

civilian workers employed at INL from 1949 to 1991. The cohort was predominantly white (96%) and

male (81%). Followup was through 1999, indicating that about 17% of the cohort was deceased. There

were 36,442 workers (57.3%) ever monitored for radiation exposure, resulting in a collective dose of

nearly 470 person-Sv and an average whole-body dose of 12.9 mSv. Elevations in leukemia were not

evident in comparisons with the U.S. population using 1940-1999 rates (SMR=0.91; 95% CI: 0.76, 1.09;

120 deaths), although results were imprecise. However, a slight elevation in risk was observed when

compared to the rates (1960-1999) from the combined population of Idaho, Montana, and Wyoming

(SMR=1.07; 95% CI: 0.89, 1.28; 119 deaths).

Schubauer-Berigan et al. [2005] also examined the dose-response for leukemia, leukemia

excluding the CLL subtype, and CLL in the subcohort of radiation-monitored workers with date of birth

(DOB) and date first monitored available (n=36,169). Loglinear and linear Poisson regression models

were developed and stratified on age group, calendar time, duration of employment, and sex. Exposure

lag periods of 5, 7, 10, and 20 years were used to examine the effects of varying lags. Slight increases in

leukemia (ERR∙100 mSv -1= 0.54; 95% CI: -0.09, 2.26; 70 cases) and leukemia excluding CLL (ERR∙100 mSv -

1= 0.54; 95% CI: -0.11, 2.38; 52 cases) were observed in linear ERR models with a seven-year exposure lag (best fit), and control for SES. Adding a quadratic term for dose did not improve model fit. There was

no evidence of increased CLL risk in any model tested.

3.3.4 LANL

As an offshoot of the HMS, LANL was selected to participate in the DOE-sponsored National

Plutonium Workers’ Study, which was intended to examine plutonium-related health effects in workers from several weapons facilities [Acquavella and Wilkinson 1983]. As a component of this research,

Acquavella et al. [1983] examined cancer incidence in LANL and Zia workers who were employed for at least one year from 1969 to 1978. Cancer cases were identified by computer matching of the study roster to the New Mexico Tumor Registry files. Expected cases were calculated from Incidence rates for

New Mexico that were specific for age, sex, ethnicity and calendar period. Statistical analyses were restricted to white non-Hispanics of both sexes, which was approximately 70% of the total enrollment.

The authors did not report study size. An elevation in leukemia incidence was observed in males

(SIR=1.44; 95% CI: 0.39, 3.70; n=4) although estimates were imprecise. No leukemias were observed in

females. A large deficit in all cancers was apparent (SIR=0.60; 95% CI: 0.44, 0.79), primarily as the result

of fewer than expected smoking-related cancers such as lung (SIR=0.01; 95% CI: 0.01, 0.37; n=2) and oral

cancer (SIR=0.15; 95% CI: 0.00, 0.86). This deficit was attributed to the high SES that was evident in the

study group.

An unpublished report by Galke et al. [1992] described the mortality in workers ( n=5,424) employed at the Zia Company from April 1, 1946 to December 31, 1978 and who were exposed to either plutonium or external radiation. Vital status followup was through 1984. Half of the white population of

this cohort was Hispanic. Almost all (96%) of the study subjects ( n=5,200) were monitored for external

radiation, and approximately half ( n=2,741) were monitored for plutonium exposure. The leukemia SMR

in white males ( n=4,942) was elevated, but not significantly (SMR=1.25; 95% CI: 0.70, 2.06; n=15). This

total included seven deaths from subtypes other than myeloid leukemia; however, the nature of these

other leukemias is not described. No leukemia deaths were observed in workers exposed to external

radiation. The leukemia dose-response (over plutonium or external radiation exposures) was not evaluated.

Wiggs et al. [1994] conducted a cohort mortality study of white males ( n=15,727) ever employed at LANL from 1943 to 1977. Vital status followup was through 1990. Among the decedents ( n=3,196), 44 cases were identified with leukemia as the underlying cause of death (SMR=1.01; 95% CI: 0.73, 1.35).

Internal comparisons of radiation monitored workers were carried out using a two-year lag and by stratifying workers over four dose categories: <10, 10-<50, 50-<100, and ≥100 mSv. Dose-response analyses did not reveal a significant or positive trend in leukemia mortality across categories of increasing dose ( P=0.44); however, it is noted that external exposures were highly right-skewed and that

21 of the 26 leukemia cases were in the baseline dose category (i.e., 0-<10 mSv).

A series of studies were conducted that examined the health effects of plutonium exposures in a group of white males ( n=26) exposed at LANL from 1944 to 1946 [Hempelmann et al. 1973; Voelz et al.

1985; Voelz et al. 1979; Voelz and Lawrence 1991; Voelz et al. 1997; Voelz et al. 1983]. Effective doses to these individuals from plutonium deposition ranged from 0.1 to 7.2 Sv. The group was medically examined at five-year intervals beginning in 1952. The last study included 50 years of followup; vital status was complete through 1994 [Voelz et al. 1997]. There were 19 subjects (73%) alive at the end of the follow-up period. Of these, only four had a history of cancer and none presented with leukemia. Of seven decedents, three had a malignancy (lung, prostate, and bone) listed as the underlying cause of death. A strong deficit in all-cause mortality was evident when compared to the U.S. population

(SMR=0.43; 95% CI: 0.18, 0.88; n=7). There was no evidence of hematopoietic disease in any of the subjects examined.

3.3.5 ORNL

Frome et al. [1990] examined mortality in white males ( n=28,008) employed at the Clinton

Engineer Works [i.e., K25, Y12 (TEC), and X10 (ORNL) facilities] for at least 30 days during the World War

II era (1943-1947) and at no time thereafter. Of these workers, only 1,158 (4.1%) were solely ORNL

employees. Vital status follow-up was conducted from January 1, 1950 through December 31, 1979.

There were 11,671 deaths (42%) by the end of study. External comparisons were made using age-

adjusted mortality rates in the U.S. population as baseline. Dose-response analyses were conducted

using Poisson regression. Radiation exposure was treated as a binary variable (ever/never). There was

no evidence of a healthy worker effect (all causes SMR=1.11, P<0.01) and cancer mortality was slightly,

although not statistically significantly, elevated (SMR=1.05, P>0.01; n=2,207) compared to the U.S. population. The increase in mortality was primarily due to lung cancer and non-malignant diseases of the respiratory system; however, leukemia mortality in these workers was also elevated (SMR=1.13,

P>0.01; n=92), although the estimate was not statically significant. A radiation dose-response was not

apparent for any outcome.

Wing et al. [1991] studied white male ORNL workers ( n=8,318) hired from 1943 to 1972 and not

employed in other DOE nuclear facilities. Vital status follow-up was though 1984. Cause-specific

mortality in ORNL workers was compared to the U.S. population. Dose-response analyses were

performed using Poisson regression models adjusted for age, cohort, and pay code. The median

cumulative dose was 1.4 mSv. A deficit in mortally compared to the U.S. population was evident (SMR=

0.74; 95% CI: 0.71-0.78; n=1,524). However, leukemia mortality was significantly elevated in these workers (SMR=1.63; 95% CI: 1.06, 2.35; n=28) and in workers monitored for internal contamination

(n=3,763) who, on average, had higher doses (SMR=2.23; 95% CI: 1.27, 3.62; n=16). A dose-response under a 20-year exposure lag was apparent for all cancers (4.9% per 10 mSv, P=0.001). The leukemia dose-response (20-year lag) was positive but not significant (9.15% per 10 mSv, P=0.44).

Frome et al. [1997] examined mortality patterns of workers ( n=106,020) employed at least 30 days from 1943 to 1985 at one or more of the Clinton Engineer Works facilities. Comparisons with the

U.S. population did not reveal elevated leukemia risks (e.g., SMR for white males = 0.980); however, significant differences in leukemia risk were observed by facility, whereby the risk was highest in ORNL workers and lowest in Y-12 workers. The dose-response was examined in a subcohort of white males

(n=28,347; 50 cases) from Y-12 and ORNL. The ERR was calculated using Poisson regression under a 2-

year exposure lag and adjusted for age, birth cohort, facility, and SES. They observed a negative,

although imprecise, dose-response (ERR∙10 mSv -1 =<0; 95% CI: <0, 6.5) for leukemia mortality in these workers.

3.3.6 PNS

PNS leukemia mortality patterns have been studied extensively, beginning with a proportionate mortality ratio (PMR) study in 1978 [Najarian and Colton 1978] and ending with the inclusion of PNS workers in the recent multisite nested case-control study [Schubauer-Berigan et al. 2007a; Schubauer-

Berigan et al. 2007b]. In the first of these studies, Najarian and Colton [1978] reviewed deaths certificates for New Hampshire, Maine, and Massachusetts and collected information on 1,722 PNS workers who died from between the years 1959-1977. Telephone interviews with the next-of-kin were conducted to obtain additional information (including radiation worker status) on 525 workers who died prior to age 80. Among these deaths, eight leukemia deaths were identified and six of these deaths were classified as nuclear workers. Expected leukemia deaths were 1.1 (O/E=5.62) and 2.8 (O/E=0.71) for nuclear and non-nuclear worker groups, respectively. There were several important limitations of this study. First, because it is a PMR study, it does not include complete information on the population at risk. Spurious elevation in cancer PMRs could occur from lower non-cancer mortality resulting from a strong healthy worker effect. Second, information on employment (ever employed at PNS) and exposure

(ever/never) was sparse and available for only 30% of the total deaths; therefore, a selection bias may

have been introduced because only a subset of the deaths could be examined. Lastly, to complete ascertainment of the cause of death and obtain exposure information, researchers interviewed the next- of-kin. These interviews may have introduced a significant bias as a result of the interviewer (interviewer bias) or the respondent (recall bias). A subsequent reanalysis of the data confirmed that the misclassification of occupational exposure obtained from the interviews and, to a lesser degree, the healthy worker effect contributed to the elevated leukemia PMR [Greenberg et al. 1985].

Shortly following the study by Najarian and Colton [1978], Rinsky et al. [1981b] reported on a cohort mortality study of white male civilian workers ( n=24,545) ever employed at PNS from January 1,

1952 to August 15, 1977. Vital status was obtained from death certificates and coded by a nosologist according to the ICD revision in place at the time of death. Cases were identified with leukemia as the

UCOD. Main analyses used the NIOSH life-table to generate the expected numbers of deaths based on

U.S death rates in five-year age groups and five-year calendar time intervals. Intra-cohort comparisons

were also conducted using the subcohort of non-radiation workers ( n=15,585) as the baseline group (10-

year age and 13-year calendar time groups) applied to person-years at risk in the radiation exposed

group ( n=7,615). Radiation exposure was determined from annual personnel dosimetry records. The

overall leukemia SMR was 0.94 (95% CI: 0.67, 1.28; n=39). The SMR in the radiation exposed subcohort

was 0.84 (95% CI: 0.34, 1.74; n=7). A significant dose-response trend in leukemia mortality was not

observed. Although sufficiently powered to detect a five-fold increase in leukemia risk as previously

reported by Najarian and Colton [1978], the study had limited power to detect an expected weak

carcinogenic effect from low-dose ionizing radiation. Moreover, the use of the non-exposed cohort as

the baseline group in dose-response analyses may have suppressed risks because of a strong healthy

worker effect that is observed in the radiation exposed cohort.

Using the base cohort described by Rinsky et al. [1981b], Stern et al. [1986] conducted a case-

control study of leukemia mortality in PNS workers. Follow-up was extended though 1980. Cases ( n=53)

were defined as decedents identified with leukemia as an underlying or contributing cause of death.

Controls were selected from the base cohort (excluding leukemia mortality cases) for each case matched on attained age of the case. Potential controls were “scored” based on the sum of the absolute difference in birth dates, date first employed, and employment duration. For each case, the four non- cases with the lowest score were selected as controls. In addition to annual radiation doses, work histories were collected to identify persons potentially exposed to organic solvents (e.g., benzene).

Conditional logistic regression did not reveal a statistically significant association between ionizing radiation at 10 mSv and leukemia mortality (OR=1.26; 95% CI: 0.95, 1.66). However, years as an electrician presented with significant elevations in leukemia risk (OR=1.67; 95% CI: 1.01, 2.78), as did years as a welder (OR=2.86; 95% CI: 1.02, 8.04).

Silver et al. [2004] conducted a mortality study of PNS workers in an expanded cohort that included the previous cohort members and all races and gender of workers employed prior to 1993

(n=37,853). Vital status follow-up was extended through 1996. There were 13,468 cohort members monitored for radiation, of which 11,791 had recorded cumulative doses greater than zero (exposed radiation workers group). External and internal analyses were conducted by NIOSH lifetable analysis.

The SMR for leukemia in the full cohort was 1.01 (95% CI: 0.84, 1.22; n= 115). Using the non-monitored group as baseline, the SRR for unexposed radiation workers (i.e., monitored workers with zero cumulative dose) was 0.59 (95% CI: 0.24, 1.46; n= 5). The SRR for exposed workers was 0.81 (95% CI:

0.47, 1.38; n = 29). However, leukemia SRRs increased monotonically with increasing dose in dose- response analyses using dose categories of 0–<1, 1–<10, 10–<50, and ≥50 mSv and a two-year lag

(P<0.01). At doses in excess of 50 mSv, the SRR was 5.13 (95% CI: 1.37, 19.19; n=7).

Yiin et al. [2005] fit a linear excess relative risk model to the data from Silver et al. [2004], using

Poisson regression to further examine the positive association between leukemia mortality and ionizing radiation. The modeling allowed for the control of potential confounders that were unaccounted for in

previous analyses. In particular, adjustments were made for solvent exposure and lifestyle factors (e.g., smoking) as measured by SES. For leukemia analyses, the best model fit was achieved using a three-year lag. A positive but nonsignificant dose-response was observed, resulting in an estimated excess relative

risk per 10 mSv dose of 0.11 (95% CI: -0.01, 0.40). Neither solvent exposure nor SES was found to a have

significant effect on the radiation dose-response; however, solvent exposure (ever/never) was

positively, though not significantly, associated with leukemia mortality (RR= 1.19; 95% CI: 0.60, 2.35).

Kubale et al. [2005] conducted a nested case-control study of leukemia mortality and ionizing

radiation exposure using the previously described cohort (n=37,853) of workers ever employed from

1952 to 1992. Four controls were matched to each leukemia case ( n=115) on attained age using

incidence density sampling methods. Of 575 cases and controls, only 201 ( n=34 cases) were radiation

monitored. Dose-response analyses were conducted using conditional logistic regression models that

were adjusted for sex, radiation worker status, and solvent exposure duration. The smaller number of

study participants in the case-control design enabled marked improvement in exposure estimates by

allowing researchers to conduct a more individualized and comprehensive exposure assessment for

both radiation and solvent exposures. A statistically significant positive association was found between

leukemia mortality and external ionizing whole-body radiation exposure, after adjusting for radiation

worker status, sex and solvent exposure (linear ERR model; ERR∙10 mSv -1=0.23; 95% CI: 0.03, 0.89).

Subsequent analyses of the full cohort (unpublished) suggested that the skewness of the exposure

distribution, the relatively small number of cases, and the selection of a 4:1 control-to-case ratio may

have worked in combination to inflate the risk estimate in this study; given that increasing the number

of controls resulted in a reduction in the point estimate. It was also noted that trimming extreme

exposures tended to increase the point estimate, suggesting that outlying doses were influential.

3.3.7 SRS

Because of the potential for exposures to a number of occupational carcinogens, the medical department of the DuPont Company established a registry as a means to measure cancer incidence and mortality in the company workforce. The mortality surveillance began in 1957 and included deaths among active and retired employees that were ascertained from life insurance claims. Information on

SRS workers was included in the registries through March 1989 [Crase and Singh 1998]. There were a number of reports stemming from the registry, including two published articles summarizing the overall cancer burden in DuPont employees. However, there were no publications available in the open literature that provided separate information on SRS cancer risks. Of the two published comprehensive surveillance reports, the first described mortality and morbidity patterns within the company ( ∼108,000 workers) from 1956 to 1974 [Pell et al. 1978]. A deficient in leukemia mortality (1957-1974) was observed in males (SMR=0.86; 130 cases) that was statistically significant at the 0.05 level. The female

SMR was slightly elevated (SMR=1.16; 16 cases) although not statistically significant. An update to the study by Pell et al. [1978] was published in 1989 that extended followup through 1984 [O'Berg et al.

1987]. There were about 101,000 workers employed at the end of the followup period. That study did not report significantly (p<0.05) elevated leukemia mortality in males (SMR=0.98; 267 cases) or females

(SMR=1.04; 28 cases).

Cragle et al. [1988] examined mortality in a cohort of white males ( n=9,860) employed at SRS for at least 90 days from 1952 to 1974. The cohort was followed through 1980. Duration of employment was used as an exposure surrogate. SMRs were calculated for subcohorts stratified by SES (hourly, salaried, or combined) and employment period (pre-1955, post-1955). Using the U.S. population rates as referent, the leukemia SMR for hourly workers was 1.63 and the SMR for salaried workers was 1.05. The

SMRs for pre-1955 hourly workers by employment duration were: 1.24 (<5 years), 2.75 (5-15 years), and

1.57 (>15 years). The leukemia SMR of workers employed 5-15 years (SMR=2.75) was significant at the

5% level.

Cragle et al. [1999] updated the study using the cohort information from the previous publication and extended followup through 1986. Radiation exposure information was obtained for

9,757 cohort members (99%) monitored in the period under observation. Leukemia dose-response was examined using trend tests and Poisson regression under a two-year exposure lag. The Poisson model assumed that the excess risk was linear in dose. Cumulative external whole-body dose was stratified into nine categories. Internal comparisons used the unexposed population as baseline. Significant deficits were observed in all-causes (salaried SMR= 0.60; 95% CI: 0.54-0.67 and hourly SMR= 0.85; 95% CI: 0.80,

0.90) and all-cancers (salaried SMR= 0.71; 95% CI: 0.58-0.87 and hourly SMR= 0.86; 95% CI: 0.76, 0.96). A total of 25 leukemia deaths were observed, six in salaried workers (SMR=1.10, 95% CI=0.40-2.40) and 19 in hourly workers (SMR=1.34, 95% CI=0.80-2.09). The trend was positive and increasing with dose for leukemia (excluding CLL) mortality (underlying and nonunderlying causes; P<0.05). The ERR∙100 mSv -1 was 1.36 (90% CI: 0.06, 5.62).

Wartenberg et al. [2001] examined the mortality experience of white and African-American SRS workers (number not reported) employed from 1952 to 1989 and followed through 1995. Dose- response analyses were not conducted. Based on comparisons with the U.S. population, the leukemia risk in white males was elevated but not significantly (SMR=1.17; 95% CI: 0.81, 1.58; n=35), as was the risk observed in African American males (SMR=1.81; 95% CI: 0.57, 3.75; n=5). No information of leukemia subtypes was provided.

Richardson and Wing [2007] examined leukemia mortality in a cohort of SRS workers ( n=18,883) hired between 1960 and 1986, employed at SRS for at least 90 days, and never worked at another DOE facility. Vital status followup was through 2002 and included information on decedents with leukemia

(n=84) as the underlying ( n=73) or contributing ( n=11) cause of death. Cumulative whole-body radiation dose from external sources and from tritium was the exposure of interest. Exposure data were abstracted from facility dosimetry records (i.e., personal dosimeter and tritium bioassay records) or imputed from employment histories through 1999 [Richardson et al. 2007]. Dose-response trends were examined using conditional logistic regression assuming a radiation effect that is linear in dose. Models examined leukemia, leukemia excluding chronic lymphocytic leukemia, and myeloid leukemia after adjusting for age, sex, race, birth year, SES, and employment status. Cumulative dose was examined using a fixed three-year lag and in time windows. Richardson and Wing [2007] found a positive, although nonsignificant, association between leukemia mortality and radiation dose under a three-year lag assumption (ERR∙100 mSv -1= 0.41; 90% CI: -0.01, 1.16; n=84). The association between cumulative dose

and leukemia excluding the CLL subtype was positive and significant (ERR∙100 mSv -1= 0.77; 90% CI: 0.14,

1.98; n=62), as was the observed risk of myeloid leukemia (ERR∙100 mSv -1= 1.23; 90% CI: 0.21, 3.54;

n=40). Exposures accrued in the time window 3-<15 years prior were most strongly related to leukemia

mortality. Curvelinearity in dose-response was not evident; an additive model provided a better fit to

the data compared to other models tested.

3.3.8 Pooled studies

A large pooled study was conceived as a component of the AEC Health and Mortality Study that

examined risk in nuclear workers with higher cumulative dose. Making use of data from previous nuclear

worker studies, Fry et al. [1996] examined mortality in white male workers who were ever employed

from 1943 to 1978 at DOE facilities or U.S Navy’s Nuclear Reactor Propulsion Program (NRPP) shipyard

facilities, and who had at least one annual dose that exceeded 50 mSv. The study enrolled 3,145 current

and former workers from 32 contractor facilities ( n=2,035) and seven nuclear shipyards (1,110). Vital

status followup was through 1984 (86.6%). Five subcohorts were evaluated, the largest of which was

white males. Dose-response was not evaluated. Compared with U. S. death rates, the SMR for leukemia

in white males ( n=2,385) was below expectation (0.47; 95% CI: 0.05, 1.68, n=2). A statistically significant deficit in all causes of death in white males was also apparent (0.88; 95% CI: 0.80, 0.97, n=448).

Although followup was best in the group of white males, we note that 13% had unknown vital status at the time of study.

Gilbert et al. [1989] examined mortality in a pooled cohort comprised of white male workers who were radiation-monitored and were employed at least six months at either the Hanford Site, the

Rocky Flats Nuclear Weapons Plant (RFP) in Denver, Colorado, or ORNL. Vital status followup was complete though 1944-1981 for the Hanford subcohort ( n=23,704) and 1943-1977 for ORNL ( n=6,332).

As expected, Hanford exhibited a deficit in leukemia mortality (SMR=0.65; 95% CI: 0.5, 1.0) compared to the U.S. population that was similar to results reported in previous cohort studies [Gilbert et al. 1989].

However, the ORNL leukemia SMR was increased (SMR=1.74; 95% CI: 1.0, 2.9). A negative, although non-significant, dose-response was observed for both facility subcohorts (i.e., Hanford ERR∙10 mSv -1 =

<0; 95% CI: <0, 0.05 and ORNL ERR∙10 mSv -1= <0; 95% CI: <0, 0.14). The combined cohort analysis was updated, which extended followup five additional years for Hanford and seven additional years for

ORNL. An appreciable change in the leukemia (excluding CLL) mortality dose-response was not apparent for Hanford (ERR∙Sv -1= -1.1; 90% CI: <0, 1.9) or ORNL (ERR∙Sv -1= <0; 90% CI: <0, 7). The leukemia ERR∙Sv -1 for Hanford and ORNL combined was -1.3 (90% CI: <0, 1.1).

Matanoski et al. [2008] conducted a pooled study of civilian workers employed at least one year at U. S. Navy nuclear shipyards including PNS. A stratified random sample of male workers ( n=72,357) was drawn from 77,509 nuclear workers and 117,718 non-nuclear workers employed at one of eight

U.S. nuclear naval shipyards through 1980. Vital status followup was through 1981. The sample was comprised of: all 28,542 workers exposed to ≥5.0 mSv, a 25% sample of nuclear workers exposed to

<5.0 mSv ( n=10,462), and a 30% sample of non-nuclear workers (n= 33,353). Average follow-up was 12.8

years for the higher-dose workers, 13.5 years for the lower-dose workers, and 13.1 years for the non-

nuclear workers. External comparison used the U.S. population as reference. Internal comparisons of exposed workers examined leukemia risk in strata of 0-<5.0 mSv, 10.0-<50 mSv, and 50 + mSv, using the

5.0-<10 mSv group as baseline. Dose-response analyses examined zero- and two-year exposure lags for leukemia. The SMR for leukemia in workers exposed to ≥5.0 mSv was 0.91 (95% CI: 0.56, 1.39; n=21),

0.42 (95% CI: 0.11, 1.07; n=4) in those exposed to <5.0 mSv, and 0.97 (95% CI: 0.65, 1.39; n=29) in non- nuclear workers. Leukemia risk in workers with two-year lagged cumulative exposures of 50 mSv or more was elevated (RR=2.41; 95% CI: 0.5, 23.8) compared to the baseline category (0.5-4.9 mSv).

Myeloid leukemia comprised 60% of all cases identified. The mortality experience stratified by shipyard was not reported; however, it was noted that information on eight (15%) PNS leukemia cases were used in this analysis. Among PNS cases, five (63%) had cumulative doses of at least 5.0 mSv.

Cardis et al. [1995] pooled nuclear workers from the U.S., the UK, and Canada to examine mortality (hereafter referred to as the Three-Country Study). The study cohort was comprised of 95,673 workers who were ever monitored for radiation exposure and were employed for at least six months in one of the three countries. The U. S. group (n=45,824) was assembled from previous studies of workers employed at Hanford ( n=32,595), RFP (n=6,638), and ORNL ( n=6,591). Vital status follow-up was through: 1986 for Hanford, 1979 for RFP and 1984 for ORNL. Dose-response estimates were obtained by

Poisson regression assuming leukemia mortality was linear in dose. Dose-response models were adjusted for attained age, sex, calendar period, and facility under a two-year exposure lag. Cumulative whole-body dose from external sources of high and low-LET radiation was the primary exposure variable. The ERR was positive for leukemia (ERR∙Sv-1= 1.55; 90% CI: -0.21, 4.7; n=146) and positive and significant for leukemia excluding CLL (ERR∙Sv -1= 2.18; 90% CI: 0.13, 5.7; n=119). The strongest association was observed for the CML subtype (ERR∙Sv -1= 11.00; 90% CI: 2.9, 30.9; n=28). Heterogeneity

tests revealed moderate differences in leukemia risk (excluding CLL) between sites ( P=0.08); however, results were heavily weighted by the Sellafield (ERR∙Sv -1= 43.5; 90% CI: 3.1, <100; n=10) and Canadian

subpopulations (ERR∙Sv -1= 48.40; 90% CI: 2.8, <100; n=10). In contrast, the ERR values were negative for

Hanford (ERR∙Sv -1= -0.90; 90% CI: <0, 2.9; n=47) and ORNL (ERR∙Sv -1= -1.06; 90% CI: <0, 4.8; n=18), and positive for RFP (ERR∙Sv -1= 4.08; 90% CI: <0, 54.2; n=4).

In a followup to the Three-Country study, Cardis et al. [2007] examined mortality in nuclear workers ( n=407,391) from 15 countries (hereafter referred to as the 15-Country Study). U.S. cohorts included Hanford, ORNL, INL, and a cohort of commercial power workers previously studied by Howe et al. [2004]. Unlike the previous study, the point estimate for leukemia risk was not statistically different from zero (ERR∙Sv -1= 1.93; (90% CI: <0, 7.14). Subtype analyses did not reveal associations between

radiation dose and mortality from CLL, ALL, or AML, but a borderline association was evident for CML.

Heterogeneity in risk by facility was not evident (P<0.1) and, except for Japan and ORNL, all subcohorts

with 10 or more leukemia deaths showed leukemia ERRs above the null. Nevertheless, only one country

(France) had a statistically significant elevated relative leukemia risk (ERR∙Sv -1=242; 90% CI 42.2, 1510; n=11 deaths). A key difference in this study compared to the Three-Country Study is that some workers

were excluded if they presented with significant exposures to high-LET radiation. This resulted in the

exclusion of a number of leukemia cases who had relatively high doses from the 15-Country Study.

Although the number of leukemia deaths in the 15-Country Study was nearly twice the number from the

previous 3-Country Study, most of the additional leukemia deaths occurred in subjects with low doses.

4.0 Study Roster

4.1 Case and Control Selection

All cases were identified using vital status information from previous studies ( n=257 cases) or extended follow-up. ICD codes from Revisions 6 through 8 were translated into ICD Revision 9 for ease of presentation. Cases were defined as decedents with leukemia [ICD-9: 204-208; ICD-10 C91-C95, excluding leukemic reticuloendotheliosis, i.e., “hairy cell” leukemia (C91.4)], identified as the UCOD.

Distinctions were made between leukemia, leukemia excluding the CLL subtype, and by major subtypes.

Non-CLL cases were comprised of acute, other non-chronic, myeloid, or monocytic leukemias (Table 4-

1). Indeterminate cases, defined as lacking sufficient information on cell type and severity (i.e., ICD-9 codes: 204.9, 208.1 208.9; ICD-10 codes: C91.9, C95.9, C95.1) to precisely determine non-CLL status, were treated as non-CLL cases in main analyses. Analyses of the four main subtypes (i.e., AML, CML, ALL, and CLL) included only cases with complete information on cell type and severity. For subtype analysis,

“myeloid” leukemias include monocytic cell type. Information from patient medical records and

EEOICPA claim files were used, to the extent practical, to reconcile indeterminate cases.

Table 4-1. Leukemia subtype classification

Severity Cell Type Chronic Acute Other Unknown Lymphoid 204.1, C91.1 204.0, C91.0 204.2, 204.8, C91.2, 204.9, C91.9 C91.3, C91.5, C91.7 Myeloid 205.1, C92.1 205.0, 205.3, C92.0, 205.2, 205.8, C92.2, 205.9, C92.9 C92.3, C92.4, C92.5 C92.7 Monocytic 206.1, C93.1 206.0, C93.0 206.2, 206.8, C93.2, 206.9, C93.9 C93.7 Other 207.1, C94.1 207.0, C94.0, C94.2, 207.2, 207.8, C94.3, 207.9, C94.9 94.4, 94.5 C94.7 Unknown 208.1, C95.1 208.0, C95.0 208.2, 208.8, C95.2, 95.7 208.9, C95.9 Codes 204-208 are from ICD-9 and codes beginning with “C” are from ICD-10.

Other demographic information, such as gender, race, and ethnicity, was abstracted from available medical and personnel records. Information on race and ethnicity was sparse. If race was unknown, the subject was assumed to be White. For those individuals without documented ethnicity,

Hispanic ethnicity was determined by matching surnames to Spanish surnames listed by Word and

Perkins [1996].

Four controls were selected for each case using incidence density sampling [Beaumont et al.

1989]. Controls were matched to their respective case on attained age and by primary study group, whereby Group 1 risk sets were formed from all facilities previously studied in the LCCS [Schubauer-

Berigan et al. 2007a] and Group 2 risk sets contained INL workers [Schubauer-Berigan et al. 2005]. To the extent that was practical, risk sets were retained from the previous study by Schubauer-Berigan et al. [2007a] by filling in behind controls which did not meet the study criteria (e.g., controls hired before

1950).

The number of matched controls was arbitrary, but based largely on the designs of similar studies. In case-control settings, the sample size is a balance between statistical efficiency and feasibility in exposure assessment; however, a number of early works [Gail et al. 1976; Ury 1975; Wacholder et al.

1992] have led many to assume that there is little profit in samples sizes greater than four controls to each case [Breslow et al. 1983]. However, this dictum is set in the rare case of testing the null hypothesis of one dichotomous exposure variable, whereby study efficiency is calculable for M controls per case by

M/(M+1). Recent works have shown that efficiency increases with the number of cases, as the true exposure-response decreases, and with decreasing skewness of the exposure distribution [Bertke 2011;

Breslow et al. 1983; Pang 1999]. Furthermore, there is evidence of a bias away from the null (when compared to the full cohort) that is inversely proportional to the number of matched controls selected to studies involving highly skewed exposure data, although this bias appears negligible in studies of 300

or more cases, match control ratios of 4:1 or greater, and small effect sizes [Bertke 2011], which are

characteristics of the current study.

4.2 Demographics

There were 369 leukemia cases identified in the base cohort; all of which were deemed suitable for study inclusion. Initial followup provided sufficient information on cell type and severity for 271 cases. The subtypes for 20 cases initially lacking subtype classification were ascertained using information from medical records, reducing the number of cases with incomplete subtype information to 69 (18.6%). There were 291 cases identifiable to the four subtypes of interest (Table 4-2). The average age at death was 67 years and the attained age among cases ranged from 26 to 93 years. The majority of leukemia deaths (90.5%) occurred in males ( n=334). There were 201 acute leukemia deaths (54.4%)

which were comprised mostly of AML ( n=147). There were only 18 ALL cases. Among chronic leukemias

(n=127) there were 52 identified as CML and 74 CLL cases. There were 31 indeterminate cases that were

potentially CLL (Table 4 -2). Leukemia excluding CLL and ambiguous subtypes comprised nearly 72% of

the cases ( n=264).

Among decedents with known leukemia subtype, CLL cases had the greatest average attained

age at 70 years, followed by AML ( 68 years), CML (63 years), and ALL (61 years). The average attained

age of the indeterminate cases was 72 years. About 84.8% of the leukemia deaths occurred after 1979

(n=313) and 63% ( n=234) from 1990 through 2005 (Figure 4-1). About 44% of decedents were 70 years

of age or older and most of these deaths occurred in 1995 or later (Table 4-4). Those who died at age

<50 years (n=31) tended to have been born later; there were 19 deaths (13.5%) among those who were

born after 1932 and died at age 49 years or younger (Table 4-5). As expected, birth date was a correlate

of the date at death (Pearson coefficient=0.41, P<0.001).

The median employment duration was elevated for cases compared to the base cohort for each site and for all sites combined. PNS cases had the highest median employment duration (25.3 years) and the earliest median year of first hire (i.e., 1951), while INL cases were later hires (median value of 1958) and were employed for less time, with a median employment duration of 7.7 years (Table 4-3).

Table 4-2. Distribution of leukemia cases by subtype.

Chronic Acute Other Unknown Total Lymphoid 74 18 1 9 10 2 Myeloid 52 136 3 1 19 2 Monocytic 0 11 0 1 12 Other 0 1 4 0 5 Unknown 1 35 1 21 58 Total 12 7 201 9 32 369 Chronic Lymphocytic (CLL) n=7 4 Non -CLL n=26 4 Indeterminate n=31

Table 4-3. Case distribution by site.

Study Facilities All Hanford INL LANL/ Zia PNS SRS X10 mean age at death (y) 67 66 67 66 71 65 67 mean year of birth 1925 1927 1925 1926 1919 1924 1927 median year of first hire 1956 1958 1958 1957 1951 1953 195 9 median employment (y) 15.4 9.4 7.7 10.4 25.3 18.7 9. 0 cases 369 58 90 46 48 78 49 non -CLL 26 4 47 64 36 34 54 29 CLL 74 8 16 7 7 19 17 indeterminate 31 3 10 3 7 5 3

Table 4-4. Distribution of the age at death by the year of death.

Year of Death Age at death <50 50 -<60 60 -<70 70+ Row Total <1965 2 0 1 0 3 (0.8) 1965 -<1975 8 13 8 3 32 (8.7) 1975 -<1985 12 14 14 11 51 (13.8) 1985 -<1995 7 16 42 39 104 (28.2) 1995+ 2 19 48 110 179 (48.5)

Total (%) 31 (8.4) 62 (16.8) 113 (30.6) 163 (44.2) 369

Table 4-5. Distribution of the age at death by birth cohort quartiles.

Year of Birth Age at death <50 50 -<60 60 -<70 70+ Row Total <19 19 1 13 18 63 95 (25.8) 19 19 -<19 25 3 10 28 59 100 (27.1) 19 25 -<19 33 8 8 35 36 87 (23.6) 19 33 + 19 31 32 5 87 (23.6)

Total (%) 31 (8.4) 62 (16.8) 113 (30.6) 163 (44.2) 369

100

140 160

140 120

120 100

100 80

Frequency 60 Frequency 60

40 40

20 20

0 0 1950s 1960s 1970s 1980s 1990s 2000s <1900s 1900s 1910s 1920s 1930s ≥1940s Panel A: death year Panel B: birth cohort

Figure 4-1. Histograms of the distribution of leukemia deaths (n=369) by decade (Panel A) and by birth cohort.

Of 1,476 controls, there were 28 workers (8 cases and 20 controls) who were selected twice and

one worker who was selected three times; therefore, the case control study group ( n=1,845) is

comprised of 1,816 individual workers (i.e., 1,447 non-cases). Most subjects were male (87.6%) and

White (85.3%). Complete information on gender was obtained; however, 251 subjects were missing

information on race. Of these, 31 workers were most likely to be Hispanic based on surname. The

remaining 220 subjects were assumed to be White and non-Hispanic in subsequent statistical analyses

(Table 4-6).

101

Table 4-6. Demographic information on cases and controls

Variable Case Control Total Cases and controls N 369 14 76 18 45 Sex (%) Male 33 3 (90. 2) 1,28 3 (86.9) 1,616 (87.6) Female 36 (9. 8) 19 3 (13.11) 229 (12.40) Race /ethnicity (%) White non -Hispanic 301 (81.6) 1249 (84.6) 1,5 50 (8 4.1) Hispanic 12 ( 3.3) 19 (1.3) 31 (1.7) Other 13 (3.5) 31 (2.1) 44 (2. 4) Unknown 43 (11.7) 177 (12.0) 22 0 (11.9 ) Birth year (%) ≤ 191 9 11 0 (29.8 ) 56 5 (38. 3) 666 (36. 1) 19 20 -≤192 7 124 (33. 6) 43 7 (29.6) 572 (3 1.0) >192 7 135 (36. 6) 474 (32. 1) 60 7 (3 2.9) Hire year (%) ≤ 1952 117 (31. 7) 549 (37.2) 668 (36.1) 19 53 -≤ 1959 131 (35. 5) 44 1 (29.9) 573 (31.0) >19 59 12 1 (3 2.8) 48 6 (32.9) 609 (32.9) Facility (%) HAN 55 (14.9) 349 (23.6) 385 (20.9) INL 91 (24. 7) 37 1 (25. 1) 287 (15.6) LAN 48 (13.0) 116 (7. 9) 212 (11.5) X10 49 (13. 3) 165 (11. 2) 678 (36.8) PNS 48 (13.0) 243 (16. 5) 87 (4.7) SRS 78 (21.1) 232 (15.7) 196 (10.6)

4.3 Occupation and Social Class

The methods used to make SES assignments are described elsewhere [Schubauer-Berigan et al.

2007a; Schubauer-Berigan et al. 2005]. In general, the collapsed Job title at first hire for each study subject was related to an occupational code listed in the 1980 U.S. Bureau of Census. The first job title

was chosen because it is thought that initial employment provides a better marker for some lifestyle

factors (e.g., smoking) that are typically established in adolescence and early adulthood [Schubauer-

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Berigan et al. 2007a; Schubauer-Berigan et al. 2005; Yiin et al. 2005]. Each of these codes was then assigned to one of six SES categories, namely (from high to low SES): professional, intermediate, skilled non-manual, skilled manual, partly skilled and unskilled.

Work history information was sufficient to make SES assignments for 1,815 study subjects selected as cases and/or controls. Job title information was not available for one worker who was first employed at INL in 1953. It appears that this individual was assigned to the ICPP over most of his career at INL, which spanned approximately five years. This worker was placed into the skilled manual category, given that most workers assigned to the ICPP in the 1950s were placed in this category (Table

4-7).

Table 4-7. Socioeconomic status assignments among cases and controls

Number Assigned (column %) Applicable 1980 Bureau of Census SES Category codes Total Cases Controls Professional 003 -005, 007 -008, 014 -015, 023, 043 - 386 (20.9) 81 (22. 0) 30 4 (20. 6) 057, 059, 064-089, 096, 113-154, 165-169, 178-179 Intermediate 006, 009 -013, 016 -019, 024 -037, 058, 288 (15.6) 50 ( 13. 6) 23 7 (16.1) 063, 095, 097-106, 155-164, 173-177, 183-206, 208-243, 489, 833 Skilled non - 253 -389 212 (11.5) 36 (9. 8) 176 (11.9) manual Skilled manual 207,403 -404, 413 -414, 416 -423, 473 - 680 (36.8) 140 (37. 9) 538 (36. 5) 476, 497, 503-679, 684-699, 824, 843, 849-855 Partly skilled 405 -406, 415, 424 -436, 444 -447, 456 - 88 (4.8) 21 (5.7) 66 (4.5) 458, 463, 468, 486-488, 683, 703- 823, 825-829, 834, 844-848, 856-863 Unskilled 407, 437 -443, 448 -455, 459, 464 -467, 196 (10.6) 41 (11.1) 155 (10.5) 469, 477-485, 494-496, 498-499, 864- 889

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It is important to note that SES is a complex term that is best described by a number of interrelated predictors such as education [Elo and Preston 1996; Schrijvers et al. 1999], income [Ecob and Smith 1999; Rahkonen et al. 2000], and occupation [Gregorio et al. 1997; Kunst et al. 1998]. In our study, information on education and income were incomplete for most study participants; therefore, job descriptions were used exclusively for SES assignments. Thus, the SES assignment is at best a crude measure of social class that is prone to errors caused by wide variations in income and education within each occupational category. It is also understood that the likelihood of exposures to occupational carcinogens is, in most cases, directly associated with job class, thus correlations between occupation- based SES assignments and exposure may be unavoidable. Moreover, the occupation-based SES schema may actually provide a better measure of exposure rather than social class, thus it may not be possible to disentangle the effects of exposure and SES on leukemia risks.

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5.0 Exposure Assessment

The exposure assessment is a key element in estimating cancer risks from ionizing radiation exposure. Both quantity and quality of exposure information are important parameters that can have a marked effect on study results. Fortunately, individualized quantitative measurements of ionizing radiation exposure are available for most U.S. nuclear workers. These measurement data support exposure estimates that are typically far superior to estimates for most other agents found in the workplace. Nevertheless, measurement data are still subject to random error and systematic biases that can significantly impact dose-response analyses, especially when risks are near the null.

To avoid an observer bias, the exposure assessment was conducted using a study roster that included personal identification, employment history, and the relevant exposure time period for each worker, but did not reveal case status. Thus, the exposure assessment was first comprised of estimates for each worker from first exposure to either the latest date of attained age of any related case or the study end date. These files were later translated into files suitable for analyses that account for exposure periods defined by each risk set and any assumed exposure lag. Exposure truncation, when necessary, was conducted assuming a uniform exposure rate over the measurement period. For example, a control who received 100 mGy in the year 1985 and whose cutoff date was July 1, 1985 was assigned a dose of 50 mGy for the last period of exposure.

Radiation exposures were assessed using the principles of dose reconstruction recently published by the NCRP [Napier et al. 2009]. The primary agent under study was exposure to penetrating low-LET ionizing radiation from work in the primary facilities, other employment, and work-related medical examinations. Workplace exposures to neutrons, and internally deposited radionuclides, such as plutonium and tritium, were also considered as potential cofactors in dose-response analyses. All radiation doses were quantified in terms of absorbed dose to hematopoietic tissue. Benzene exposures

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were assessed as “scores”, which accounted for both the intensity and duration of the potential exposure.

5.1 External low-LET ionizing radiation

Occupational radiation dose to bone marrow is most likely to be acquired by exposures to penetrating low-LET radiations (i.e., photons) or high-LET radiations (e.g., neutrons) from sources outside the body (hereafter referred to as “external dose”). Under typical working conditions in the facilities under study, whole-body exposures to penetrating gamma or x-ray radiations of energies greater than 100 keV contribute most to the total dose accumulated in a working lifetime. Thus, the primary study variable ( Drbm ) was defined as the cumulative absorbed dose to hematopoietic bone marrow resulting from penetrating photon exposures encountered in the workplace from external sources. The dose reconstruction efforts to estimate Drbm are discussed in this section.

The discussion on external low-LET exposures is divided into three parts. The first part provides

background information on monitoring practices at study facilities. The background is followed by a

description of the methods used to estimate doses from low-LET irradiation. A summary of the results is provided at the end of the section. Examining the effects on the relation between Drbm and leukemia from other radiations was also planned; therefore, separate estimates of the cumulative dose from neutrons, plutonium deposition, and tritium were made and are discussed in later sections of the report.

5.1.1 Facility dosimetry

Early personal monitoring at MED facilities was conducted based on the likelihood of exceeding threshold doses and relied on the pocket ionization chamber, otherwise known as the pencil dosimeter or pocket meters. Personal radiation monitoring began at ORNL in October 1943 using paired pocket meters, which provided the dose of record until replaced by film badges on June 26, 1944 [Hart 1966].

Pocket meters were also the primary dosimeter at Hanford between January and September 1944

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[Wilson et al. 1990]. During these periods, both facilities used Victoreen Projection Minometers [Parker

1946b] that were manufactured from tubular bakelite and machined aluminum parts [Hart 1966]. These pocket meters were calibrated using radium standards [Parker 1946b], had a useful range of 1.29 x 10 -6

C⋅kg -1 (5.0 mR or ∼0.05 mGy) to 1.16 x 10 -5 C⋅kg -1 (∼0.45 mGy) [Hart 1966], and provided dose

information in increments of 1.29 x 10 -6 C⋅kg -1 [Deal and Morgan 1947].

Pocket meters were prone to false readings, and the results from a single pocket meter reading were considered unreliable [Deal and Morgan 1947]. The false readings occurred from discharges due to insulator leakage or mechanical shock, and consequentially resulted in higher than actual indicated exposures. To minimize the impact of false readings, workers’ doses were recorded from the lower reading of two issued meters [Parker 1946a]. The pocket dosimeter was largely replaced by film dosimetry before the end of the MED era. Because of the short period of service and the MED era employment exclusion, only a few study subjects were potentially exposed during the use of pocket meters as primary dosimetry.

Most external dose information in our study originates from worker dosimetry using photographic film emulsions. "Film-badges" were first introduced at MED facilities in early 1944 [Pardue et al. 1944]. These dosimeters provided a permanent record of exposure incurred over a monitoring period. Film-badges were more reliable than pocket meters and were less affected by working conditions. Nevertheless, film-badge monitoring was relatively insensitive compared to dosimetry in use today and biases from geometry, incident radiation energy, and environmental factors were evident.

Film badge monitoring continued on a large scale until the emergence of the thermoluminescent dosimeter (TLD) in the mid 1970s, which continues to be the primary personal dosimeter in most industrial settings. Details on film metering at study facilities are available elsewhere [Daniels and

Schubauer-Berigan 2005; Daniels et al. 2004; Schubauer-Berigan et al. 2005].

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Beginning in the mid-1980s, facilities within the DOE complex began preparations for accreditation under the newly developed DOE Laboratory Accreditation Program (DOELAP) [DOELAP

1986]. The DOELAP was established to ensure the integrity of personnel dosimetry results by providing standards for acceptance. The accreditation process includes both an evaluation of dosimeter performance against established criteria and an onsite assessment of system quality assurance, documentation, and technical adequacy. The penetrating dose performance standard used in these evaluations is in terms of the International Commission on Radiation Units (ICRU) operational quantity of personal dose equivalent penetrating to a depth of 10 mm [Hp(10)]. Irradiation standards for testing response to low doses of high-energy photons consist of on-phantom exposures from collimated 137 Cs to doses between 0.3 and 100 mSv [McDonald et al. 1992]. Accreditation is required for all DOE facilities.

To this end, reported dosimetry results are considered normalized to Hp(10) in the anterior-to-posterior

(AP) direction following DOELAP accreditation.

5.1.1.1 Hanford

Hanford used a two-element film dosimeter with the DuPont Type 552 film from October 1944 to March 1957. [Wilson et al. 1990]. Penetrating exposures were assessed by measuring the optical density under 1-mm silver filters using Type 502-sensitive film. The two-element badge was replaced by a multi-element dosimeter in April 1957, which remained in service until replaced by TLDs in January

1972 [Nichols et al. 1972]. Dosimeter calibrations were performed in-air using 226 Ra calibration sources

and units of exposure. The detection levels using Type 502 film and Type 508 film were 1.03 x 10 -5 C⋅kg -1

(∼0.40 mGy) and 5.16 x 10 -6 C⋅kg -1 (∼2.0 mGy), respectively [Wilson et al. 1990].

Hanford introduced the TLD to a limited number of workers in June 1971. This TLD was designed to be worn for periods of up to one year by workers not expected to receive more than 10% of the quarterly dose limit [Wilson 1987]. Starting in January 1972, a multipurpose TLD was placed in service

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for radiation workers. Both dosimeters used elements consisting of 7LiF:Tl block (TLD 700 phosphor) for beta-gamma dose measurements. TLD dosimeter calibrations were performed in-air using a radium source, until replaced by 137 Cs in 1977 [Wilson et al. 1990]. Dosimetry processing was performed monthly, and the detection level was 0.2 mSv. Starting in September 1984, calibration exposures were changed to on-phantom, and a coefficient of 1.03 was used to convert in-air exposure to tissue dose

[Wilson et al. 1990]. The first successful DOELAP performance test was in 1988. The dosimetry system has been accredited under DOELAP since January 1, 1990.

5.1.1.2 INL

The ORNL-type two-element film badge was put in service at INL on August 20, 1951 [Cipperley

1958b]. Penetrating radiation results were taken from under the 1-mm cadmium filter. The dosimeter used the DuPont Type 552 film packet (in most areas) or the DuPont Type 558 packet. The minimum sensitivity (radium gamma) was about 7.74 x 10 -6 C⋅kg -1 (∼0.30 mGy) for Type 552 film packet and about

2.58 x 10 -6 C⋅kg -1 (∼0.10 mGy) for Type 558 packet. By March 30, 1958, INL had adopted the use of a multi-element dosimeter that combined a dosimetry system with the security credentials [Cipperley

1958a]. The dosimeter used a modified Hanford-Type film badge that included three filters (0.10 cm Cd,

0.013 cm Ag, and 0.5 cm Al) and a plastic holder. The badge used DuPont Type-558 film packets, which provided a detection capability of about 2.58 x 10 -6 C⋅kg -1 (∼0.10 mGy). INL reported dosimeter accuracy for x- and gamma ray exposures under “normal conditions” to be ±0.12 mGy in the range of 0.10-0.40 mGy and within ±12% in the range 0.40 mGy to 10.0 Gy [Cipperley 1966]. This design continued to be used into the mid 1970s, although later badges incorporated Kodak Type 3 film as a replacement for

DuPont film. Penetrating radiation calibrations of all film dosimeters were in-air and corrected to dose- equivalent using 226 Ra calibration sources. Calibration geometries were AP and without backscatter

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correction. Film badge monitoring intervals were: weekly (1951-1958); and biweekly or monthly, except in high dose areas (1958-1966).

TLDs were first issued to a limited group (<200) of INL workers in November of 1966. Beginning

January 1, 1967 INL issued TLDs to all low-risk workers (i.e., < 2.0 mSv∙y -1); however, high exposure

groups (e.g., ICPP and NRF workers) continued to use film dosimetry until 1975. The first TLD consisted

of two Teflon discs that contained LiF and was supplied by Harshaw. The radium sensitivity was reported

to be 2.58 x 10 -6 C⋅kg -1 (∼0.10 mGy) and was accurate to within ±15% but ≥ 2.58 x 10 -6 C⋅kg -1. This TLD

design continued to be used, with slight modification, until replaced by the four-element Automatic

Thermoluminescent Analyzer System (ALTAS) in February 1974. The system had a sensitivity of about

7.74 x 10 -6 C⋅kg -1 (∼0.30 mGy). INL began phasing out the ATLAS dosimeter beginning in December 1974.

By May 1975, INL had deployed the Harshaw two-chip system (TLD 700 phosphor), which continued to be used until replaced by the Panasonic four-chip dosimeter system in 1986. This was also the year of

INL’s DOELAP accreditation. This dosimeter used lithium borate (Li 2B4O7) TLD elements with plastic and

aluminum filtration for a marked improvement in penetrating dose measurements. The minimum

recording values using this system ranged between 0.1-0.15 mGy. TLD calibrations utilized a 226 Ra calibration source until replaced by 137 Cs in 1981. During this time, INL also began using a calibration phantom and calibrated to a tissue absorbed dose rather than exposure. In 1987, INL began using a

Shepherd panoramic irradiator with a 137 Cs source for TLD calibration. The irradiator did not use a

phantom, but results were adjusted using information from previous TLD irradiations that included a

phantom.

5.1.1.3 LANL

Select LANL workers wore film badges as early as 1944 [Hempleman 1944]. First generation

dosimeters used Kodak Eastman Type K film packages with a 0.5-mm lead filter (for penetrating dose

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discrimination) housed in a sectioned case manufactured from 0.5 mm of brass [Ferry 1944]. The lead cross was replaced by a brass filter in 1947 [Buckland 1947]. In 1950, LANL issued a new two-filter dosimeter that also used a lead filter for deep dose determination. [Lawrence 1956]. This badge was replaced in April 1951 by a similar dosimeter using two 0.5-mm filters, one made of brass and the other made of cadmium [Littlejohn 1961], that enabled energy-response correction for penetrating radiation

[Storm 1951]. LANL film dosimetry remained essentially unchanged until the use of the multi-element

“Cycolac” film badge beginning in October 1962 [Dummer 1963]. The Cycolac badge (named after the type of plastic used for the holder) contained two five-element (bismuth, erbium, gold, tantalum, and either molybdenum or gadolinium) composite filters and an unfiltered area to improve measurements of low-energy x-rays and mixed radiation fields common to plutonium work [Storm and Shlaer 1964b].

The Cycolac badge fully replaced the brass-cadmium badge by 1968 and was used until film monitoring was completely discontinued [Lawrence 1978]. Calibrations were performed in-air using 226 Ra calibration

sources, until replaced by 60 Co sources in or about 1960 [Littlejohn 1961]. The routine use of calibration

phantoms began on or about 1977 [Storm et al. 1977]. The detection threshold for film badge

monitoring was 1.03 x 10 -5 C⋅kg -1 (∼0.40 mGy) per monitoring period [Lawrence 1978].

On April 28, 1978, LANL began issuing TLDs to visitors. Beginning August 28, 1978, LANL began replacing film badges with TLDs for its regularly monitored workers [Storm 1978]. The TLDs consisted of a Cycolac plastic holder containing three Harshaw TLD-700s ( 7LiF) and one Harshaw TLD-600 ( 6LiF)

mounted in an aluminum plate. Penetrating gamma radiation was determined using one TLD-700

covered with a 90 mg-cm -2 thick copper filter [Cortez et al. 1977]. This TLD continued to be used until

replaced by a new eight-element design in 1998. A complete description of the dosimeter is provided by

Storm et al. [1981].

Dosimetry exchanges were performed monthly through 1979, and then monthly or quarterly

thereafter, based on an assessment of exposure potential. Calibrations were performed using a 60 Co

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calibration source [Cortez et al. 1977]. The dose-response under the filter was linear for x-rays in excess of 35 keV and doses <2.0 Gy. The reported sensitivity of the dosimeter was 0.1 mSv [LASL 1984]. The TLD underwent pilot testing for DOELAP accreditation in 1985 [Gasell 1985]. A positive bias was observed during the DOELAP pilot testing that was attributed to a 10% difference in dosimeter response between

LANL in-air calibration and the backscatter from the on-phantom 137 Cs irradiation by DOELAP [Lawrence

1986]. LANL corrected for this bias in February 1986 by reducing recorded doses by 10%. The LANL

dosimetry program successfully completed DOELAP performance testing in 1987.

5.1.1.4 ORNL

ORNL began film monitoring in 1944 with a two-element film meter that incorporated a 1-mm

cadmium filter and Type 502 sensitive film for penetrating dose assessment [Hart 1966]. A four-element

dosimeter was introduced in 1953, which also used the cadmium filter and Type 502-sensitive film

[Davis et al. 1954]. In 1959, the film packet was replaced by DuPont Type 554, which contained the Type

555-sensitive emulsion. Sometime between 1968 and 1969, the DuPont Type 554 film packet was

replaced by Eastman Kodak Type 2, which continued to be used until replaced by TLDs in 1975. Film

calibration procedures used 226 Ra sources [Craft et al. 1953] and doses were calculated in-air prior to

1954. Beginning in 1954, calibration procedures included 5 cm of plastic as a phantom to simulate

backscatter conditions [Davis et al. 1954]. Mitchell et al. [1993] reported a detection level of 7.74 x 10 -6

C⋅kg -1 (∼0.30 mGy); however, dosimetry records suggested a range of recording thresholds over the period of film badge use. Beginning in 1944, exposures were reported above 2.58 x 10 -6 C⋅kg -1 (∼0.10

-6 -1 mGy). By mid-December 1947, the reporting threshold was increased to 7.74 x 10 C⋅kg . The threshold was then reduced again to 2.58 x 10 -6 C⋅kg -1 by the mid 1950s.

The ORNL TLDs replaced film badge monitoring in 1975 and used combinations of Harshaw LiF phosphors (i.e., TLD-100, TLD-600, and TLD-700) for photon dose assessment. Several variations in TLD

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design have been used throughout the years, depending on the radiation exposure characteristics expected to be encountered. Beginning in 1980, efforts to consolidate personnel dosimetry services among the Oak Ridge Complex prompted Union Carbide Corporation, Nuclear Division (UCC-ND) to issue the UCC-ND TLD. The two-element dosimeter was manufactured by the Harshaw Chemical Co. and incorporated LiF phosphors [McLendon 1980]. This dosimeter continued to be used until replaced by the four-element Harshaw Model 8800 of the Centralized External Dosimetry System (CEDS) in January

1989. The CEDS satisfactorily completed the DOELAP performance testing in October 1989 and was accredited on January 31, 1992.

5.1.1.5 PNS

Although most nuclear work at the shipyard did not commence until the late 1950s, film meters were first placed into service at to monitor a small group of radiographers beginning in 1950. Early badges were equipped with Kodak Eastman Type K film and a 0.5-mm lead filter. In 1952, Eastman DF-7 film replaced the Type K film packet. In 1957, PNS used a DuPont SX-233 film packet with Type

555-sensitive film and a 1-mm cadmium filter for penetrating dose measurements. The film meter remained unchanged until July 1969, when the Kodak Type 3 Film Radiac Pack was placed in service.

From July 1950 to July 1958, calibration was performed in-air with 226 Ra calibration sources. Calibrations using 60 Co sources and a 10-cm wood block to simulate backscatter began in August 1958. The 60 Co source was replaced by 137 Cs in 1964. The detection threshold was reported to be 5.16 x 10 -6 C⋅kg -1

(∼0.20 mGy) from 1950 through 1969.The detection threshold was reduced by 50% to 2.58 x 10 -6 C⋅kg -1

with the introduction of Kodak Type 3 film in 1969 [PNS 1980]. Monitoring periods were bi-weekly prior

to 1960 and monthly thereafter.

PNS first used TLDs (DT-526/PD) in 1974. The dosimeter consisted of two CaF 2:Mn phosphors

that sandwich a wire-strip heating element encased in a glass bulb. The entire bulb assembly was

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housed in a tamper-proof container designed to be worn on the belt of the worker. The dosimeter is manufactured by Harshaw Chemical Co. and is reported to have an exposure range of 1.29 x10 -6 to 1.29

C⋅kg -1 of air for gamma energies of 80 keV to 1.3 MeV, with an accuracy of ±10% or 1.29 x10 -6 C⋅kg -1 of

air, whichever is greater [PNS 1980]. The detection threshold for this dosimeter is 0.01 mSv per

monitoring period [Brann 1988]. The DT-526/PD continues to be the primary personal dosimeter for

shipyard employees. The TLDs were processed daily (when exposed) and the results were summed

monthly for reporting purposes. PNS first completed accreditation by the National Voluntary Laboratory

Accreditation Program (NVLAP) in 1988. This program provides third-party accreditation to testing and

calibration laboratories, similar to the DOELAP program.

5.1.1.6 SRS

SRS began its personal monitoring program by issuing the ORNL two-element film dosimeter to

workers in 1951 [Taylor et al. 1995]. ORNL dosimetrists processed SRS badges until March 1953;

however, the ORNL dosimeter continued to be used until SRS introduced its design in late 1959. The SRS

dosimeter used DuPont Type 555-sensitive film and a 1-mm silver filter for penetrating radiation [Wright

1959]. This dosimeter design continued to be used until replaced by the TLD in 1970 [Taylor et al. 1995].

Calibrations were conducted at ORNL until March 1952, and then at SRS using 226 Ra. All calibrations were conducted in-air. The reported detection level was 7.74 x 10 -6 C⋅kg -1 (∼0.30 mGy).

On April 1, 1970, SRS became the first of the study sites to introduce the TLD as the primary means for measuring beta-gamma exposure [Taylor et al. 1995]. The existing film badge housing was modified to incorporate two 7LiF TLD chips manufactured by Harshaw Chemical Co., one centered over

the open window and the other behind an aluminum shield. This basic configuration was used with

minor modifications until replaced by the commercially available Panasonic Models UD-802 and UD-813

on July 1, 1982 [Taylor et al. 1995]. The UD-802 used two CaSO 4:Tm elements for whole-body gamma

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7 dose monitoring. The UD-813 was a two-element dosimeter incorporating Li 2B4O7:Cu chips and was

used exclusively for FB-Line and HB-Line workers exposed to lower-energy x-rays from plutonium. The

UD-813 was replaced by the four-element UD-812 in 1987, which added the CaSO 4:Tm elements of the

UD-802. The TLDs were calibrated using a 226 Ra source until replaced by a 137 Cs source in January 1987

[Taylor et al. 1995]. Calibration phantoms were used beginning with the first dosimetry cycle in 1986.

Also starting in 1986, a conversion coefficient of 1.03 was used to correct recorded exposures to penetrating dose [Taylor et al. 1995].

SRS began preparing for the DOELAP accreditation process in the mid 1980s. During initial preparations for accreditation, an overall negative bias in the TLD dosimetry system was discovered, prompting program changes and a subsequent increase in recorded dose. Based on these changes, SRS recommended gamma dose bias factors 0.894 and 0.962 for correcting doses prior to 1986 and from

1986 to 1987, respectively. The dosimetry system satisfactorily completed DOELAP performance testing in 1987 [Taylor et al. 1995].

5.1.2 Methods

Measurement data from personal monitoring of x- and gamma ray exposures are abundant and are the primary source of exposure information for this study. Records of external ionizing radiation exposure measurements for each study participant were abstracted from: 1) dosimetry databases maintained by individual sites, 2) datasets constructed for previous health studies conducted by NIOSH,

3) the DOE Radiation Exposure Monitoring Systems (REMS), and 4) the USNRC Radiation Exposure

Information Reporting System (REIRS). Searches were expanded to include information on exposures that occurred during periods of employment outside of the primary study facilities. Raw data from multiple sources were carefully combined to eliminate record duplication. For example, the Oak Ridge facilities (i.e., ORNL, Y12, and K25) shared information and it was common to find identical information

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in the separate exposure files maintained by each facility. Final dosimetry tables were constructed and scrutinized for aberrant exposure patterns and any remaining overlapping exposure periods.

Questionable entries were flagged and reconciled by hand using all available information (e.g., hard copy exposure records, REMS/REIRS data, and claimant files). The master electronic exposure files used for data abstraction are described in Table 5-1.

Recorded values were adjusted to normalize results between facilities and across time.

Estimates of tissue absorbed dose were calculated using combinations of conversion coefficients recommended by the ICRP [ICRP 1987; ICRP 1996]. Dose in monitoring intervals were summed from first exposure to the time of the attained age of the index case, subtracting any exposure lag period. A

-1 uniform dose rate (i.e., Drbm ∙d ) was assumed to prorate the period of last exposure (if truncated by

study restriction).

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Table 5-1. External dosimetry data files for penetrating radiation exposure information.

Dose ID filename FAC Years Obs. Vars. Workers Vars Comments 1 rems_ext.sas7dbat ALL 1990-2008 7,939 59 641 G, N Extraction from DOE REMS database based on roster search

2 reirs_ext.sas7bdat ALL 1900-2008 707 23 707 G, N Extraction from USNRC REIRS database based on roster search

3 Canndose.sd2 Amchitka 1971-1972 7,249 19 2,329 G Data from site exposure database

4 emd_prev_det_rou.sas7dbat FEMP 1953-2005 10,564 28 1,567 DDE Exposure file from HIS20 database for early workers at the FEMP (offsite)

5 edd_inactive_ird_exp - FEMP 1952-2005 346,105 19 34,024 G,N Exposure file from HIS20 database .sas7dbat for early workers at the FEMP (onsite)

6 dos_master.sas7bdat HAN 1999-2001 68,088 40 11,262 G, N, T Table from REX

7 dos_master_data.lst HAN 1951-2003 105 40 65 G, N, T Table from REX (offsite dose)

8 dos_summ_results.sas7bdat HAN 1940-1999 984,722 24 199,817 G, N, T Table from REX (onsite dose)

9 dos_summ_results_data.lst HAN 1902-1992 16,050 24 9,194 G, N, T Table from REX (offsite dose)

10 dose.sas7bdat INL 1900-1999 1,248,673 19 36,570 G, N From previous cohort study [Schubauer-Berigan et al. 2005]

11 hp_external.sas7dbat (k25) K25 1981-1988 32,957 49 5,584 G, N Table from Lockheed Martin dosimetry database from K25.

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Dose ID filename FAC Years Obs. Vars. Workers Vars Comments 12 orgdp_pgdp_tld.sas7dbat K25 1945-1988 166,246 43 13,083 G, N Table from Lockheed Martin dosimetry database from K25.

13 lamulw02.d4 LAN 1980-1990 495,916 18 7,356 G, N, T From Wiggs et al. [1994]

14 niosh2010.sas7bdat LAN 1946-2007 10,006 26 92 G, N, T Extraction from LANL database sent to NIOSH by J. Voltin in 2010 for current study 15 voltin_ext_1.sas7bdat LAN 1945-2003 14,603 25 169 G, N Extraction from LANL database sent to NIOSH by J. Voltin in 2003 for LCCS 16 lan_zia_ext.sas7bdat LAN 1944-1985 174,997 13 27,581 G, N, T Data received from LANL in 1997

17 lamulw02.d3 LAN 1944-1985 130,165 15 16,874 G, N, T From Wiggs et al. [1994]. Incomplete from 1980-1985.

18 person_extdose.sas7bdat ORNL 1989-2000 328,106 36 18,259 G,N Table from Lockheed Martin dosimetry database.

19 hp_external.sas7bdat ORNL 1980-1988 60,904 49 9,239 G,N Table from Lockheed Martin dosimetry database.

20 ornl_tld.sas7bdat ORNL 1943-1987 821,087 20 36,172 G,N Table from Lockheed Martin dosimetry database.

21 pk1.data.t93014.tld2.txt Piketon 1981-1986 92,700 12 3,836 G Data from site exposure database

22 pk1.rawdata.t92064.film*.txt Piketon 1953-1980 277,279 14 5,401 G Data from site exposure database

23 dose24.sas7bdat PNS 1934-1996 88,739 53 13,475 G, N From previous cohort study [Daniels et al. 2004; Silver et al. 2004]

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Dose ID filename FAC Years Obs. Vars. Workers Vars Comments 24 srsabstrniosh.sas7dbt SRS 1965-1989 1,752 5 855 G File coded for NIOSH/UNC study [Richardson et al. 2007]

25 fayerwet.sd2 SRS 1952-1978 221,330 13 8,194 G, N, T File coded for DuPont health study by William Fayerweather

26 srpabstr.sd2 SRS 1951-1991 5,242 11 1,059 G, N, T File coded by ORISE researchers for health study [Cragle et al. 1999]

27 exposure.sas7dbat SRS 1950-1999 390,526 19 43,998 G, N, T From HPAREH database

28 hp_external.sas7dbat (y12) Y12 1965-1999 457,100 49 25,435 G, N Table from Lockheed Martin dosimetry database from Y12

29 y12_tld.sas7dbat Y12, 1945-1991 799,829 43 35,551 G, N Table from Lockheed Martin PAD dosimetry database from Y12

30 gammadetail.sas7dbat RFP 1952-1970 251,212 35 5,340 G Table from the Neutron Dose Reconstruction Project (NDRP) database from RFP [Falk et al. 2005] 31 neutrondetail.sas7dbat RFP 1952-1970 110,782 38 3,638 N Table from the Neutron Dose Reconstruction Project (NDRP) database from RFP [Falk et al. 2005] 32 TLDdetail.sas7dbat RFP 1970 10,223 17 3,427 G, N Table from the Neutron Dose Reconstruction Project (NDRP) database from RFP [Falk et al. 2005] 33 Iarc89_dose.sas7bat HAN 1944-1989 369,136 14 37,012 G, N, T Table obtained by NIOSH in 1996 that contains Hanford dose data from first IARC study [Cardis et al. 1995]

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5.1.2.1 Hanford

In November 2001, Hanford dosimetry services provided NIOSH with 66 data tables extracted from the Hanford Radiation Exposure Databases (REX) that included personnel, external dosimetry, bioassay, and in vivo dosimetry information [MacLellan 2001]. These data were appended with an additional submittal to NIOSH in March 2004 [MacLellan 2004]. External dosimetry data included annual exposure summaries for shallow, deep, neutron, tritium, and extremity exposure. The data provided in

REX are rounded to the nearest 0.1 mSv. Zero values were recorded as placeholders for non-monitoring; thus, monitoring status cannot be established. This file also provided work locations for a number of individuals.

5.1.2.2 INL

The records used to identify workers and assess individual exposures to ionizing radiation were collected from the site from the mid-1990s to the early 2000s. Details on these record systems are described elsewhere [Schubauer-Berigan et al. 2005]. In general, NIOSH researchers collected several electronic databases and hard copy records from INL employment and health physics record systems for use in a previous cohort study. Among these record systems were three electronic files containing external dosimetry records for monitored workers. The Master Update Dump (MUD) database contained the file EXP_HIST that provided 207,663 observations of annual dosimeter results for 49,480

INL workers exposed prior to 1986. The MUD database was superseded by the Radiation Dosimetry

System (RDS), which was maintained by the Operation Dosimetry Unit (ODU) from 1986 to the time period of the last NIOSH data capture in 2001. The RDS housed both internal and external exposure data and contained 797,941 external dose observations. Finally, exposure information for civilian workers exposed at the NRF was kept in a separate database after 1973 (NRF database) that held 4,900 observations. Information was abstracted from these record systems to populate a single dose file (i.e.,

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dose.sas7bdat) for the INL cohort by linking on several key variables. Substantial data cleaning was necessary to prevent duplication in the combined file because of overlap between MUD and NRF databases, especially in the years from 1951 to 1973. The final external dose file contains 1,248,673 observations that span the time period from beginning of INL operations through 1999.

5.1.2.3 LANL

Early in the NIOSH OERP program, LANL provided a set of Excel® spreadsheets containing exposure information on workers ( n=27,581) from 1944 through 1985. It was presumed that the dataset

represents all available dosimetry information on LANL and Zia workers prior to 1985. However,

inspection of these data for the LCC exposure assessment revealed that the exposure information was

incomplete, of suspect quality, and that records tracing the data to the actual source were not available.

Thus, these data were not preferred for this study.

To reduce uncertainty, NIOSH sought internal and external dosimetry information through subsequent requests made to site dosimetry services. LANL dosimetrists abstracted data from their dosimetry database systems using rosters of study subjects known to be employed at LANL. The first extraction took place in 2003 for 169 LANL/Zia workers participating in the LCCS. Of these workers, 47 were selected to the current study and their information was made available for the current study. A second extraction of dosimetry information on the remaining 92 study subjects was completed in 2010.

External exposure data were available for all monitoring periods. The data included fields for shallow, deep, and neutron exposures. All exposure data provided were rounded to the nearest 0.1 mSv. Null values were used as placeholders for non-monitoring periods.

5.1.2.4 ORNL

In the early 2000s, NIOSH received a large collection of data files from the Martin Marietta radiation exposure database containing measurement data for all Oak Ridge DOE workers, including

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current and former workers with employment at ORNL. The electronic dosimetry datasets had been consolidated by badge number during data migration by ORNL. Unfortunately, badge numbers were not unique and the potential existed for merged dosimetry results for persons with duplicate badge assignments. Hard copy dosimetry records and work histories were used to identify study subjects with a potential for merged electronic data. External dosimetry information for these study subjects was coded from the hard copy records.

5.1.2.5 PNS

The results for each shipyard monitoring interval and for each exposed worker were recorded on individual exposure record cards. The information from the cards was then transcribed to form DD

1141, Record of Occupational Exposure to Ionizing Radiation (currently NAVMED Form 6470/10). These forms were filed in the medical records of each employee, and a copy was kept in the radiological files

maintained at PNS. Annual exposure data from the form DD 1141s were coded into an exposure

database maintained by PNS. These data were obtained from the site and used to construct an exposure

file for 13,475 nuclear workers who participated in a previous cohort study [Silver et al. 2004]. The

cohort study file (DOSE24) was augmented using other record sources and adjusted to account for a

number of potential sources of exposure bias (e.g., prior doses from other employment and

administratively assigned doses). These adjustments were described elsewhere [Daniels et al. 2004]. The

data included fields for shallow, deep, x-ray, and neutron exposures rounded to the nearest 0.01 mSv.

Null values were used as placeholders for non-monitoring periods.

5.1.2.6 SRS

Three electronic databases were received from site dosimetry systems and migrated into SAS

datasets. These data files are: 1) FAYERWET.SD2 (also known as the “Fayerweather file”); 2)

exposure.sas7bdat (HPAREH file); and 3) SRPABSTR.SD2. In addition, a dataset (UNC/NIOSH.sas7bdat)

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was obtained that contained supplementary exposure information coded by researchers under a collaboration between the University of North Carolina (UNC) and NIOSH [Richardson et al. 2007]. A description of these data is provided below.

The Fayerweather File was originally constructed by Dr. William Fayerweather, an epidemiologist for the DuPont Corporation who was conducting health studies of SRS workers. The file was completed in 1986 and contains exposure data for select individuals who terminated prior to 1979.

The dataset consists of 221,330 exposure records for 8,192 workers employed from 1952 through 1978.

When constructing the dataset, Dr. Fayerweather used a template whereby doses were initially set to zero as placeholders for each year of exposure for each worker. As dose values were identified by hard copy records, the zero values were replaced with record values. Thus, workers have records for all years,

regardless of actual monitoring status, and one cannot distinguish between a monitored and

unmonitored zero value. Incomplete fields in the file include tritium and neutron exposure fields,

although recorded exposures greater than zero are typically reliable. Given that the file was compiled

from hard copy printouts, the deep and shallow dose values include contributions from neutron and

tritium dose. Because tritium and neutron fields were incomplete, hard copy logbook information was

used to separate the tritium and neutron dose components from the deep and shallow dose data.

The Health Protection Annual Radiation Exposure History (HPAREH) database became fully

functional in 1979. HPAREH consolidated data from the personnel radiation exposure files and logbooks

(1951-1972) and magnetic tapes (1973-1979). Only individuals who were actively working at the site in

1979 were originally included in the dataset. However, as the site researches dose history to support

litigation, worker requests, and other dose reconstruction efforts, data from individuals terminated prior to 1979 have been subsequently added to HPAREH. Nevertheless, the Fayerweather information remains as the primary source of dose information for SRS workers terminated prior to 1979. NIOSH received data tables extracted from HPAREH in 2000 (HPAREH_2000). The data tables were converted

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into a single SAS dataset containing 390,526 observations of annual exposure data pertaining to 43,998 individuals exposed from 1950 through 1999. Shallow, deep dose, tritium, and neutron components were reported separately.

During the course of the Savannah River Health and Mortality Study [Cragle et al. 1999], researchers from the Oak Ridge Institute for Science and Education (ORISE) identified 1,052 individuals who were not included in the Fayerweather file or the HPAREH file. ORISE researchers manually abstracted data from hard copy records for these individuals and created the electronic file SRPABST.

The file contains a total of 5,242 records for a total of 1,052 individuals. The data are for the period from

1951 through 1991. Similarly, Richardson et al. [2007] abstracted dosimetry information from hard copy information on 855 workers for another SRS epidemiologic study to create the UNC/NIOSH dataset. The file contains exposure information on workers who were employed in the period from 1964 to 1979 and were missing from the Fayerweather and SRPABST files.

5.1.2.7 Bias and Uncertainty

Variations in reported exposures between study facilities and across time resulted from differences in incident photon energy, exposure geometry, dosimeter type, and dosimetry methods. The dosimetry response for each facility and monitoring cycle was examined to develop methods for normalizing reported doses across study facilities and estimating the dose to the hematopoietic bone marrow for each study subject. Dosimetry response was defined as the ratio of the recorded dose to

Hp(10) or to HT. Final dose estimates were made by adjusting facility records of exposure, using bias factors for each dosimeter and facility type (Table 5-3). These bias factors were derived using methods suggested by Daniels and Schubauer-Berigan [2005] for the LCCS and others for the 15-Country Study

[Thierry-Chef et al. 2001; Thierry-Chef et al. 2007; Thierry-Chef et al. 2002].

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In general, bias factors ( B) were defined such that:

D B = m (5-1) D

where D m is the measured dose and D is the unknown true dose. The parameter B was assumed to be lognormally distributed with geometric standard deviation ( S) taken from a two-parameter lognormal

-1 -1 distribution. An uncertainty factor, K, was defined as the interval [ Dm(KB) , DmKB ] estimated to contain

D at the 95% confidence level. For lognormally distributed values, K = ( S)1.96 .

Daniels and Schubauer-Berigan [2005] calculated average bias factors and estimated

measurement uncertainty using Monte Carlo simulation techniques. Uncertainty was evaluated in two

parts. First, energy- and geometry-specific bias factors ( Bpe, eg ) were calculated for ranges of photon

energy ( pe ) and exposure geometry ( eg ):

Bdos = exp[ ∑ fpe feg ln( Bpe ,eg ]) (5-2)

where fpe is the fraction of exposure from photons within the two energy bands (i.e., 100-300 keV and

300-3000 keV); and feg is the fraction of exposure in the anterior-to-posterior (AP), rotational (ROT), or

isotropic (ISO) exposure geometries. The average values for f100-300 and f300-3000 were 0.2 and 0.8, respectively and f100-300 was assumed to be uniformly distributed from 0.15 to 0.25 with f300-3000 set equal

to the quantity 1- f100-300 . The value fAP was assumed uniformly distributed from null to unity and quantity fISO +f ROT was assumed equal to 1-fAP . Values for fISO and fROT were solved by the quantities fISO = a(1-fAP ) and fROT = 1-fISO where variable ( a) followed a triangular distribution ranging from zero to one.

Average values for fAP , fISO , and fROT were 0.5, 0.5, and 0.0, respectively.

Second, an evaluation of the random uncertainty associated with dosimeter processing and dose interpretation was performed. Where practical, uncertainties were combined using Monte Carlo simulation or by simple approximation. In some instances, data were not available to estimate

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simulation parameters. When sufficient data were unavailable, parameters were assigned based on expert judgment. For example, ±20% was assumed as the range of uncertainty (at 95% confidence) for processing of TLDs. The final bias factor was expressed as the product of individual bias factors needed to normalize reported exposures to absorbed dose to active bone marrow ( Dt).

5.1.2.8 Other Sources of Uncertainty

Other sources of dosimeter error have resulted from equipment malfunctions, procedural

violations, and human errors (i.e., recording errors, dose interpretation, etc.) that are independent of

dosimeter type. Some of the more important sources of errors are discussed below.

5.1.2.8.1 Badge Orientation

In most cases, radiation workers were required to position dosimetry facing forward on the

outer clothing on the front surface of the body, between the neck and waist. However, an occasional

worker has been observed not wearing issued dosimetry when required. Others have been observed

with dosimeters in front pockets or in other unapproved orientations [Lawrence 1985]. This can result in

partial shielding of the dosimeter from pens, pencils, or other carried objects resulting in a negative bias

in the dose estimate. In addition, whole-body doses are generally estimated from a single dosimeter

assuming a uniform body exposure perpendicular to the body (e.g., the dosimeter). However, exposure

energies, geometry, and shielding conditions at some work locations may result in heterogeneous body

exposures that are poorly estimated by a single whole-body dosimeter. Although measurement errors

are likely given badge orientation, these errors are not anticipated to be differentially distributed among

cases and controls. Moreover, a large systematic bias in measurements that stem from poor monitoring

geometry is not likely because dosimetry procedures typically required multiple dosimetry or relocation

of the primary dosimetry in jobs involving partial exposures (e.g., glovebox worker).

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5.1.2.8.2 Extraneous Exposure

When not worn by the worker during occupational exposures, dosimetry should be stored in a low-radiation environment. However, there have been observations of dosimeter response resulting from badge storage in high background areas or near small radiation sources. Likewise, some unusual exposures have been traced to non-occupational sources, such as medical x-ray procedures or irradiation by medical isotopes used in treatment of the worker. Under such circumstances, a positive bias may be introduced to recorded occupational exposures. In most cases, significant exposures of this type have been investigated following dosimetry processing and corrections have been applied to the dose of record.

5.1.2.8.3 Working Conditions

The industrial work environment common to most of the study facilities can be detrimental to the equipment used for radiation monitoring. Although uncommon, dosimeter failures did occur. Pocket meters were observed to over-respond from the mechanical shock of dropping the dosimeter, or from colliding with piping and other workplace obstructions. Film dosimeters have been destroyed by dropping into liquids or overexposure because of damaged wrappings. Dosimeters have been lost due to broken capture devices. In these cases, dosimetry results were either lost and not recorded, or were replaced with estimated values. Because personal monitoring programs were designed to ensure worker exposures did not exceed protection limits, most estimating practices were conservative to ensure that

a particular exposure limit would not be exceeded. Therefore, lost or faulty dosimetry is likely to result

in a slight overestimation of dose. However, quality assurance procedures and routine audits of

dosimetry programs helped to ensure dosimeter reliability over most years of facility operations. A

notable exception is earlier operating periods when PIC measurements served as the dose of record.

However, a large bias from unreliable PIC measurements is not expected because few study subjects

were employed during this time.

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5.1.2.8.4 Equipment Defect and Processing Errors

Some uncertainty is expected from equipment defect and errors associated with dosimetry processing. For example, typical errors associated with film badge monitoring were observed from defective film emulsions, no film in packet, developing errors, and badge contamination. Of 289,516 film badges processed in 1945, 384 (0.13%) defective film badges were identified [Parker 1946b].

5.1.2.8.5 Lower-Energy X-ray Contribution

Penetrating (>100 keV) photon exposures were used to calculate low-LET dose. However, some study subjects were exposed to predominantly lower-energy radiation. Plutonium workers at LANL,

Hanford, ORNL, and SRS were primarily exposed to 17 keV and 60 keV photons resulting from the decay of 239 Pu and 241 Am. Special calibration and dosimetry techniques were developed by each exposure site to account for the lower-energy photon dose contribution. In most cases, a fraction of the lower-energy

exposure was summed with the results from penetrating exposures to provide a single effective

penetrating radiation result. Although this practice provided for worker protection, deep dose estimates

are likely biased as a result of attenuation. For example, the typical photon radiation quality

encountered at LANL DP West was characterized by the following spectra: 1) 65-70% 17 keV; 2) 10-20%

60 keV; 3) 1-10% 100 keV; and 4) 0-7% >200 keV [Dummer 1963]. Using these data and assuming there

is insignificant bias in the film badge exposure estimate, the estimated absorbed dose to bone marrow

should be reduced by nearly 30-fold from the badge reading.

Given the expected low contribution to bone marrow dose, efforts were made to exclude the

lower-energy component from dose estimates supporting the epidemiologic study. Where possible,

penetrating photon radiation results were extracted from readings under the shielded portion of the

film badge only. For TLDs and some film monitors, separate accounting of the lower energy component

was not available, and corrections could not be readily implemented. Therefore, the reported

penetrating radiation results may overestimate dose for study subjects assigned as plutonium workers.

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Table 5-2: Film badge filter characteristics by facility and era used

-1 Mass Energy-Absorption Coefficient µµµen ρρρ 2 Filter Filter Thickness Filter Density ρρρ Density Thickness (cm g) ID Facility Era Material (cm) (g cm -3) (g cm -2) 118 keV 208 keV Hanford 1944-1962 1 Silver 0.10 10.50 1.05 0.68 0.16 SRS 1959-1970

2 LANL/Zia 1949-1949 Brass 0.05 8.41 0.43 0.19 0.06

3 LANL/Zia 1951-1961 Cadmium 0.05 8.65 0.44 0.71 0.17

PNS 1952-1974

SRS 1951-1958 4 Cadmium 0.10 8.65 0.87 0.71 0.17 ORNL 1944-1979

INL 1951-1966

5 PNS 1950-1951 Lead 0.05 11.35 0.58 1.65 0.54 LANL/Zia 1950-1951

Lead 0.05 11.35 6 LANL/Zia 1944-1948 1.28 0.85 0.27 Brass 0.08 8.41

7 Hanford 1962-1972 Tantalum 0.05 16.65 0.85 1.52 0.44

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Table 5-3: Bias factors ( B) and uncertainty factors ( K) used to estimate absorbed dose to the active bone marrow ( Brbm , K rbm ) from recorded exposures.

1 Brbm Krbm Sites and approximate time span

1.48 1.28 HAN (1944-1962); SRS (1951-1969)

1.36 1.29 HAN (1963-1970)

1.39 1.29 LANL (1945-1948)

1.65 1.31 LANL (1948-1950)

1.32 1.28 LANL (1951-1961)

1.27 1.28 LANL (1962-1978)

1.42 1.29 LANL (1944); PNS (1950-1952)

1.49 1.29 PNS (1953-1957); ORNL (1944-1953)

1.29 1.28 PNS (1963-1974)

1.37 1.30 ORNL (1954-1958); INL (1951-1957)

1.23 1.29 ORNL (1959-1974); INL (1958-1966)

1.35 1.28 PNS (1958-1963)

1.20 1.24 HAN (1971-1983); INL (I967-1980);ORNL (1975-1989); SRS (1970-1989)

1.26 1.21 HAN (1984-1989); INL (1981-1985); LAN (1978-1986); PNS (1974-1988)

1.33 1.22 HAN (1990-2005); INL (1986-2005); LAN (1987-2005); ORNL (1990-2005); PNS (1989- 2005); SRS (1990-2005) 1HAN=Hanford site; INL= Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; PNS=Portsmouth Naval Shipyard; SRS=Savannah River Site

5.1.3 Results

5.1.3.1 Cohort Level

The base cohort (n=105,245) was matched to the unadjusted dosimetry information to examine

dose distributions by study facility. Only results reported for measurements of low-LET penetrating were

abstracted. Data from site dosimetry databases were available for all facilities except LANL. The dose distribution for LANL was developed from dosimetry files created for a previous epidemiologic study

[Wiggs et al. 1994], which contained information for about three-fourths of the LANL/Zia subcohort. The

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SRS Fayerweather file did not discriminate between monitored and unmonitored years, thus the number of monitoring person-years were estimated from positive doses only.

There were 101,750 workers (96.7%) found with at least one dose measurement recorded in the primary facility dosimetry records. The remaining subjects without matched data ( n=3,496) were employed mainly at LANL ( n=2,904), whereby dosimetry information was known to be incomplete.

Others without dosimetry information ( n=592) may have resulted from poor linkage to matching variables (e.g., incorrect SSN used in match). However, it is unlikely that additional information on this group would greatly influence cohort level results because of its small size (0.6% of the cohort); thus, more complex matching was not attempted.

The mean cumulative dose to each worker from all facilities combined was 17.8 mSv. Among the individual sites, SRS had the highest mean cumulative dose (29.8 mSv), followed by PNS (24.2 mSv),

Hanford (16.9 mSv), INL (13.8 mSv), ORNL (8.8 mSv) and LANL (8.2 mSv). The overall cumulative dose distribution and the dose distributions for each facility were highly right-skewed, whereby most workers had doses less than 10 mSv (74.9%), nearly 90% were less than 50 mSv, and about 0.2% had doses of

400 mSv or greater (Table 5 4).

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Table 5-4. Cohort level cumulative dose (mSv) statistics for low-LET penetrating radiation

Statistic ALL PNS Hanford SRS ORNL INL LANL dose file ID 23 6, 8 24-27 19- 21 10 13, 17 year last monitored 2001 1996 2001 1999 2000 1999 1990 workers by primary 105,245 9,657 23,177 12,585 14,057 33,474 12,296 site monitored Workers 100,510 9,619 22,739 12,541 13,977 32,401 9,233 excl. site overlap 2 monitored Workers 101,750 9,625 28,241 14,202 15,676 33,391 9,403 monitored years 1,022,369 66,007 269,090 170,927 3 167,971 272,583 83,035 avg. monitoring 10.0 6.9 9.5 12.0 10.7 8.2 8.8 duration (y) No. (%) by dose in mSv: 0 ≤ H<10 76,165 (74.9) 6,153 (63.9) 21,529 (76.2) 9,234 (65.0) 13,566 (86.5) 26,300 (78.8) 8,060 (85.7) 10≤ H<20 7,790 (7.7) 920 (9.6) 2,832 (8.4) 1,082 (7.6) 816 (5.2) 2,219 (6.6) 462 (4.9) 20≤ H<50 8,192 (8.1) 1,201 (12.5) 2,025 (7.2) 1,423 (10.0) 672 (4.3) 2,431 (7.3) 490 (5.2) 50≤ H<100 4,636 (4.6) 707 (7.3) 1,103 (3.9) 973 (6.9) 309 (2.0) 1,315 (3.9) 229 (2.4) 100≤H<200 3,148 (3.1) 422 (4.4) 686 (2.4) 968 (6.8) 185 (1.2) 742 (2.2) 123 (1.3) 200≤H<400 1,599 (1.6) 215 (2.2) 419 (1.5) 514 (3.6) 101 (0.6) 309 (0.9) 32 (0.3) H≥400 220 (0.2) 7 (0.1) 97 (0.3) 8 (0.1) 27 (0.2) 75 (0.2) 7 (0.1) Mean dose (mSv) 17.8 24.23 16.9 29.8 8.8 13.8 8.2 collective dose 1,811.0 233.2 478.3 422.7 137.8 461.90 77 (person-Sv) 1HAN=Hanford site; INL= Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; PNS=Portsmouth Naval Shipyard; SRS=Savannah River Site

2Of workers by primary site, the number of workers with at least one site dose record for penetrating radiation.

3The SRS Fayerweather file did not discriminate between monitored and unmonitored years, thus the number of monitoring person-years were estimated from positive doses only.

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About 8.2% of the study subjects had dosimetry information from more than one facility (Table

5-5). The largest combination of sites was INL and Hanford, where 3,916 workers (3.8%) were exposed at both facilities. In contrast, less than 0.3% of PNS workers were exposed at another facility.

Table 5-5. Combinations of sites with exposure information on 10 or more workers in the base cohort.

Site combination No. workers Percent of total cohort INL 28,754 28.3 HAN 21,776 21.4 ORNL 13,012 12.8 SRS 12,071 11.9 PNS 9,600 9.4 LANL/Zia 8,195 8.1 INL + HAN 3,916 3.8 HAN + SRS + ORNL 1,214 1.2 HAN +ORNL 863 0.8 HAN + LANL/Zia 822 0.8 HAN + SRS 492 0.5 INL + ORNL 150 0.1 ORN +LANL/Zia 127 0.1 INL + SRS 114 0.1 INL + LANL 110 0.1 INL +HAN +SRS 100 0.1 INL + HAN + ORNL 99 0.1 SRS + ORNL 90 0.1 INL + HAN + SRS + ORNL 65 0.1 INL + HAN + LANL/Zia 56 0.1 HAN + ORNL + LANL/Zia 35 0.0 SRS + LANL/Zia 32 0.0 HAN + PNS 13 0.0 HAN + SRS + LANL 10 0.0 1HAN=Hanford site; INL= Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; PNS=Portsmouth Naval Shipyard; SRS=Savannah River Site

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Temporal trends in exposure are consistent with weapons production schedules and changes in radiation protection standards (Figure 5-1). Beginning in 1951, average exposures steadily increased with the number of persons monitored to peak values in the 1960s, when plutonium production (SRS,

Hanford), submarine overall (PNS), and reactor testing activities (INL) were at highest levels. The highest annual average exposure (5.1 mSv) occurred in 1965. An exponential decrease in annual average doses was observed in the period between 1965 and 2000, which coincides with a reduction in plutonium production at SRS and Hanford and complex-wide improvements to radiation protection including a reduction in dose limits and the implementation of ALARA principles.

6 avg. dose no. monitored 40,000

5 30,000 4

3 20,000 workers Dose (mSv) Dose 2 10,000 1

0 0 1950 1960 1970 1980 1990 2000 Year

Figure 5-1. Average annual dose and number of monitored workers in the base cohort.

5.2 Other occupational exposures

Employment in nuclear facilities other than the study facilities was likely for some participants; therefore, a potential existed to underestimate occupational exposures to workers who were employed in nuclear facilities elsewhere. To reduce this bias, the search for relevant exposure data was expanded to include information from other facilities. The informational sources searched were comprised of

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electronic exposure records that were collected or constructed for previous NIOSH studies and two national dose registry sources (i.e., REMS and REIRS).

The REMS database contained annual individual radiation exposure records for all DOE workers employed since 1987 and for some workers dating back to 1969 ( ∼ 2.4 M person-years). The REIRS database contains exposure information for workers employed at certain USNRC licensed facilities since

1969 ( ∼ 1.0 M person-years). REIRS data are available for individuals by licensee employment period; therefore, results may span several years. Participating USNRC program facilities include: commercial nuclear power reactors; industrial radiographers; fuel processors, fabricators, and reprocessors; manufacturers and distributors of byproduct material; and independent spent fuel storage installations

[Anzenberg et al. 2010].

5.2.1 Methods

The study roster of 1,816 individual cases and controls was linked to exposure data in each source file by SSN, name, and DOB. Partial matches were examined to determine appropriate matches.

Similar to information obtained from the primary facilities, exposure data were abstracted for the individual for the time period from first exposure to the date of the attained age of the respective case

(i.e., cutoff date). Exposure assessors were blinded to case status. Abstracted data were adjusted to account for overlapping information. A hierarchy was used to select information in overlapping time periods whereby primary facility records were preferred to exposure records from other facilities, which were preferred to records from the REMS and REIRS databases. To facilitate exposure lags, dose was assumed to be evenly distributed over the exposure interval. Annual dose estimates were similarly constructed for dose intervals spanning multiple years.

Detailed information on dosimetry practices was not available to derive bias factors for estimating bone marrow absorbed dose. Therefore, reported values of dose equivalent from low-LET

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exposures were adjusted to absorbed dose to bone marrow using the set of bias factors derived for

ORNL exposures described in Table 5-3. ORNL factors were selected as the default because the site was a leader in the development of early personal dosimeters and many facilities used ORNL-designed dosimetry equipment and monitoring methods. Likewise, values of absorbed dose from neutrons were calculated using the dose conversion coefficients derived for other/unknown Hanford workers. These coefficients were selected because the information used in derivation was most complete for Hanford workers and believed most germane to exposure scenarios at other facilities.

5.2.1.1 NIOSH files

The NIOSH system of records was searched for exposure files containing personal identifiers that could be linked to individual study subjects. These files originated from previous efforts to collect exposure information for health studies in the Division of Surveillance, Hazard Evaluations, and Field

Studies (DSHEFS) and for the examination of coworker patterns in support of dose reconstruction performed by the Division of Compensation Analysis and Support (DCAS). All files searched are identified in Table 5-1.

Some site dosimetry systems included records of previous exposures at other facilities as a means to track lifetime dose for compliance purposes. These files were used, to the extent practical, to supplement the information on exposures to study subjects. In particular, dosimetry files obtained from the Hanford Site contained highly specific information on exposures during previous employment at other facilities. INL dosimetry records also identified “offsite” doses, but information on location and exposure period were inadequate to determine potential overlap with other files.

5.2.1.2 REMS and REIRS

Data from the REMS and REIRS databases provided information on exposures at other facilities.

The REMS database also provided exposure information for subjects employed at a primary study facility

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after the end date of other available exposure records. The REMS and REIRS databases originated from a single system of AEC records that are now maintained by one contractor; therefore, the databases share many similarities in design and availability. Permissions were sought from the individual agencies to abstract exposure data from both databases. The roster of cases and controls was provided to the database manager to conduct the search and data abstraction. Two searches were conducted. The first search was completed using an initial roster of 1,588 study participants. A second search involving 240 subjects was conducted because of revisions to the roster following new vital status information.

5.2.2 Results

There were 9,837 and 797 external dose observations abstracted from REMS and REIRS, respectively. After removal of duplicate information and adjustment to annual dose intervals, final doses were accumulated for 277 study participants resulting in 2,028 person-years of exposure. Bone marrow dose from penetrating low-LET radiation was 5.4 person-Gy. Between these two sources, nearly all

(98.6%) of the dose was attributed to the REIRS abstraction. The mean cumulative dose to exposed study subjects identified in REMS and REIRS combined was 19.5 mGy (range 0-395 mGy).

Data from the REMS/REIRS abstraction were combined with information from other available exposure files for a total of result in 3,197 dose observations pertaining to employment outside of the study facilities (Table 5-6). A total of 6.0 person-Gy from penetrating low-LET radiation was distributed

among 329 study subjects with exposures spanning the time period between 1936 and 2005. Many of

these workers were exposed at multiple facilities in addition to their exposures in study facilities. Nearly

all (90%) of the offsite collective dose was attributed to work in the nuclear power industry; there were

229 workers (12.6%) with at least one dose assignment from a commercial nuclear power station.

Employment at other DOE facilities ( n=97) resulted in about 0.4 person-Gy. The remaining dose (0.2

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person-Gy) was attributed to workers employed at Naval Reactors (NR) and other military institutions

(n=21), universities ( n=5), and unknown locations (n=34).

Neutron doses were also reported by some facilities but the collective dose was small in comparison to gamma radiation exposures. The reported collective equivalent dose from neutrons was

8.8 person-mSv. The maximum cumulative dose equivalent from neutrons was 2.9 mSv, which was assigned to a plutonium worker employed at the former Mound Plant (MMP) in Miamisburg, OH from

1972 to 1979.

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Table 5-6. Cumulative doses (mGy) from penetrating low-LET radiation exposures at other facilities

No. Facility monitored mean median minimum maximum total ANL 6 6.8 0.6 0.0 38.3 40.8 BNL 2 8.7 8.7 0.0 8.7 17.4 FMPC 4 0.0 0.0 0.0 0.0 0.0 ORGDP 16 0. 9 0. 4 0.0 1.8 8.0 LBL 2 3.7 3.7 0. 2 7.3 7. 5 LLNL 3 0. 1 0.0 0.0 0. 3 0. 3 MILITARY 5 16.3 14.6 0.0 50.0 81.5 MMP 12 0.1 0.0 0.0 1.0 1.2 NR 16 1.0 0. 7 0.0 3.9 15.2 NTS 9 14. 4 0.0 0.0 85.9 129.3 USNRC L 229 23. 6 2. 4 0.0 394.7 5,405 RF P 9 0. 4 0.1 0.00 0.5 1.8 SNL 1 0.2 0.2 0.2 0.2 0.2 UMTRA 1 0.4 0.4 0.4 0.4 0.4 UNIV 5 2. 4 1.6 0.0 5.2 11.8 WVDP 1 0.2 0.2 0.2 0.2 0.2 Y12 32 5.3 0.4 0.0 66. 7 169.6 other or unknown 33 4. 1 0. 1 0.0 29.6 13 6.0 all facilities combined 339 17.9 1. 1 0.0 39 5 6,033 ANL= Argonne National Laboratory, IL; BNL=Brookhaven National Laboratory, NY; FMPC=Feed Materials Production Center, OH; LBL= Lawrence Berkeley Laboratory, CA; LLNL=Lawrence Livermore National Laboratory, CA; MILITARY= U.S. Armed Forces; MMP= Miamisburg Mound Plant, OH; NR=U.S. Naval Reactors; NTS=Nevada Test Site, NV; USNRCL=USNRC licensees; ORDGP=Oak Ridge Gaseous Diffusion Plant, Oak Ridge, TN; RFP=Rocky Flats Plant, CO; SNL= Sandia National Laboratory, NM; UMTRA=Uranium Mill Tailings Remedial Action, NM; UNIV=U.S. Universities; WVDP= West Valley Demonstration Project, NY; and Y12=Y12 Plant, Oak Ridge, TN

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5.2.3 Limitations

Because of limitations in data availability, it was not feasible to construct a complete exposure history for each study subject. Exposure data files were not available for all affected sites and during all

periods of potential exposure. Of files searched, some were incomplete because of restrictions imposed

by the inclusion criteria developed for the previous study. Other files lacked exposure data in certain

periods because of incomplete monitoring practices. Of important note, personal monitoring was

performed to verify compliance with protection standards in place at the time of exposure. Personal

monitoring was often restricted to persons who were deemed likely to exceed imposed limits. Thus the

completeness, accuracy, and precision of personal dosimetry systems varied with time and with the

perception of hazards in each facility.

A centralized registry for U.S. nuclear workers that effectively records an individual’s career

radiation dose is not available. Although the REMS and REIRS data on DOE and USNRC workers covers a

large fraction of the U.S. nuclear workforce employed since the 1980s, other occupational exposures

during U.S. military service or work in other professions not reporting to these systems were likely

among some study subjects. REMS and REIRS are limited to information collected primarily over the past

two decades when effective exposure controls and dose reporting systems were largely in place. There

is little information on early exposures in these systems. For example: although dose reporting to REIRS

began in 1969, the requirement to report annual doses for active employees was not fully implemented

until 1994. Doses reporting prior to 1994 stemmed from the initial requirement to report radiation dose

histories of workers only at the time of termination. Therefore, dose information from 1969 to 1994

may be incomplete for those workers actively employed during those years. In attempts to fill these

gaps, the USNRC requested facilities to voluntarily report career doses to active employees. However, a

subsequent review of the database conducted by Muirhead et al. [1996] suggested that the REIRS

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database substantially underestimates doses of workers prior to 1994 due to the transition in reporting requirements between 1969 and 1994.

These sources of uncertainty are likely to result in underestimating exposures to some study subjects, especially during the early years of radiation work. Nevertheless, dosimetry information was available from many of the large national laboratories (e.g., SRS, Hanford, ORNL, LANL, and INL) employing most weapons-workers during the height of nuclear weapons production and accounting for the bulk of exposures to ionizing radiation to study subjects. There was little evidence of a large contribution of dose to the study group from employment in other facilities. Therefore, a significant bias in exposure estimates resulting from employment elsewhere is not likely.

5.3 Work-related medical X-ray exposure

5.3.1 Introduction

Routine physical examinations including chest radiography were customary as a condition of employment or continued qualification as a radiation worker in all study facilities. Although current radiographic posterior-to-anterior (PA) chest x-ray examinations result in negligible bone-marrow dose

(∼0.05 mGy∙exam -1; [Preston-Martin and Pogoda 2003]), early examinations using stereo image photofluorography (also known as photoroentgenography or mass miniature radiography) could have resulted in doses over twenty-fold higher [Anderson and Daniels 2006; Cardarelli et al. 2002]. Some study sites also required that employees undergo a full or partial lumbar radiographic series to establish a baseline for employment, which may have resulted in bone marrow doses between 0.1 and 4.1 mGy∙exam -1 [Preston-Martin and Pogoda 2003]. Unlike ionizing radiation exposure from other medical sources, whereby dose from diagnostic procedures is likely to be nondifferentially distributed in workers, a strong correlation between workplace ionizing radiation exposure and work-related x-ray exams (WRX) is likely. Thus, estimates of bone marrow dose from early WRX of the chest using

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photofluorographic techniques (PFGs) and lumbar series were calculated for each subject in the case- control study.

5.3.2 Methods

Organ dose coefficients for early medical x-ray examinations of nuclear workers were previously estimated for epidemiologic studies [Anderson and Daniels 2006; Cardarelli et al. 2002; Daniels et al.

2005] and for dose reconstruction supporting worker compensation programs [Burns and Fleming 2007;

Johnson et al. 2010; Rohrig 2009; Thomas 2009; Thomas et al. 2010]. In general, these studies use historical information on equipment and procedures to model the radiation field characteristics that are used to calculate dose to a specified tissue per exam (e.g., mGy∙exam -1). Many of the studies examined

x-ray procedures that were performed at the primary study facilities. For example, Anderson and

Daniels [2006] derived coefficients for estimating the absorbed dose to hematopoietic tissue in workers

at Hanford, LANL, ORNL, PNS, and SRS for studies on leukemia risk by Schubauer-Berigan et al. [2007a;

2007b]. Given their direct relation to the current effort, the results reported by Anderson and Daniels

[2006] were used as the primary informational source for WRX dose estimates in the current study.

Supplemental information was provided by research supporting dose reconstruction under the Energy

Employees Occupational Illness Compensation Act of 2000 (hereafter referred to as EEOICPA). Where

results between Anderson and Daniels [2006] and EEOICPA studies were in conflict, the difference was

reconciled based on the weight of evidence presented. However, cautious interpretation of information

from EEOICPA studies was necessary because of the claimant-friendly policy to select assumptions that

tend to overestimate dose.

Anderson and Daniels [2006] calculated facility, time, and procedure-specific dose coefficients

(mGy∙exam -1) using the software package PCXMC (Personal Computer program for X-ray Monte Carlo) developed by the Finnish Centre for Radiation and Nuclear Safety [Servomaa and Tapiovaara 1998].

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PCXMC uses a Monte Carlo simulation approach based on the stochastic transport of x-rays as they interact and deposit energy in tissue, to estimate organ doses from diagnostic x-ray examinations. These site-specific dose coefficients were multiplied by the number of exams within the year of exposure to estimate the annual absorbed dose to the bone marrow from WRX. Historical records were searched to determine exam frequencies, type of equipment used, and equipment operating parameters.

PCXMC-derived organ dose coefficients were calculated using estimates of entrance air kerma

(mGy), entrance exposure (C∙kg -1), or the product of the beam current and time (mAs). Required input

parameters included beam size (height and width) on the patient, focus-to-skin distance (FSD),

coordinates of the beam on the patient, angle of the projection, kVp, target angle, and quantity and type

of added filtration. The height and weight of subjects were unknown; therefore, values for the standard adult phantom were used (174 cm; 71.1 kg). Photon spectra for transport modeling were based on full- wave bridge-rectified PFG equipment by adjustment via a separate algorithm to account for the voltage ripple expected from the full-wave rectified equipment compared to the three-phase constant potential x-ray machines used PCXMC models [Nestle 2005].

Examination frequencies for all affected sites were estimated using available exam procedures, work histories, and individual medical records. For annual exams at SRS and Hanford, where no other exam date was specified, it was assumed that the exam took place on the anniversary of the hire date.

For PNS, annual exams were assumed to occur on the worker’s birthday. Although routine medical examinations were conducted at INL, there was no evidence that PFG chest x-ray examinations or lumbar series were ever conducted; thus, WRX doses were not estimated for INL workers [Rohrig 2009].

The dose coefficients and exam frequencies for all other study facilities are shown in Table 5-7. Dose coefficients from Daniels and Anderson (2006) were preferred over EEOICPA dose reconstruction coefficients except for SRS workers. Excluding SRS, dose coefficients from the two sources were in

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reasonable agreement with a tendency toward larger values used in EEOICPA studies. A brief description

of the assumptions used to estimate exam frequencies is provided in the following subsections.

5.3.2.1 Hanford

Hanford worker examinations prior to 1956 were performed at the Kadlec Hospital in Richland,

WA. Although specific techniques during this period were not available, radiation dose measurements

from the early equipment were conducted in 1959 for the retrospective dose reconstruction supporting

the DOE HMS [Kirklin et al. 1969]. These measurements were used as input to models used to calculate

dose coefficients.

Stereoscopic PFG, single PFG, and direct radiography of the chest were conducted at the

Hanford Site from 1944 to 1962. Based on previous reviews of individual medical records, simple PA

chest exams using stereoscopic PFG methods were most common prior to 1953, followed by single PFG

and direct radiography thereafter. During these times, employees received pre-employment, annual,

and termination chest exams. There was no evidence of routine lumbar series conducted at Hanford

[Anderson and Daniels 2006].

5.3.2.2 LANL

Routine chest examinations prior to 1957 were performed by the occupational medical service

using both standard radiography and PFG methods; however, monthly progress reports suggested that

most examinations were PFG [Johnson et al. 2010]. The actual date of first PFG use at LANL is uncertain;

although pre-employment chest exams appear as early as 1943 [Johnson et al. 2010], evidence of PFG

use prior to 1948 was equivocal. Therefore, only those radiation workers employed between March 16,

1948 and December 31, 1956, were assumed to receive a pre-employment a PFG [Anderson and Daniels

2006]. During this time, workers with evidence of potential exposures to uranium, plutonium, polonium,

or beryllium were assumed to have received exams at six month intervals. In the absence of evidence

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indicating the potential for internal exposures (e.g., plutonium workers designation, bioassay results, medical record), the exam period was extended to once per year. [Anderson and Daniels 2006].

The conclusion by Anderson and Daniels [2006] that early LANL PFGs were stereographic was based on the assumption that examinations were similar to those performed at Hanford and ORNL. In contrast, additional information reported by Johnson et al. [2010] was judged sufficient to conclude that stereographic imaging was not routinely performed at LANL. Therefore, the dose coefficient used to estimate WRX dose to LANL workers was reduced by a factor of two compared to that reported earlier

[Anderson and Daniels 2006].

5.3.2.3 ORNL

All radiographic examinations for ORNL workers prior to October 3, 1947, were conducted by staff at the Oak Ridge Hospital in Oak Ridge, TN. Detailed reviews of films and medical records during this period suggests that most pre-employment examinations were stereoscopic PFGs [Anderson and

Daniels 2006; Burns and Fleming 2007]. In addition, there was evidence that pre-employment examinations for craft employees (pipefitters, carpenters, etc.) included a lumbar spine series to uncover pre-existing back problems. These lumbar series took place from April 6, 1950, to September

23, 1953.

Subjects who began work at the facility prior to October 1945 were assigned dose from a single pre-employment stereoscopic PFG, except those hired from October 1, 1943 to February 1, 1944, and

October 1, 1944 to January 1, 1945. ORNL used conventional direct radiography equipment for beginning October 3, 1947. Craftsmen who started work from April 6, 1950 to September 23, 1953, received a pre-employment lumbar spine series [Anderson and Daniels 2006]. Job titles were used to select crafts positions based on broad categories of boilermaker, electrical, general labor, maintenance, mechanical, millwright, metal work, pipefitting, and welding. Because the Oak Ridge facilities shared

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early medical services, it was further assumed that workers employed at K25, Y12, or with the

Tennessee Eastman Corporation (TEC) were subject to the same pre-employment screening.

5.3.2.4 SRS

The use of photofluorographic techniques at SRS is uncertain. Some evidence suggests that routine mass tuberculosis screenings of the site workforce using mobile PFG units may have taken place as early as 1951 [Thomas 2009]; however, data were insufficient to determine who was screened and how often screening occurred. In or around February 1954, SRS procured a mobile x-ray unit that was capable of mass chest screening by PFG and conventional techniques. It appears that annual stereoscopic PFGs were conducted from 1954 to 1957 for workers assigned in remote areas of the plant

(i.e., areas D, C, K, L, P, R, H, and F). It is likely that workers stationed near the medical facility in Building

719-A would not routinely receive PFG examinations but would be examined using direct radiography techniques [Anderson and Daniels 2006]. Nevertheless, work history information was not sufficiently detailed to identify workers by specific station and facility records were insufficient to estimate the workload of x-ray equipment used during that time. Moreover, a review of the medical records for some study subjects employed at SRS from 1954 through 1957 did not elucidate patterns of PFG use. Because the largest fraction of workers would have been stationed in the remote areas compared to Building

719-A, we assumed that all workers employed from 1954 through 1957 received stereographic PFG exams by the mobile x-ray unit during this period.

The SRS WRX dose coefficient reported by Thomas [2009] was nearly two-fold less than that reported by Anderson and Daniels [2006]. In the latter study, the coefficient was based on the early equipment and techniques used at Hanford and ORNL because information regarding these parameters was not available. In the study by Thomas [2009], the coefficient was based on equipment and operating parameters specified in a response to the request for proposals for purchasing the mobile x-ray unit. The

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dose coefficient reported by Thomas [2009] is consistent with values reported by others regarding PFGs

during the same period [Laughlin et al. 1957]; therefore, [Antoku and Russell 1971]dose coefficient was

used to estimate WRX dose for SRS workers.

5.3.2.5 PNS

PNS conducted PA chest exams using non-stereographic PFG prior to May 1, 1966. Examination

frequencies were estimated assuming: 1) workers hired from January 1, 1948, through December 31,

1966, received at least one exam in the year of hire; 2) workers employed anytime during 1948, 1951,

1952, and/or 1956 received an annual exams for mass tuberculosis screening; 3) radiation workers

employed from January 1, 1958 through December 31, 1962 received a baseline exam; and 4) Radiation

workers employed from January 1, 1963 to May 1, 1966 were examined each year in which at least one

occupational dose record is reported. For the year 1966, only those workers with birth months earlier

than May were assigned dose because annual physicals coincided with birthdates [Daniels et al. 2005].

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Table 5-7: Calculated dose coefficient (bone marrow, mGy∙exam -1) for PFG and lumbar spine examinations at study facilities.

Anderson Exam and Daniels Projection Frequency EEOICPA 1 (2006) Date Range Comments Hanford (PFG, stereo) Pre-emp, 2.8 2.2 June 1, 1944- annual, May 31, 1953 termination Hanford (PFG, single) Pre-emp, 1.4 1.1 June 1, 1953- annual, December 31, termination 1962 ORNL (PFG, stereo) Pre-emp 2.4 2.0 1943-October Except those hired between October 1, 1943 and February 3, 1947 1, 1944, and October 1, 1944 and January 1, 1945. Includes K25 and Y12 workers. ORNL (Lumbar Series) Pre-emp 1.5 1.4 April 6, 1950- Limited to: construction workers, pipefitters, welders, September 23, painters, laborers, carpenters, millwrights, helpers, 1953 electricians, mechanics, and craft foremen. SRS (PFG, stereo) Annual 0.92 2.2 1954-1957 Information on SRS PFG use is incomplete. This use of stereo imaging is uncertain. Current assumptions may overestimate exposures to persons, especially those working in areas other than D, C, K, L, P, R, H, and F. LANL (PFG, single 2) Pre-emp, varied 1.3 1.1 March 16, Exam periodicity: every 6 mo. for those exposed to uranium, 1948- plutonium, polonium, or beryllium; annually for those with December 31, external exposure only. 1956 PNS (PFG, single) Pre-emp, varied N/A 1.5 January 1, Workers employed in 1948, 1951, 1952, and/or 1956 1948-May 1, received annual PFG; Radiation workers employed between 1966 January 1, 1958 and December 31, 1962 received a baseline PFG; Radiation workers employed between January 1, 1963 and May 1, 1966 received annual PFG PFG= photofluorographic chest x-ray; EEOICPA= Energy Employee Occupational Illness Compensation Act of 2000. 1Dose conversion factors reported in technical basis documents supporting EEOICPA dose reconstruction activities [Burns and Fleming 2007; Johnson et al. 2010; Thomas 2009; Thomas et al. 2010]. 2Single PFG images were conducted at LANL based on information in Johnson et al. [2010].

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5.3.3 Results

Cumulative doses from WRX were estimated for 883 study participants (48.6%) resulting in about 5.7 person-Gy of exposure (Table 5-8). Among those exposed, the overall mean and median cumulative doses from WRX were 6.4 mGy and 4.4 mGy, respectively, and individual cumulative doses ranged from 0.9 mGy to 28.6 mGy. Most (44.2%) of the collective dose was attributed to PNS. In contrast, relatively few ORNL workers were exposed ( n=34). Two subjects were exposed at more than one facility.

Table 5-8. Cumulative dose (mGy) to study participants from work-related x-ray examinations

ALL Hanford LANL ORNL PNS SRS OTHER N 88 6 267 63 34 255 25 7 10 Mean 6.4 7.6 3.0 1. 8 9.9 3.4 1. 8 Median 4.4 6.6 2.2 2.0 10.5 3. 7 2.0 Range 0.9 -28.6 2.2 -28.6 1.1 -13.2 1.4 -3.4 1.5 -13.5 0.9 -3.7 1.4 -2.0 Sum (person -Gy) 5.7 2. 0 0.1 9 0. 06 2. 6 0. 87 0.0 2

5.3.4 Limitations

The WRX dose estimates are relatively imprecise compared to other estimates of low-LET exposures. This imprecision is caused by a number of assumptions on modeling parameters that remain largely uncertain. First, details on the equipment and procedures used were not available in most cases.

When equipment was specified, actual measurement data needed to characterize entrance exposures were sparse. Second, exam frequencies were determined by work histories that were incomplete for some study subjects or lacked the specificity necessary to accurately predict the occurrence of periodic examinations. Lastly, worker doses will also vary based on patient physiology, technician work practices, rate of retakes, and other factors that may influence exam frequency (e.g., availability, worker classifications, and medical priorities), whereas models used in this study were based on the characteristics of a reference male worker receiving scheduled exams without requiring reexamination.

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5.4 Neutron Exposure Assessment

Some study subjects may have received an important contribution to their total occupational dose from neutron exposures over the course of their employment in reactor facilities, plutonium production, and accelerator operations. It is common for persons with significant neutron dose to have concomitant dose from low-LET radiation. Thus, it is conceivable that a disregard of neutron exposures could introduce a bias in the observed dose-response between low-LET exposure and leukemia.

Therefore, quantification of neutron exposures to individual subjects was conducted using available monitoring data and the methods described in this section.

5.4.1 Facility dosimetry

Estimation of neutron dose using measurements obtained from passive personal neutron dosimeters has been a long-standing technical challenge. Uncertainty in the measurement of neutron dose is substantially greater than that from measurements of gamma and x-ray radiation. Although several personal dosimeter types have been used over the years, each with its own strengths and weaknesses, the most common dosimeters used by study participants were: 1) Pocket Ionization

Chambers (PICs) mainly in early accelerator research in MED facilities, 2) Kodak Nuclear Track Type A

(NTA) film emulsion that was widely used between 1950 and the early 1970s, and 3) thermoluminescent neutron dosimeters (TLNDs) that were placed into service in the mid-to-late 1970s and thereafter. The earlier devices provided information on exposures to intermediate and fast neutrons typified by unmoderated fission spectra or common (α, n) reactions. Slow and thermal neutron personal exposure measurements were sparse in most situations until the widespread use of albedo TLND neutron dosimetry. A possible exception is LANL, where multi-element film dosimetry was developed for routine quantification and reporting of slow and thermal neutron exposures.

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Pocket ionization chambers (PICs) with enriched boron ( 10 B) liners were used for early neutron

monitoring in MED facilities. These devices were sensitive to gamma radiation in addition to slow

neutron energies, without discrimination. PIC usefulness was limited to detecting the potential for

neutron dose, but was inadequate for dose quantification. PICs were issued on a case basis and were

processed daily; however, measurement records were not retained at all facilities or for all monitoring

periods. Exposures to most study subjects during this time are not likely because participation was

limited to those workers first hired at LANL, Hanford, or ORNL after 1950. Nevertheless, exposures

during employment at other early accelerator research facilities cannot be ruled out for some study

subjects.

Beginning in the early 1950s, the AEC realized that some workers had a significant potential for

fast neutron (i.e., neutron energies on the order of 10 keV to 10 MeV) exposures in certain work

locations, which prompted the nearly universal use of fine grain photographic emulsions for measuring

the neutron dose to workers occupying exposure prone areas. The Kodak Nuclear Track Type A (NTA)

film packet was the preferred choice for neutron dosimeters, which contained dental film with a single emulsion layer ( ∼30 µm thick) resting on a cellulose triacetate base (203 µm thick) [Kathren et al. 1965].

Proton recoil following fast neutron interaction with low atomic number nuclides in the film emulsion formed “tracks” of silver halide gains that were microscopically observed and mathematically related to dose. Monoenergetic neutrons will produce recoil protons with a uniform energy distribution from null to the energy of the incident neutron (dependent on the angle of incidence). However, tracks from protons of 500 keV or less are not distinguishable in NTA film and poor resolution of tracks was evident at energies less than 800 keV. Nevertheless, NTA film provided a relatively reliable method of quantifying doses from fast neutrons of the fission spectrum (i.e., most neutron energies between 1 and

3.0 MeV) [Cheka 1954; Höfert and Piesch 1985]. Additionally, the film emulsion was sensitive to thermal neutron radiation by the recoil proton from the reaction 14 N(n,p) 14 C. Later uses of NTA film incorporated

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thermal neutron absorbers (e.g., natural cadmium, resonance peak-energy= 0.178 eV, cross-section

7,800 barns) and used the difference in the number of tracks between unfiltered and filtered portions of the film to quantify exposures to thermal neutrons [Thornton et al. 1961].

NTA films were typically integrated into the personal badges and worn by the worker along with their low-LET film dosimetry. Processing of NTA film was conducted for those workers suspected of having a viable potential for significant neutron dose. NTA film dosimeters share many of the sources of uncertainty associated with beta-gamma film dosimetry, such as angular dependency, variations in emulsions, latent image fading, backscatter, and energy dependence [Cheka 1954; Höfert and Piesch

1985; Kathren et al. 1965; Oshino 1973]. The overall relative uncertainty in the estimate of a 1.0 mSv dose from PuBe neutrons [ 9Be(α, n) 12 C] is about 30%. The recording threshold for fast neutrons varied among sites and over time. The estimated limit of detection for most sites was approximately 0.14 to 0.4 mSv and doses were typically quantified in increments >0.2 mSv [Höfert and Piesch 1985].

NTA films were subsequently replaced by TLNDs beginning in the early 1970s. TLNDs detect neutron radiation using a system of phosphors to absorb thermal or intermediate energy neutrons.

TLNDs often use more than one method of neutron radiation detection and often incorporate non- penetrating and penetrating photon dose capabilities. Comparisons between the NTA- and TLND- measured neutron dose in several workplaces showed that the TLND-measured doses were typically greater, suggesting that NTA-derived doses were largely underestimated. Consistent with beta-gamma dosimetry used in DOE sites, widespread compliance with DOELAP neutron dosimetry performance criteria resulted in standardization of dosimeter calibration and response between facilities. DOE study

facilities received DOELAP accreditation in the late 1980s. A similar Navy-wide accreditation process also

effectively standardized PNS neutron dosimetry.

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The general types of neutron dosimeters and periods of use for each of the study facilities are shown in Table 5-9. Details on dosimetry procedures, equipment, and calibration techniques, are provided in the following subsections.

Table 5-9. General characteristic of common neutron dosimeters.

Dosimeter Type Parameter PIC ( 10 B) NTA TLND Era: Prior to 1950 1950s - 1970s l970s to date 252 Calibration: Thermal neutrons PoBe, PuBe, accelerator, PuF 4, Cf (bare and moderated) PuF 4 PuBe Response: Thermal,<0.4eV Limited by size of Response of Ag, Cd MDL ~0.01 mSv PIC chamber activation but likely of limited sensitivity in mixed neutron/photon fields. Intermediate Minimal response No response < 700-800 keV MDL ~0.50 mSv

Fast, > 1 MeV Essentially no MDL~ 0.1 mSv MDL ~0.3 mSv response PIC= Pocket Ionization Chamber, NTA= NTA Nuclear Track Emulsion, TLND= Thermoluminescent Neutron Dosimeter, MDL= minimum detection level

5.4.1.1 Hanford

A few positive neutron doses were found in the REX database dating back to 1946, although others have reported that recording of neutron dose measurements began with NTA film in the early

1950s [Fix et al. 1997; Watson 1951]. Prior to then, some workers used PICs to detect slow neutron exposure without dose quantification. NTA film was initially housed in the Hanford two-element beta- gamma dosimeter holder until an NTA-specific holder was assigned to select workers in 1958 [Swanberg

1958]. From 1950 to 1958, NTA films were processed concurrently with beta-gamma personal dosimetry. During this period, 66,290 dosimeter films were processed, of which only 107 (0.16%) resulted in estimated doses equal to or greater than 0.9 mSv [Watson 1959]. Of the positive films, all but six were collected from workers in the 200 West and 300 Areas. Furthermore, all incidences of positive neutron exposure (i.e., ≥0.9 mSv) had corresponding positive gamma exposure (i.e., ≥0.3mSv). Based on

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these observations, film processing and dose assessment was primarily limited to workers in the 200

West plutonium facilities beginning in 1959. Films from workers in other areas were processed on a case basis if a credible neutron exposure potential existed and the photon doses exceeded 1.0 mSv [Fix et al.

1997]. Hanford began using TLNDs in 1972.

Over time, NTA film was calibrated using a PoBe or PoB (normalized to PoBe) source (1950-

1955), Van de Graaff accelerator with beryllium target (1955-1958), and a PuF 4 neutron source (1958-

1972) [Fix et al. 1997]. The PuF 4 was also used to calibrate the TLND in a specially arranged configuration

to match the laboratory and workplace dosimeter response for the same measured dose using tissue

equivalent proportional counter (TEPC) measurements. Subsequently, a 252 Cf source was used (1980-

2005). Hanford initially calculated and reported a single collision RBE dose, which was defined as the dose, adjusted for relative biological effectiveness of the incident neutrons, that resulted from the first collision of a neutron with an atom in the subject material [Attix 1966]. A value of 10 was assumed as the RBE value (i.e., the “Quality Factor”) for fast neutrons [NBS 1957]. Hanford began assessing multiple- collision RBE dose beginning in July of 1963. The multiple-collision dose recognized that a neutron incident to human tissue was likely to undergo more than one interaction before dissipating all of its kinetic energy. The difference in recorded dose between the two calibration references was an increase in dose of about 37% using the multiple-collision RBE assumption.

5.4.1.2 INL

INL began personal neutron monitoring with NTA film incorporated into the ORNL-type two- element film badge placed into service on August 20, 1951 [Cipperley 1958b]. Subsequent film badge designs also included Kodak Type A film packets for personal neutron monitoring [Cipperley 1966; INL

1960]. Descriptions of early monitoring practices report a detection capability of about 0.14-0.20 mSv.

Although the capability of neutron monitoring was evident, it appears that not all dosimeters were

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equipped with NTA and not all NTA film was examined [Sommers 1969]. INL began issuing TLDs to employees who routinely received yearly exposures <5.0 mSv in 1966 [Cipperley 1966]. The early TLD design did not have fast neutron measurement capability; thus, NTA film was assigned to workers by

Health Physics staff on a case basis [Sommers 1967]. By 1975, INL dosimetry included the albedo TLND dosimetry system, as described by Hankins [1973], for quantifying neutron exposures [Gesell et al.

1996]. The albedo dosimeters attached easily to existing dosimeters and were particularly useful for quantifying doses from low energy or scattered neutrons. However, the dosimeter response was significantly energy-dependent (primarily in the range 0.1-10 MeV) and spectral/area-specific adjustments were necessary to quantify neutron doses. Adjustments were made using Facility Neutron

Conversion Factors (FNCFs), which ranged from 0.17 to 2.7.

Dosimeter exchange frequencies varied by neutron exposure potential and by dosimeter type. A review of historical exposure records revealed that NTA film monitoring intervals were conducted weekly, biweekly, and monthly. Similarly, TLD exchanges were conducted biweekly, monthly, and quarterly.

NTA film calibration was conducted using a PoBe neutron source. A 252 Cf source was used for

TLND calibrations. DOELAP accreditation of the TLND dosimetry system was first accomplished in 1988

[Gesell et al. 1996].

Most INL workers were not routinely exposed to neutrons; however, some neutron fields were evident in a few work areas and personal dosimetry was conducted for individuals assigned to these areas. In particular, INL operated three reactors [MTR (1952-1970), TREAT (1952-1994), and ZPPR (1969-

1992)] that included neutron beam ports for irradiating test samples.

The Material Test Reactor (MTR) was one of eight reactors operated in the Test Reactor Area

(TRA), and was a water-cooled and water-moderated high-flux reactor designed for materials testing.

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Workers at the MTR conducted nearly 20,000 neutron irradiations prior to final reactor shutdown in

1970. Neutron monitoring of MTR workers was conducted using NTA film during this time. In the early years of MTR operation, Sommers [1959] examined doses in occupied areas and compared neutron measurements from NTA film to neutron “long counter” measurements and found that the film significantly underestimated dose from the moderated fission spectrum. Sommers [1959] suggested that the bias could be reduced by conducting calibrations using fission spectra instead of PoBe. Actual workplace spectra from the MTR were not available; however, Rohrig and Bump [2006] recently recommended that the reported dose equivalent from NTA measurements of MTR workers should be increased by a factor of 2.0 ± 0.3 based in a reanalysis of field measurements by Hankins [1961]. This correction is in reasonable agreement with dose coefficients (mSv∙mSv -1) calculated for reactor operators at Hanford reported by Fix et al. [2005].

Other sources of INL neutron exposures are shown in Table 5-10. Exposure scenarios in these locations typically involved only a few workers performing relatively simple tasks in highly controlled situations (e.g., instrument calibration, waste characterization). Moreover, potential exposures would have occurred in more recent years when marked improvements in neutron dosimetry are noted. Thus, these sources were not likely to result in significant undetected neutron dose to most persons at risk of exposure.

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Table 5-10. Other INL Neutron sources

Loca tion Period Comment TAN 1986 -2005 Spent fuel storage pad NRAD Facility 1977 -2006 TRIGA reactor operated for neutron radiography and the development of radiography techniques RWMC SWEPP The Passive Active Neutron (PAN) assay system generates 14-Mev neutrons for waste characterization. CPP -1649 1985 -2005 PuBe, AmBe sources used to calibrate criticality detectors CF -633 (HPIL) 1970 -2002 252 Cf calibration facility CF -636 (DOELAP Irradiation 1952 -1994 AmBe calibration facility Facility) TRA MTR -635 (blockhouse) Unknown AmBe, RaBe, and 252 Cf calibration sources

CF=Central Facilities; CPP=Chemical Processing Plant; HPIL=Health Physics Instrument Laboratory; NRAD=Neutron Radiography Facility; RWMC= Radioactive Waste Management Complex; SWEPP=Stored Waste Examination Pilot Plant; TAN=Test Area North; TRIGA=Training Research and Isotope General Atomic.

5.4.1.3 LANL

Prior to 1949, LANL personnel who worked with the cyclotron and other neutron sources used a Victoreen pencil PIC to measure and record neutron exposures in “n-units.” A 1968 study by Hankins [1968] who examined the response of the neutron PIC, indicated that one n-unit corresponded to about 60 mGy of cyclotron-produced neutron dose ( ±25%). In late August 1949,

Nuclear Track Plates (NTPs) were first issued for the evaluation of fast neutron exposures in those

workers presumed to be at risk. However, the first formalized neutron dose reporting (in medical

records as % tolerance) did not take place until May 1950 [Lawrence 1956b]. NTPs were comprised of a

100 µm nuclear emulsion (silver-bromide) on a glass backing issued on a case basis and worn for

approximately four weeks. As with the use of PICs, dose quantification using NTP measurements were

performed only if “appreciable” exposures were anticipated. LANL incorporated the use of NTA film in

the brass-cadmium beta-gamma dosimeter in the late 1950s. The NTA film was contained in the

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Eastman (Kodak) Experimental Personnel Neutron Type B film packet along with another emulsion (Fine

Grain Positive) for high range gamma exposures [Littlejohn 1961]. LANL began phasing out the brass- cadmium dosimeter beginning in early 1960s. Its replacement, the Cycolac multi-purpose film dosimeter, also included the Eastman Type B packet [Storm and Shlaer 1964a]. LANL began assigning site-specific TLNDs to workers in September 1978 and completed the changeover from film dosimetry badges to TLND-containing badges on January 1, 1980; however, NTA film continued to be used in some situations until completely discontinued in mid 1981 [Lawrence 1981]. The LANL Track-Etch Dosimeter

(TED) was placed into service in 1995. The TED contains three dosimetry-grade CR-39 track-etch plastic foils. The foils are placed in a hemispherical-shaped ABS plastic case on the sides of a triangular polystyrene pyramid to minimize angular dependence of the TED. The LANL TED, which is sensitive to only neutron radiation, is used for special field conditions involving higher-energy neutron radiation.

Dosimeter exchange frequencies varied by dosimeter type, monitoring era, and exposure scenario. NTPs were typically exchanged every four weeks, although more frequent exchanges may have occurred in special circumstances. Likewise NTA films were exchanged and processed monthly for most workers. The LANL TED was processed routinely every quarter.

Two Van de Graaff accelerators (0.5, 0.6, 0.8, 1.0, 1.5, 4, 5, 8, 17 and 20 MeV) and one

Cockcroft-Walton accelerator (2.5 and 14 MeV) were used to calibrate NTP and early NTA-based dosimetry [Littlejohn 1961]. Later, NTA films were calibrated to workplace spectra determined by field measurements as a secondary standard. For example, NTA film for persons working in DP West, where a higher fraction of intermediate and low energy neutrons were present compared to other work locations, used a higher dose coefficient (0.24 mSv∙track -1∙mm 2) than most other areas (~0.08 mSv∙track -

1∙mm 2).

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5.4.1.4 ORNL

ORNL initiated fast neutron exposure monitoring in early 1945 [Bradley 1945]. During this time, routine neutron monitoring was performed on a bi-weekly basis for about 100 workers assigned to the graphite reactor building (i.e. the “pile” building, bldg. 105). Measurements were conducted using

“Eastman Special Fine-Grain Particle Safety Film”, which is believed have been a precursor to NTA film.

Dose records were reported in percent of 8-hour tolerance, which was set at 200 n∙cm -2∙sec -1 for fast

neutrons [Morgan 1947]. This tolerance value was considered equal to receiving a daily dose of 20 mrep

(milli-roentgen equivalent physical). By 1949, NTA film was used in all assigned employee badges and

exchanged on a weekly basis [Thornton et al. 1961]; however, the film was processed for only those

workers determined to have been potentially exposed. By the mid-1950s, NTA film exchange was

weekly for those few workers with a high potential for exposure, quarterly (if processed), and every six

months for all others. Typically, less than 25% of the films were developed and read ( ∼2,000 annually)

[Gupton 19633 #9300; Morgan et al. 1967]. Electronic data suggest that separate recording of neutron

dose began in 1962. The policy for recording neutron dose prior to 1962 is unknown, although it is

possible that positive doses were summed with the low-LET results.

Early neutron exposures were expressed in terms of number of tracks per field (of view) or,

more specifically, number of tracks in a certain number of fields. For example, fast and thermal neutron

tolerance levels were both 20 tracks per 12 fields in early 1949. The previous value (from late 1947

through the eleventh week of 1949) was 16, although it is not clear if this value was also based on 12

fields. It appears that the fast and thermal neutron tolerance level increased to 22 tracks as of the 38th week of 1949. The practice of recording beta-gamma and neutron film badge readings on the same data card was discontinued on December 30, 1951.

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Albedo neutron dosimetry was incorporated into the ORNL film badge in the late 1970s. These dosimeters used both NTA and TLDs (TLD-600 and TLD-700) to quantify exposures to a broad spectrum of neutron energies [Gupton 1978]. The “red-dot” dosimeters used by most ORNL workers in the early to mid 1980s used TLNDs exclusively; however, special dosimetry (i.e., white-dot dosimeters) continued to include NTA film in unique neutron exposure situations [Berger and Lane 1985]. NTA films were discontinued in the mid-1980s following the implementation of the Panasonic neutron TLND dosimeter used in tandem with the two-chip Harshaw dosimeter. In January 1989, neutron and beta-gamma dosimeters were integrated into the four-chip Harshaw TLD system (Harshaw Model 8806B), resulting in the discontinuation of the Panasonic neutron dosimeter by 1990. ORNL dosimetry systems received

DOELAP accreditation on January 31, 1992.

Calibrations of early film techniques were conducted using a RaBe source [Bradley 1945]. It appears that PoBe sources may have been used for NTA calibrations beginning in the late 1940s [Kalmon

1949] and the Graphite Reactor in the early 1950s [Davis et al. 1954]. By the late 1960s, calibration of

NTA film dosimeters used either unmoderated PoBe or PuBe sources, with dose expressed in terms of equivalent exposure from a PoBe source. Specifically, ORNL used a fluence-to-dose equivalent conversion factor of 2.23 x 10 -5 n∙cm -2∙Sv -1, based on spectral comparisons between the calibration

sources and a PoBe spectrum. By the 1980s ORNL neutron calibrations were routinely conducted using

PuBe or 252 Cf neutron sources [Gupton 1981]. TLND responses were determined using the Health Physics

Research Reactor (HPRR) in combination with a variety of shielding components to mimic workplace

spectra [Berger and Lane 1985]. Dose algorithms were derived based on the response characterization

and the characterizations of workplace neutron spectra.

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5.4.1.5 PNS

PNS began personal neutron monitoring using Kodak type A NTA film in October 1957 [PNS

1980]. Film processing and calibration were initially conducted at the Brookhaven National Laboratory

(BNL) until July 1958, when PNS neutron dosimetry systems became fully operational. Calibrations at

BNL used a PoBe source (1957-1958), while PNS NTA film calibrations were performed using a PuBe source. NTA film monitoring was replaced by TLNDs featuring two-chip design in 1978 and a four-chip design in 1989 [Daniels et al. 2004].

Neutron exposures were possible only during submarine reactor operations, which were limited to a few civilian workers involved in power testing and operations during sea trials. Moreover, occupancy of the reactor compartment was prohibited during criticality and the compartment was heavily shielded to prevent neutron exposures in adjacent occupied spaces. From 1959 to 1975, only

111 PNS workers were found with recordable neutron exposures. These workers had a collective dose of about 0.04 person-Sv from neutrons, which was approximately 0.03% of the total collective dose during that period [Murray and Terpilak 1982]. The rarity of neutron exposures in civilian shipyard workers is also evident in later years, based on reviews of annual dosimetry reports. Therefore, it is not likely that

neutron exposures at PNS, detected or otherwise, contributed significantly to the total dose in

study subjects.

5.4.1.6 SRS

ORNL provided SRS with NTA film dosimeters and dosimetry processing services beginning in

1951 [Taylor et al. 1995]. NTA film was housed in the ORNL film badge issued and processed weekly. SRS implemented their in-house personnel dosimetry capabilities on August 3, 1953, which included onsite processing of NTA film. SRS continued weekly dosimeter processing until July 14, 1960, when the badge

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exchange cycle was extended to two weeks. The TLND dosimeter replaced NTA in 1971. SRS implemented a commercial Panasonic TLND system on January 1, 1995.

NTA film calibrations were conducted using PoBe (ORNL; 1951-1953), PuBe (1953-1964), and

PuF 4 neutron sources (1965-1971). SRS used a factor of approximately 1.3 to adjust for an underestimate in dose resulting from neutron spectra of energies below the film’s detection threshold. SRS reported a minimum recordable dose of 0.3 mSv. The TLNDs were calibrated using a PuF 4 source until replaced by

252 Cf in 1995. In the 1980s, SRS began exposing calibration TLNDs on-phantom (used to simulate worker’s body). As such, previous calibrations do not include response from neutron radiation backscatter response [Taylor et al. 1995].

5.4.2 Methods

Facility, dosimetry, and employment records were searched to gather information on neutrons exposures and health physics practices. Patterns of neutron exposures for workers included in the cohort were examined using available records of personal measurement data. Individual unadjusted neutron measurements were combined, while avoiding duplication, to estimate annual doses to each worker at the facility level and for the cohort. Annual doses were summed over the years monitored to provide estimates of cumulative neutron dose.

For each of the cases and controls, the recorded measurements were adjusted to normalize results between facilities and across time and to provide estimates in terms of absorbed dose to red bone marrow. The specific methods used for adjusting neutron doses were developed for a previous study [Schubauer-Berigan et al. 2007a] and are discussed elsewhere [Fix et al. 2005]. In general, historical records from each study facility were searched for information on personal neutron dosimetry

equipment, calibration, and dose reporting procedures. Individual employment histories were used to

determine sources of neutron exposure. Information on neutron spectrum in various work locations was

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used to adjust energy response functions for the respective dosimeter types. Using these data, Fix et al.

[2005] calculated a set of dose conversion coefficients to adjust reported doses to absorbed dose to bone marrow. Final dose estimates were determined by multiplying the recorded dose by the dose conversion coefficient selected based on facility, time-period, and exposure type determined from review of the worker’s employment history in the period of exposure. A description of the dosimeters used is provided in Table 5-11 and the criteria for grouping workers to specific spectra are provided in

Table 5-12 and Table 5-13. The final dose conversion coefficients used to make adjustments are shown in Table 5-14.

The methods developed by Fix et al. [1997] pertained to workers at the Hanford site, LANL, SRS,

PNS, and ORNL. Site specific spectral data were not available for INL. In the case of INL neutron exposures, it was assumed that all workers were exposed during reactor operations. The INL DCFs were based on the average of DCFs for reactor operators at Hanford, ORNL, PNS, and SRS stratified by dosimeter type. There were several instances were recorded neutron dose was available from facilities other than the study facilities. General assumptions about the source term and dosimetry used were unavoidable in the absence of facility and job information. For example, the INL composite DCFs were applied in cases were reactor operations were assumed to be the most likely source term (e., g., commercial power). LANL plutonium worker DCFs were applied for other plutonium facilities (e.g., RFP and MMP).

It is important to acknowledge that this analysis relied on the available monitoring data without considering the data accuracy, precision, or completeness of personal neutron exposure information.

Missed dose from neutrons from incomplete monitoring or insensitive dosimetry systems was not assessed. It is noted that safety requirements during early nuclear facility operations were less stringent; thus, personnel neutron dosimetry was used sparingly and available early monitoring data were mostly uninformative. However, significant undetected neutron exposures were likely in only a few operations

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and are limited mainly to MED-era plutonium workers and persons working in accelerator research. A large bias is not likely to result from neutron exposure misclassification because study inclusion criteria have vastly reduced the inclusion of these workers.

Table 5-11. Neutron dosimeter type, period of use, and exchange frequency 1 [Fix et al. 2005]

Dosimeter Period of first use and exchange frequency 2 3 Type Hanford LANL ORNL PNS SRS INL PIC(B) 1944 D 1943 D 1943 D n/a n/a n/a NTA/NTP 1945 5 1950 W,B 1950 4 M M 1957 V 1951 6 W,B 1951 W,B,M 1949 TLND - Site 1972 M 1978 M 1975 Q 1978 M 1971 M 1975 B,M,Q TLND - 1995 M 1990s M 1989 Q 1989 M 1995 M 1995 M Commercial TED 1995 7 M 1990s Q n/a n/a n/a n/a 1HAN=Hanford site; INL= Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; SRS=Savannah River Site 2Radiation Worker routine exchange periods: D= daily, W= weekly, B= biweekly, M= monthly, Q= quarterly V= Not routinely conducted, case-basis for persons suspected of exposure. 3Dosimeter types: PIC= Pocket Ionization Chamber, NTA= NTA Nuclear Track Emulsion, NTP= Nuclear Track Plate (used exclusively at LANL), TLND= Thermoluminescent Neutron Dosimeter, TED= Track Etch Dosimeter 4LANL used NTP beginning in 1949 and then used NTA beginning in about 1951. 5A few assignments in 1945. Routine use began in 1949. 6SRS NTA dosimeter services originally provided by ORNL until SRS dosimetry system was functional on 8/1953. 7Hanford terminated TED system in about 2000 for lack of plutonium workplace relative low-energy neutron sensitivity and accuracy.

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Table 5-12. Workplace neutron spectra for worker exposure groups [Fix et al. 2005]

Worker exposure Avg. Neutron Description of Representative group Site Energy (MeV) Spectrum Reactor Hanford, SRS, 0.169 Spectrum measured by multisphere at SRS K-Reactor ORNL, PNS, and in 1985. INL Pu Worker Hanford, SRS 0.557 A composite of 3 multisphere-measured spectra, assuming that the worker was exposed during 3

activities: glovebox work, handling cans of PuO 2, and working in a Pu vault Other Hanford, SRS 0.363 Combination of spectra for Reactor and Pu workers Isotope Handler ORNL 0.565 Multisphere spectrum near cell wall of waste transfer area, ORNL REDC Other ORNL 0.367 Combination of Isotope Handler and Reactor Pu Worker LANL unknown A composite of 3 multisphere measurements made in the LANL Pu Facility Accelerator LANL 2.16 A composite of 5 spectra for shielded work locations worker outside accelerators Other LANL 0.639 Combination of Pu and Accelerator workers HAN=Hanford site; INL= Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; SRS=Savannah River Site

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Table 5-13. Group and exposure scenarios [Fix et al. 2005]

Exposure Group Hanford INL LANL ORNL PNS SRS Reactor 100 areas All 3001, 3005, All 100 Areas 3010, 3042 --- 7500, 7702 7710, 7900 Plutonium 200W area : TA-1: D Bldg, 221F, B Line 234-5, 231-Z, D-5 Sigma 221H, B-Line 271, 2736-Z; Vault; 772F 235F 300 area : TA-21 : DP 736A 308, 309, Site & DP ------324, 325, Site-West, 3745A, Bldg 21 3745B TA-55 : Pu Facility, PF Site, Vault Isotope --- 3038 Handler 7920 REDC, rm 111 7920 REDC, rm 211 ------7920 REDC, TDF 7920 REDC, waste cask

Accelerator LAMPF at TA------53 Other/ worker not worker not worker not --- worker not Unknown identified in identified in identified in identified in --- other other other other categories categories categories categories HAN=Hanford site; INL=Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; SRS=Savannah River Site

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Table 5-14. Calculated dose conversion coefficients for converting recorded neutron dose to absorbed dose to bone marrow [Fix et al. 2005].

DCF Worker Exposure RBM Dose Site Category Time Period Dosimeter Type (mGy/mSv) Hanford Reactor Worker 1950-1956 NTA film 0.603 Hanford Reactor Worker 1957-1962 NTA film 0.414 Hanford Reactor Worker 1963-1971 NTA film 0.303 Hanford Reactor Worker >1971 TLD 0.020 Hanford Plutonium Worker 1950-1956 NTA film 0.061 Hanford Plutonium Worker 1957-1962 NTA film 0.042 Hanford Plutonium Worker 1963-1971 NTA film 0.031 Hanford Plutonium Worker >1971 TLD 0.020 Hanford Other / Unknown 1950-1954 NTA film 0.092 Hanford Other / Unknown 1957-1962 NTA film 0.063 Hanford Other / Unknown 1963-1971 NTA film 0.046 Hanford Other / Unknown >1971 TLD 0.019 SRS Reactor Worker 1951-1953 NTA film 0.515 SRS Reactor Worker 1954 - 1965 NTA film 0.433 SRS Reactor Worker 1966-1971 NTA film 0.279 SRS Reactor Worker >1971 TLD 0.018 SRS Plutonium Worker 1951-1953 NTA film 0.052 SRS Plutonium Worker 1954-1965 NTA film 0.044 SRS Plutonium Worker 1966-1971 NTA film 0.028 SRS Plutonium Worker >1971 TLD 0.016 SRS Other / Unknown 1951-1953 NTA film 0.078 SRS Other / Unknown 1954-1965 NTA film 0.066 SRS Other / Unknown 1966-1971 NTA film 0.042 SRS Other / Unknown >1971 TLD 0.017 ORNL Reactor Worker 1945-1973 NTA 0.515 ORNL Reactor Worker >1973 TLD 0.020 ORNL Isotope Handler 1945-1973 NTA 0.052 ORNL Isotope Handler >1973 TLD 0.019

Other / Unknown 1945-1973 NTA 0.078 ORNL

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DCF Worker Exposure RBM Dose Site Category Time Period Dosimeter Type (mGy/mSv) ORNL Other / Unknown >1973 TLD 0.019 LANL Plutonium Worker 1951 to 1978 NTA 0.056 LANL Plutonium Worker >1978 TLD 0.019 LANL Accelerator Worker 1951 to 1978 NTA 0.078 LANL Accelerator Worker >1978 TLD 0.037 LANL Other / Unknown 1951 to 1978 NTA 0.056 LANL Other / Unknown >1978 TLD 0.027 PNS Reactor Worker 1957-1978 NTA film 0.433 PNS Reactor Worker >1978 TLD 0.018 INL Reactor Worker 1951-1974 NTA film 0.437 INL Reactor Worker >1974 TLD 0.019 Others Reactor Worker 1950-1978 NTA film 0.437 Others Reactor Worker >1978 TLD 0.019 Others Accelerator Worker 1950-1978 NTA film 0.078 Others Accelerator Worker >1978 TLD 0.037 Others Plutonium Worker 1950-1978 NTA film 0.056 Others Plutonium Worker >1978 TLD 0.019 Others Other / Unknown 1950-1978 NTA film 0.078 Others Other / Unknown >1978 TLD 0.019

HAN=Hanford site; INL= Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; SRS=Savannah River Site

5.4.3 Results

5.4.3.1 Cohort level

Neutron dosimetry measurements were available for 17,314 workers or 16.5 percent of the total cohort ( n=105,245). Exposure measurement data were not available for PNS workers, although few workers are anticipated to have a positive neutron dose [Daniels et al. 2004]. Neutron-exposed workers contributed less than four percent to the collective dose and the mean cumulative dose from neutrons

(3.7 mSv, unadjusted) was about five-fold less than that from low-LET penetrating radiation. Cumulative

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neutron doses ranged between the null and nearly 300 mSv. Similar to low-LET occupational exposures, the dose distributions for each facility and for all facilities combined were right-skewed, whereby about one-half of workers were assigned doses of less than 1.0 mSv (55.1%) and nearly 95% had dose assignments less than 20 mSv. Only a few workers ( n=8) were assigned cumulative neutron doses of 200 mSv or greater (Table 5-15).

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Table 5-15. Cohort level cumulative dose (mSv) statistics for neutron radiation

Statistic Combined Hanford SRS ORNL INL LANL Dose file ID 2 6, 8 24-27 19-21 10 13, 17 Year of first reported dose 3 1946 1946 1954 1962 1952 1949 Year last monitored 2001 2001 1999 2000 1999 1990 Neutron m onitored 4 17,314 3,965 2,306 2,870 2,930 5,377 Gamma monitored 101,750 28,241 14,202 15,676 33,391 9,403 Percent monitored 5 17.0 14.0 16.2 18.3 8.8 57.2 No. (%) by dose in mSv: 0 < H <1 7,137 55.06 2,384 (60.13) 1,045 (45.32) 568 (50.09) 1,829 (64.06) 1,569 (52.67) 1 ≤ H< 10 4,479 34.55 1,311 (33.06) 802 (34.78) 431 (38.01) 917 (32.12) 1,036 (34.78) 10 ≤ H< 20 622 4.80 131 (3.30) 211 (9.15) 65 (5.73) 68 (2.38) 149 (5.00) 20 ≤ H< 50 472 3.64 104 (2.62) 179 (7.76) 43 (3.79) 26 (0.91) 118 (3.96) 50 ≤ H< 100 161 1.24 29 (0.73) 56 (2.43) 21 (1.85) 10 (0.35) 45 (1.51) 100 ≤ H< 200 84 0.65 5 (0.13) 12 (0.52) 5 (0.44) 5 (0.18) 57 (1.91) H ≥ 200 8 0.06 1 (0.03) 1 (0.04) 1 (0.09) 0 (0.00) 5 (0.17) Mean dose (mSv) 4 3.72 3.22 7.50 2.22 2.25 3.99 Maximum (mSv) 287.0 200.0 217.9 274.6 125.9 287.0

Collective dose 64.48 12.78 17.3 6.37 6.59 21.45 (person-Sv)

1Combined= all workers, some of which worked in multiple facilities; HAN=Hanford site; INL= Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; SRS=Savannah River Site 2file descriptions are provided in Table 5 1. 3The year of first reported neutron dose is based on inspection of electronic dosimetry data files. 4HAN and SRS data files do not discriminate monitoring status for all monitoring periods. Values shown for these facilities are based on positive dose values only. Statistics for all facilities include null values 5The ratio of neutron monitored to gamma monitored multiplied by 100.

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5.5 Plutonium Deposition

There is a paucity of information on the leukemogenicity of internally deposited alpha-emitters; most evidence of an effect stems from studies of patients receiving high-doses from Thorotrast injections. Previous studies of plutonium workers have not found evidence of elevated leukemia risk attributable to plutonium exposure. Nevertheless, plutonium deposition in hematopoietic tissue results in biologic damage that is consistent with other known carcinogens and may contribute to the overall leukemia risk in nuclear workers. Moreover, leukemia deposition in workers is positively correlated with external exposures, and thus may confound the observed effects for low-LET ionizing radiation that is being examined in this study. An assessment of plutonium deposition in study participants was undertaken primarily in efforts to adjust for the potential nuisance effects on models examining a dose- response for low-LET ionizing radiation.

5.5.1 Dose evaluation

All available plutonium urinalysis data for each exposed study subject were assembled from existing records. Site records were reviewed for information on bioassay methods, sample collection frequencies and methods, chemical extraction and recovery, counting techniques, reporting requirements, and detection levels. The absorbed dose to bone marrow from incorporated plutonium was assessed for moderate to highly exposed study participants using methods previously developed by

Daniels et al. [2006]. In general, a quantitative dose assessment was performed for each worker who had at least one detectable plutonium urinary excretion ≥1.7 mBq d -1. Dose estimates were calculated using a computer spreadsheet that incorporated plutonium retention functions, excretion functions, and dose conversion coefficients developed from reference modeling using the Integrated Modules for

Bioassay Analysis (IMBA) Expert [2002]. Dosimetry models used the most recent ICRP default parameters and applied standard assumptions on the route of entry, compound solubility, and time of

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uptake in the absence of information on these variables. The standard assumptions were: 1) each bioassay result represents total daily excretion from a single uptake; 2) the route of entry is inhalation;

3) the bioassay data collection occurred three days following intake; and 4) the inhaled material was

50% Type M and 50% Type S fresh weapons-grade plutonium. Under these assumptions, the 50-year committed absorbed dose to bone marrow per excretion was 0.039 mGy per mBq∙d-1.

5.5.1.1 Category Assignment

Study subjects were separated into three exposure categories based on the available bioassay data. The category definitions were based largely on an analysis of detection and reporting thresholds.

Subjects with at least one plutonium excretion result >17 mBq∙d-1 were placed into the “probable” exposure category (Category III). Those workers with “possible” plutonium uptakes were defined as having a maximum sample result between 17 mBq∙d-1 and 1.7 mBq∙d-1. The low exposure group

(Category I) included the remaining monitored workers. A fourth exposure category (Category IV)

included unmonitored subjects. Using the standard assumptions for a single uptake, Category III

corresponded to committed absorbed doses >0.65 mGy and Category II doses ranged from 0.07 to 0.65

mGy.

5.5.2 Results

5.5.3 Site characteristics

In most cases, sufficient information from site plutonium in urine bioassay procedures was

available that revealed the steps necessary for sample collection, radiochemical analysis, and the

reporting of the results dating back to the mid 1940s. Notably, procedures on the safe handling of

plutonium were not readily available at the start of the MED era. Instead, safety procedures, including

plutonium bioassay methods, were developed by researchers at the primary MED facilities (ORNL, LANL,

and Hanford) in concert with meeting nuclear weapons production for the war effort. Thus, MED-era

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bioassay methods were relatively crude and error-prone at a time when significant exposures were likely because of insufficiently developed and implemented exposure controls. However, continuous improvement, through trial and error by MED and early AEC medical researchers and health physicists, led to relatively sensitive and accurate plutonium urinalysis procedures by the mid-1950s. These methods were readily adopted by other AEC facilities with little modification, which may explain the many similarities observed in plutonium bioassay methods used in the study facilities.

Sample collection was similar at all sites. Large urine volumes were desired because of the minute amounts of plutonium excreted. Samples were collected over 24 hours or an equivalent of a 24- hour sample was used. The equivalent sample was typically defined as all urine passed by the worker for one half-hour before retiring and for one half-hour after arising for two consecutive days. The frequency of sample collection was in accord with the worker's potential for exposure; however, incidents involving the potential for plutonium uptake were a likely cause for sampling at all facilities. Baseline, termination, and periodic (typically annual or more frequent) urine samples were common in plutonium workers.

Radiochemical analyses at Hanford, LANL ORNL, and SRS were previously described by Daniels et al. [2006]. INL analyses are described in §5.5.3.2 below. In general, similar methods were used across these sites, although some procedural differences existed between sites and over time. Essentially, radiochemical analyses involved three basic steps: urine digestion (typically acid “wet ashing” with HNO 3 or HCl 4), plutonium extraction, and radioactivity quantification. Bioassay programs used a variety of

plutonium extraction methods over the years. These methods typically involved precipitation, solvent

extraction, or ion exchange, either separately or in combination. For example, SRS (1952-1959)

extraction procedures used two bismuth phosphate (BiPO 4) co-precipitation steps and two lanthanum

fluoride (LaF 3) co-precipitations, followed by plutonium extraction in thenoyltrifluoroacetone (TTA)

[Sanders 1956]. Solvent extraction using TTA or Tri-isooctylamine (TIOA) provided superior plutonium

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yields and was commonly used in early procedures. Yields (plutonium recovery) varied markedly by extraction method, but most facilities had achieved values greater than 80% by the mid- to-late 1950s.

Early radioassay was primarily performed by gross alpha counting using gas-flow proportional counters,

electrodeposition and autoradiography with NTA film, or zinc-sulfide (ZnS) scintillation counters.

Widespread use of alpha spectrometry substantially improved plutonium bioassay results. Spectrometry

methods surfaced as early as 1968 (LANL), but were not standardized at most facilities until the mid-to-

late 1980s. Spectrometry methods enabled isotopic discrimination and the use of plutonium tracers to

uniquely determine chemical yields.

A summary of extraction and quantification methods is provided in Table 5-16. The various

sources of excreta data used in the dose assessment are shown in Table 5-17. Details on the detection

capabilities and available bioassay records are provided in the following sections.

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Table 5-16. Major plutonium in urine bioassay methods in study sites

Site Extraction methods Radioassay methods

HAN LaF 3–TTA (1946-1964) gas-flow proportional counter (1946-1952)

LaF 3–TTA (1965-1983) electrodeposition and autoradiography (1952- 1983) ion-exchange (1983-) alpha spectrometry (1983-) INL Various methods were employed at INL at the electrodeposition and Frisch Grid ionization discretion of the chemist. However, the methods chamber or ZnS scintillation (1950s-1970s) below seemed to be used most often. alpha spectrometry (1970s-) LaF –TTA (1950s-1962) 3 alkaline earth phosphate precipitation and ion- exchange (1963-1981)

NdF 3 – Nd (OH) 3 coprecipitation and ion-exchange (1990s-)

LANL NH 4[C 6H5N(O)NO] (i.e., cupferron) in CHCl 3 gas-flow proportional counter (1944-1957) (chloroform) - LaF precipitation (1945-1949) 3 electrodeposition and autoradiography (1957-

BiPO 4–LaF 3 co-precipitation (1949-1957) 1963)

LaF 3–TTA (1957-1963) ZnS scintillation (1963-1966) alkaline earth phosphate precipitation and ion- alpha spectrometry (1966-) or exchange (1963-1981) thermal ionization mass spectroscopy (TIMS)

BiPO 4–LaF 3 or Ca(COO) 2 –LaF 3 co-precipitation (1997- (1981-1995) alkaline earth phosphate precipitation and ion- exchange (1995-)

ORNL BiPO 4–LaF 3 co-precipitation (1947-1957) gas-flow proportional counter (1947-1957)

BiPO 4–LaF 3 or Ca(COO) 2 –LaF 3 co-precipitation ZnS scintillation (1957-1989) (1957-1964) alpha spectrometry (1989-) NO – H O dissolution and ion-exchange (1964-) 3 2 2

SRS BiPO 4–LaF 3 co-precipitation followed by TTA electrodeposition and autoradiography (1954- solvent extraction (1952-1959) 1964)

NO 3 – H2O2 dissolution and ion-exchange (1959- solid-state surface-barrier detectors (1964-1988) 1966) alpha spectrometry (1988-) TIOA liquid extraction (1966-1988) Ion exchange (1988-) 1HAN=Hanford site; INL= Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; SRS=Savannah River Site.

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Table 5-17. Sources of plutonium bioassay information from 1945 through 2005.

Pu Filename FAC 1 Years Obs. Vars. Workers workers 2 Comments exc_result.sas7dbat HAN 1945-2001 456,844 34 29,850 28,013 REX excreta results int_master.sas7dbat HAN 1945-2001 3,289 27 2,408 612 REX confirmed intake data db_bioassay.sas7dbat INL 1972-1998 92,483 17 3,677 2,779 RDS database excreta results ocas_bio.sas7dbat INL 1952-1986 147,206 18 50,480 1,741 OCAS coded information [Geckler 2010] post1990_invitro.sas7dbat LANL 1973-2005 145,472 18 7,335 6,665 LANL internal dosimetry data abstracted from site systems pre1990_invitro.sas7dbat LANL 1944-2000 283,476 19 13,649 11,576 LANL internal dosimetry data abstracted from site systems ornl_invitro.sas7dbat ORNL 1949-1988 104,956 13 8,015 6,269 Table from Lockheed Martin dosimetry database hp_urin.sas7dbat ORNL 1988-1995 18,329 43 1,694 496 Table from Lockheed Martin dosimetry database bioassay.sas7dbat ORNL 1992-2000 120,590 34 2,797 1,861 Table from Lockheed Martin dosimetry database srs_bioassay.sas7dbat SRS 1953-1992 3,760 10 131 119 Data previously coded for the LCCS [Daniels et al. 2006] SRS Bioassay Database 1989- SRS 1990-2004 1,388,448 18 27,687 16,820 Database obtained from DCAS .mdb that contains SRS bioassay data after 1989 srbiom.dat SRS 1951-1978 24,647 13 321 320 Coded bioassay information from researchers conducting the DOE- sponsored National Study of Workers in the Nuclear Industry [Crase and Singh 1998]. 1HAN=Hanford site; INL= Idaho National Laboratory; LANL=Los Alamos National Laboratory incl. Zia; ORNL=Oak Ridge National Laboratory; SRS=Savannah River Site. 2Workers that have at least one recorded plutonium bioassay sample.

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5.5.3.1 Hanford

Hanford initially established the permissible plutonium level in urine at 46.7 mBq d -1 based on a one-year lag between uptake and sample, assuming an excretion rate at the end of one year of 0.004% per day [Healy 1948]. Urine bioassay methods were developed to detect, at a minimum, a resample value of 25% of the maximum permissible value (11 mBq∙d-1). By June 1949, improvements in counting methods had resulted in a reduction of the resample level to 5.5 mBq d -1. Plutonium sample electroplating and autoradiography resulted in a reported detection level of about 3.0 mBq∙d-1 in 1952.

By March 1953, methods were in place to routinely detect plutonium in urine at levels <1.7 mBq∙d-1

[Wilson 1987].

The REX database contained two data files that provided complete bioassay information on about 30,000 workers employed at Hanford from 1945 to 2001 (Table 5-17). One file (exc-result) contained bioassay sample information for all excreta samples, while the other file (int_master) provided information on the intake (e.g., intake date, route of entry, and incident location) for workers who were evaluated according to site procedures (i.e., confirmed uptakes). There were 456,844 bioassay records, of which 62% were plutonium assays (Figure 5-2). Of all workers who were monitored for incorporated radioactivity, most (94%) had at least one plutonium bioassay result. However, of 2,408 evaluations of confirmed internal depositions, only 612 (25.4%) were related to incorporated plutonium.

Of these, two study subjects had confirmed plutonium uptakes, although only one had a committed effective dose (i.e., 0.8 mSv) assigned by site dosimetrists.

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14000

12000

10000

8000

6000

4000

No. bioassay No.of Pu samples 2000

0 1950 1960 1970 1980 1990 2000 Sample year

Figure 5-2. Hanford plutonium bioassay samples (1950-2000).

5.5.3.2 INL

Routine urinalysis programs at INL were focused on incorporated fission products rather than deposited actinides. Plutonium bioassay was rare (<1% of samples) and was limited to persons suspected of exposure during incidents. The greatest potential for plutonium exposures was in workers at the ICPP. About 70% of the available plutonium urine bioassay samples were traced to work in ICPP areas. Incident samples included 24-hour urine and feces collections, which were analyzed for soluble and insoluble plutonium compounds, respectively. From 1952 to 1963 less than 100 urine samples were analyzed for plutonium compared to nearly 100,000 bioassay results for other radionuclides. Based on a review of the available sample data, routine plutonium bioassay did not begin until the early 1980s.

Because plutonium exposures were less likely at INL than in some other study facilities, early INL bioassay methods were somewhat less capable of detecting low levels of plutonium in urine. Review of early sample data (i.e., <1972) suggested a general detection limit of about 20 mBq∙d-1. There was a notable improvement in plutonium analyses beginning in 1972, whereby detectable plutonium in urine

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near 1.0 mBq∙d-1 was routinely reported. By 1974, methods were in place to routinely detect 239 Pu in urine concentrations of 0.3 mBq∙l -1.

Information on the actual bioassay procedures used in the early years is sparse. It appears that

an array of extraction and quantification methods was available to INL researchers and choices were left

entirely up to the chemist conducting the analysis. However, it is likely that the plutonium bioassay

methods selected mimicked proven methods used at other major AEC facilities. For example, there is

some evidence that a LaF 3–TTA solvent extraction procedure similar to that used by the Hanford site was employed by INL in the 1950s; however, there is no information on chemical yields. It is not known if the TTA was prepared in benzene. It also appears that INL may have used a precipitation of actinides on calcium phosphate [Ca 3(PO 4)2] from raw urine during this time, which reportedly had plutonium

yields around 95%. By 1962, INL had developed methods that involved wet-ashing urine with nitric acid

(HNO 3) and hydrogen peroxide (H2O2) and separation from bulk urine salts by precipitation from HNO 3

as the fluoride with a rare-earth carrier (LaF 3). The fluoride precipitate was then dissolved and the plutonium extracted as a chloride complex with a secondary amine (presumably LA-1) and electroplated to a platinum disc for alpha-counting. The average chemical yield of this process was about 95% with a coefficient of variation from 5 to 95% [Horan and Dodd 1962]. Actinide extractions using high molecular weight amines, such as the secondary amine Amberlite (LA-1) or the quaternary ammonium compound

Aliquat-336 (N-Methyl-N,N-dioctyloctan-1-ammonium chloride) were apparently used in the mid-to-late

1960s [Puphal 1994]. More recently, INL used co-precipitation on neodymium hydroxide [Nd (OH) 3] and the hydroxide precipitate was dissolved in concentrated HCI and passed though ion-exchange columns to extract the plutonium.

Early urine samples were analyzed by gross alpha counting using scintillation or by spectroscopy using a Frisch Grid ionization chamber. A 256-channel analyzer was first available for the Frisch Grid

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chamber in 1958 [Horan 1960]. Solid state alpha spectrometry replaced scintillation counting in the early 1970s, most likely accounting for the improved sensitivity ( ∼0.5 mBq∙d-1) discussed previously.

Electronic bioassay data were available from two sources. Files obtained from the INL Radiation

Dosimetry System (RDS) provided bioassay information from 1972 to 1998. Because very few plutonium urinalyses were conducted prior to 1980, the RDS data were the primary source of plutonium bioassay information on INL study subjects (Figure 5-3). A dataset from a large coding effort conducted by the

NIOSH Division of Compensation Analysis and Support (DCAS) also provided bioassay information from

1951 to 1986 [Geckler 2010]. These data overlapped the RDS information, but did provide important information on plutonium bioassay conducted at INL prior to 1972. In the case of overlapping years

(1972-1998), the RDS database was preferred. Unique identifiers were not available in the DCAS file; therefore, these data were matched to study subjects by name (last, first, and middle) for those subjects employed prior to 1972.

Incidents involving significant plutonium uptakes by INL workers were rare. Two incidents prior to 1977 were documented by INL researchers. The first incident involved widespread contamination following a spill of milligram quantities of highly insoluble 240 Pu dioxide from a welding glove box in the

Reactor Services Building (MTR-635) of the Test Reactor area (TRA) in 1963 [Sommers 1966]. Low-level

contamination quickly spread throughout many TRA buildings, sidewalks, and roadways. High level

contamination was present throughout MTR 635 and the Maintenance Building (MTR-653). Fecal

samples ( n=80) were collected from 22 workers, which indicated 10 workers with plutonium excretion.

One of these workers was subsequently treated with the chelating agent diethylenetriaminepentaacetic

acid (DTPA). All workers were estimated to receive annual lung doses less than 150 mSv. Plutonium

urinalysis was conducted for 69 workers ( n=103 samples), resulting in plutonium excretion levels from

not detected (ND) to ∼4,200 mBq∙d -1; however, only five workers had at least one sample above

detection. The second event occurred in 1972, when 13 laboratory personnel were contaminated with

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239 Pu at the ICPP during a routine mass spectroscopy sample preparation procedure [Wenzel 1973].

Fecal analyses conducted immediately following the incident were positive for plutonium contamination in all 13 workers. Urine was collected from those workers with the highest fecal plutonium concentrations. Plutonium in all urine samples was below detection. The committed dose to the lung for all but one worker was less than 4.0 mSv. One worker was reported with a committed lung dose of about 140 mSv. However, the ratio of 238 PU to 239 Pu observed in this worker was not consistent with that

of his coworkers, suggesting that his exposure occurred elsewhere. It was later determined that the

highly exposed worker inhaled an insoluble form of 239 Pu during cleanup of the ICPP X-cell (former laboratory) about seven months prior to the mass spectroscopy sample contamination event. It is unknown if other personnel were involved in the earlier event.

2000

1500

1000

500 No. bioassay No.of Pu samples

0 1950 1960 1970 1980 1990 2000 Year

Figure 5-3. INL plutonium bioassay samples (1952-1998)

5.5.3.3 LANL

Early sample techniques were designed to detect, with reasonable confidence, a tolerance level

(body burden) of 1.0 µg plutonium, assuming an excretion rate of 0.01% per day. The investigation level

was reported to be one count per minute above background for a 24-hour excretion [Langham 1947;

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Langham et al. 1945]. Assuming a nominal counting efficiency of 50% results in an investigation level of

33.3 mBq∙d-1. Although action levels remained essentially unchanged from 1945 to 1957 [McClelland

1954; McClelland 1958], a review of results reported in early medical records found urine assays approaching 1.7 mBq∙d-1 by 1950. TTA/co-precipitation and autoradiography procedures, beginning in

June 1957, increased measurement sensitivity to about 0.83 mBq∙d-1 [Campbell et al. 1972].

Site bioassay data for the years 1944-2005 indicate that the number of plutonium bioassay samples rapidly increased in the mid-1970s and peaked in 1981. After 1981, the frequency of sampling remained relatively constant, whereby about 4,500 ±800 bioassay samples were processed annually

(Figure 5-4).

8000

6000

4000

2000 No. of No.Pu bioassay samples

0 1945 1955 1965 1975 1985 1995 2005 Year

Figure 5-4. Number of plutonium bioassay samples at LANL (1945-2005).

5.5.3.4 ORNL

Early (1947-1949) plutonium bioassay methods had a detection capability of approximately 6.3 mBq∙d-1. From 1950 to 1967, methods were sufficient to detect, on average, plutonium in urine concentration of 6.3 mBq∙d-1. Continued improvement in bioassay lead to detection levels < 1.0 mBq∙d-1 by 1968.

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ORNL bioassay information was available electronically for urine samples collected from 1949 to

2000. There were 241,415 individual bioassay samples analyzed during this period, of which about

23.5% ( n=56,745) were identified as plutonium assays. The number of plutonium bioassay samples increased during the 1960s during peak production periods and again in the 1990s supporting site remediation activities (Figure 5-5).

4000 3500 3000 2500 2000 1500 1000

No. bioassay No.of Pu samples 500 0 1950 1960 1970 1980 1990 2000 Year

Figure 5-5. Number of plutonium bioassay samples at ORNL (1950-2000).

5.5.3.5 SRS

The reported plutonium bioassay sensitivity was about 0.6 mBq∙d-1 from 1952 to the mid-1960s.

The introduction of TIOA and surface barrier detectors in 1964 greatly simplified bioassay methods, but

resulted in a slight increase in reported sensitivity at 1.7 mBq∙d-1. Alpha spectrometry enabled separate reporting of 238 Pu (0.8 mBq∙d-1) and 239 Pu (1.2 mBq∙d-1) for routine samples in the early 1980s, and for all

samples by 1988 [Taylor et al. 1995].

Electronic bioassay information was not readily available on many SRS participants although

there was evidence that databases of excreta data were kept on certain workers over many years of

observation. For example, one file contained information on SRS plutonium worker that was intended to

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be used for a DOE-sponsored epidemiologic study that was planned in the mid-1980s [Crase and Singh

1998]. NIOSH became the custodian of these data in the 1990s. Although the intended study was never conducted, the data proved useful in reconstructing plutonium doses for some of current study subjects.

Prior to 1989, bioassay results were recorded on cards that were kept in the worker’s personal dosimetry file. Beginning in 1989, plutonium bioassay results on all monitored workers were available in

electronic form. For this study, information prior to 1989 was abstracted from: 1) information collected

for previous studies and 2) personal dosimetry folders and plutonium bioassay logbooks for those

subjects not previously studied. In addition, SRS maintained an internal dose report on employees

(n=982) identified with confirmed uptakes of radionuclides. These were workers who had potential

uptakes resulting in a committed effective dose equal to or greater than 0.10 mSv. There were 12 study

subjects found in the report with at least some portion of their estimated total internal dose resulting

from incorporated plutonium.

5.5.4 Results

The results from 5,302 in vitro urine samples analyzed from 1945 to 2005 formed the basis for

the category assignments in Table 5-18. The majority of urine sample data (41.5%) pertained to SRS

workers, followed by workers at Hanford (31.9%), LANL/Zia (14.4%), ORNL (9.5%), and INL (2.8%). There

were 563 study subjects (31.0%) with at least one bioassay result. Of these, four subjects had plutonium

bioassay information from two sites.

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Table 5-18: Plutonium exposure category assignments for study subjects ( n=1,816)

Category 1 Hanford INL LANL 2 ORNL SRS Total 3 (%) Samples I 1,485 144 720 424 1,9 48 4,7 21 (89.1) II 11 7 1 36 75 17 3 40 2 (7.6) III 89 2 5 5 78 179 (3.4) Total 1,691 147 761 504 2, 199 5,3 02 (%) (31.9) (2.8) (14.4) (9.5) (41.5) Persons I 190 15 42 47 142 436 (76.0) (Highest category II 23 1 20 35 28 10 7 (18.9) achieved) III 6 2 4 5 7 24 (4.2) Total 219 18 66 87 177 567 (%) (38.6) (3.2) (11.6) (15.3) (31.2) 1Where x is the urine bioassay result in mBq∙d -1, and categories are defined as follows: Category I is 0 ≤ x < 1.7; Category II is 1.7 ≤ x < 17; Category II is x ≥ 17. 2Includes study subject with employment at both LANL and Zia. 3There are four persons with monitoring results from multiple facilities. The column values are based on the highest category assignment obtained by the person, which may differ from the row totals.

There were 131 persons quantitatively assessed and assigned doses. The mean and median cumulative dose to bone marrow was 1.16 mGy and 0.16 mGy, respectively, and dose estimates ranged from <0.01 to 50.8 mGy without an exposure lag. The collective dose to bone marrow from plutonium was 0.15 person-mGy. SRS contributed most to the collective dose (89.5 person-mGy), followed by ORNL

(34.2 person-mGy), Hanford (19.4 person-mGy), LANL/Zia (8.0 person-mGy), and INL (1.7 person-mGy).

5.6 Tritium Exposures

5.6.1 Background

Tritium is a radioisotope of hydrogen that emits a low-energy beta particle (5.7 keV average energy), resulting in exposure hazards if taken into the body. The most likely uptake results from tritiated water or tritium oxide (HTO), which chemically behaves the same as ordinary water. Once introduced into the body, tritium is quickly (within one to two hours) distributed uniformly among soft tissues. Tritium is eliminated from the body with a biological half-life of approximately 10 days.

Tritium is produced by tertiary fission and neutron activation of deuterium in reactor coolant water. The predominant forms of tritium have been HTO and tritium gas. Some research and

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development projects at study facilities may have involved organically-bound tritium compounds (OBTs) or stable metal tritides, although study participant exposures to these compounds were not identified.

Workers involved in heavy water reactor operations, tritium production, recovery and processing, heavy water rework, laboratory analysis, and research and development were the mostly likely to be exposed to tritium.

Doses were calculated in terms of whole-body equivalent dose, and were typically combined with external exposure information to report the penetrating dose. In general, the basic dosimetry methods have changed little over time. Tritium doses were determined by fitting the urine data to an exponential, integrating the concentration curve, and multiplying by a conversion factor. However, some variation in dosimetry techniques has occurred, given differing assumptions of RBE, critical organ mass, biological half-life, and incident energy. Therefore, efforts were made to normalize doses across facilities and time in support of this study.

5.6.2 Methods

Historical documentation, dosimetry records, and medical records were evaluated to discover bioassay data and information pertaining to known incidents leading to confirmation of tritium uptakes.

Site records were reviewed to glean information on bioassay methods, sample collection frequencies and methods, chemical extraction and recovery, counting techniques, reporting requirements, and detection levels. HTO was assumed as the toxicant. Annual absorbed doses to the active marrow were tallied for each study subject from first exposure up to the cutoff date minus the exposure lag period.

Evidence of tritium bioassay without recorded dose indicated a potential for exposure and resulted in a zero dose assignment as a placeholder.

Assimilated tritium is distributed evenly throughout tissue as a component of body water.

Tritium dose models were based on complete assimilation following exposure and early models

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incorporated general assumptions that included; a wR of 2 for the tritium beta, 50 liters of body water as

the critical organ; and a biological half-life of about 10 days [Parker 1950b; Thompson and Kornberg

1954]. By 1954, the reference body water mass was reduced to 43 l [Thompson and Kornberg 1954],

which was also adopted by the ICRP by 1959 in its second publication [ICRP 1959]. The ICRP also

recommended a wR value of 1.7, an absorption fraction of 1.0, and a biological half-life of 12-days

[Beasley and Rouse 1961]. In 1986, the ICRP revised its tritium dose modeling methods whereby the w R was reduced to unity and the critical organ was changed from the body water to the soft tissues, increasing the organ mass to 63 kg. To account for changes in methods over time, recorded doses from

SRS and Hanford were normalized to reflect the recommendations of the ICRP [1992]. LANL updated all previous tritium dose estimates using the metabolic model and RBE recommendations in ICRP

Publication 30; thus normalization was not required. Given that the equivalent doses reported were normalized and interchangeable with absorbed dose, no further correction was needed (Table 5-19).

Table 5-19: Dose conversion coefficients used to normalize tritium doses.

Prior ICRP ICRP 2 Value ICRP 30 Value Time period <1954 1954-1985 ≥1986

Biological half-life 10 12 10 (days) Target tissue Body Water Body Water Soft Tissue Target mass (kg) 50 43 63

Radiation weighting 2 1.7 1.0 factor (w R) Dose coefficient 0.4 0.4 1

Missed dose from tritium exposures was estimated for those study subjects without reported doses but who had bioassay data available. Estimates were derived using a simple single compartment metabolic model. For this model, HTO was assumed to be completely absorbed into the systemic circulation without regard to intake pathway and is uniformly distributed in all body fluids, including

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urine. The worker was assumed to have a total volume of body water of 42 L distributed uniformly in 63 kg of soft tissue. The water exchange rate is 3 l ⋅d-1; therefore, the exchange rate constant, k, is 3/42=

0.0714 d-1. The removal by radioactive decay (half-life ∼12.3 y) is negligible compared to biologic removal mechanisms (half-life ∼10 d).

The activity of tritium in the body water at any given time, t, follows the first order equation:

ͯ&/ ̻ʚͨʛ Ɣ ̻͙ͤ -1 where A0 is the instantaneous tritium activity at uptake. The instantaneous dose rate, R (mGy ⋅d ), from the incorporated tritium at uptake is:

42,000 ͡͠ 5,700 ͙͐ 1 ͙̈́ͣͩ͠ 1,440 ͢͡͝ ́ͭ͡ · ͛͟ ͯͫ ͌ Ɣ ̽ͤ Ɛ Ɛ Ɛ ͥͬ Ɛ Ɛ ͯͧ Ɣ 8.7 Ɛ 10 ̽ͤ 63 ͛͟ ͕ͧͨ͛ͦͨͣ͘͢͢͝͝͝͝ 6.24 Ɛ 10 ͙͐ ͘ 10-1 ͙̈́ͣͩ͠ where C0 is the instantaneous urine concentration at time of uptake (dpm ⋅ml ). For this model, C0 was

determined by adjusting the bioassay urine activity concentration to account for removal between

uptake and sample collection. If the date of uptake was unknown, it was assumed to have occurred 3

days prior to the first urine sample.

The internal dose is a function of activity retention; thus, cumulative dose, D, for a single uptake

is proportional to the number of disintegrations that occur over the elapsed time since uptake ( t1-t0):

/u ͯ&/ ͌ ͯ&ʚ/uͯ/tʛ ̾ʚͨͥ Ǝ ͨͤʛ Ɣ ǹ ͙͌ ͨ͘ Ɣ ƫ1 Ǝ ͙ Ư /t ͟ The equation above was used to account for the dose contribution indicated by previous bioassay samples (i.e., uptakes) in dose estimates involving successive samples. The total committed dose (mGy) estimated from n samples (uptakes) was calculated by:

) ͌$ ̾ʚ∞ʛ Ɣ ȕ $Ͱͥ ͟ where Ri is the instantaneous dose rate for uptake i that was determined following any adjustment for

preceding uptakes.

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5.6.3 Results

5.6.3.1 Hanford

On or about October 1, 1948, Hanford initiated the P-10 Project, which consisted of the design, construction, and operation of facilities to extract tritium from irradiated lithium-bearing slugs [Reed

1952]. The tritium production facility (Building 108-B) was located in the 100B Area, where tritium was produced through 1954. One study subject was identified with tritium exposure while working in

Building 108-B during tritium production.

Six tritium-exposed individuals worked at the Plutonium Recycle Test Reactor (PRTR, Building

309) in the 300 Area. The PRTR was constructed to test fuels, experimental process tubes, fabrication techniques, and physics parameters associated with plutonium fuels. This reactor was moderated using heavy water (D 2O) obtained from SRS. The reactor operated from 1960-1968 with numerous interruptions. Tritium was produced either by neutron activation of the deuterium in the heavy water or by neutron activation with the 6Li in the LiOH used to chemically treat the primary coolant [Beasley and

Rouse 1961]. Operations and maintenance activities resulted in tritium uptakes by workers within this facility. By June 1964, Hanford management recognized that internally deposited tritium was posing a major problem at the PRTR [McConnon 1964]. Tritium exposure was said to account for 30% of the total penetrating dose of operations and maintenance personnel at PRTR. Tritium dose controls were implemented in 1964, which limited an individual to a maximum tritium excretion of 1,850 kBq∙l -1 and a monthly maximum dose of 3.5 mSv. Individuals exceeding these limits were removed from the area.

Bioassay samples were given following contact with tritium and follow-up samples were required for urine concentrations >185 kBq∙l -1 [McConnon 1964].

A tritium in-urine analysis program began at Hanford on or about May 1949 [Parker 1950a].

Detection of tritium in urine was initially performed using methods suggested by Healy [1949], which

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placed the urine sample into an acetylene generator and measured the radioactivity of the resultant gas with a vibrating reed electrometer. Depending on the background level, the detection limit at the 90% CI was estimated to be between 24.5 kBq∙l -1 and 44.4 kBq∙l -1 [Therburn 1950]. Initially, the bioassay throughput was approximately six samples per month [Parker 1950a]. In 1950, as the hazards of tritium became better understood, processing increased to 250-350 samples per month on two shifts using this methodology. Additional improvements continued, and by 1951 the capacity of the laboratory to analyze urine samples was six to eight samples per hour per instrument, with a monthly workload of over 1,000 samples [Kornberg 1951].

In 1949, the maximum permissible content of tritium in the body was 3.7 x 10 4 kBq (1 mCi). In

early 1950, this limit was revised to an activity density in urine of 2.4 x 10 4 kBq∙l -1 based on a total

-1 exposure limit of 3.0 mSv wk and wR of 2 for the tritium beta [Parker 1950b]. Dose calculations

considered body water as the critical organ and assumed the biological half-life was approximately 10

days [Thompson and Kornberg 1954]. Initially, the body was estimated to consist of 50 liters of water. In

1954, a value of 43 liters was assumed [Thompson and Kornberg 1954]. Beginning with PRTR operations,

dose calculations were consistent with ICRP Publication 2 [ICRP 1959] based on a wR of 1.7, an

absorption fraction of 1.0, and a biological half-life of 12-days [Beasley and Rouse 1961]. Tritium dose,

as calculated from bioassay results, was added to the whole-body dose from external penetrating

radiation. In 1989, Hanford adopted the ICRP Publication 30 methodology for dose calculations, where

the w R changed from 1.7 to 1.0 and the critical organ was changed from the body water to the soft tissues, increasing the organ mass to 63 kg [ICRP 1979]. The combination of these changes resulted in a reduction in assigned tritium dose [Carbaugh 2000]. Hanford did not reevaluate historic dose assignments following implementation of newer dosimetry techniques; therefore, recorded doses were normalized.

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A total of 1,155 annual tritium dose records for 488 workers were recorded in the REX Database.

These data were identified by “H” code in the DR_DOS_TYPE_CD field of the DOS_SUMM_RESULTS table. The records ranged from 1949 to 1972, with the majority (81.0%) occurring after 1961. There were nine study subjects identified with potential tritium exposure (Table 5-20). Cumulative doses to bone marrow among the nine workers ranged between 0.04 mSv and 6.12 mGy, with mean and median doses of 1.27 and 0.16 mGy, respectively.

Table 5-20: Hanford tritium exposures

Adjusted Unadjusted Dose Work ID Period Dose (mSv) 1 (mGy) Location Job Title 91412 1968 0.4 0.16 309 Instrument Technician 102628 1962 3.2 1.28 309 Technician 104219 1966 0.6 0.24 324 Health Physics Technician 104938 1965 0.1 0.04 309 Engineer 112425 1951 0.2 0.08 309 Engineer 117533 1962 -1965 8.6 3.44 309 Technologist 121751 1966 -1968 15.3 6.12 309 Plumber steamfitter 124368 1951 0.1 0.04 3706, 300 Chemical Helper 125998 1953 0.1 0.04 108B, 100B Millwright Journeyman 1Equivalent whole-body dose reported by the site without normalization

5.6.3.2 INL

Historical records on INL tritium exposures is sparse, although it appears that limited environmental monitoring began in June 1961, followed by personal monitoring shortly thereafter

[Horan and Dodd 1962]. The most important tritium source term appears to be HTO as a by-product waste material from processing alloy fuels at the ICPP. Sampling and analysis were consistent with methods used at other AEC facilities. In 1974, INL reported a detection limit for tritium in urine using liquid scintillation counting methods with a 10-minute counting time of 0.6 kBq∙l -1.

Separate reports of whole-body dose equivalent from tritium exposures were not available in site records; however, bioassay information was found that contained tritium urinalysis results for a few

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INL workers. A total of 3,704 urine samples were collected and analyzed for 689 workers between the years 1962 and 1998, and nearly 36% ( n=1,321) were performed in 1991 (Figure 5-6). Of these samples,

1,958 (53.1%) results were greater than the null. Of the positive results, the mean and maximum tritium

urine concentrations were 0.3 kBq∙l -1 and 111.0 kBq∙l -1, respectively. There were four cases and controls

identified with tritium bioassay samples (n=17). Among the cases and controls, one worker was

identified with one positive result of tritium in urine (<0.01 kBq∙l -1). Thus, tritium doses from employment at INL appear below to be negligible for the current study and dose estimates were not performed.

1400

1200

1000

800

600

400 Number of Numbersamples

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0 1970 1975 1980 1985 1990 1995 2000 Year

Figure 5-6. The number of tritium urinalysis samples at INL between the years 1974 and 1998.

5.6.3.3 LANL

As with Hanford and ORNL, records indicate LANL began researching the usefulness of tritium in the 1940s [Johnson 1949]. Approximately 5,180 TBq of tritium was estimated to have been released into the atmosphere due to LANL operations prior to 1973 [Valentine 1974]. Table 5-21 lists the atmospheric releases by point of origin and hazard rating.

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Table 5-21: LANL tritium releases 1943-1972

Area and Bldg. No. Nomenclature Potential Release Period Potential Magnitude TA-1-69 Building U 1944 -1957 Moderate TA-1-71 Building W 1950-1955 Minimal TA-1-83 Building Z 1944-1955 Minimal TA-3-16 Van de Graaff 1956-1972 Minimal TA-3-29 CMR (Chemistry & Metallurgy Research) 1961-1972 Not reported TA-3-34 Cryogenics 1955-1958 Minimal TA-9-21 Laboratory 1960-1966 Minimal TA-21-3 Laboratory Building 1951-1956 Not reported TA-21-5 Laboratory Building 1965-1972 Not reported TA-33-86 HP Laboratory Building 1958-1972 Large TA-35-2 Laboratory 1958-1972 Moderate TA-43-1 Health Research Laboratory 1955-1972 Minimal

The majority of tritium-related work consisted of Van de Graaff accelerator use by workers assigned to the Physics Division, Group 9, (i.e., P-9), initially at the W Building (TA-1-71) and the New

Van de Graaff Building (TA-3-16). Indicated as a large release hazard, the Gas Handling Facility (TA-33-

86) began operations in June 1955. It was the first facility at LANL to handle large quantities of tritium gas for the nuclear weapons development program. Its purpose was to conduct research and development on tritium-handling technology that would feed SRS tritium production. However, SRS was not ready to handle and fill gas reservoirs during the mid 1960s, and LANL (TA-33-86) took over the production work for a brief period of time. LANL processed and repackaged tritium gas into small- volume, high-pressure gas containers, which were used in several weapons systems and devices that were tested at the Nevada Test Site (NTS).

Tritium bioassay data were available for study subjects dating back to September 1950. Whole- body exposure evaluation via urinalysis began at LANL on February 29, 1952 [Lawrence 1956a]. The

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biological half-life was initially assumed to be 14 days until reduced to 10 days in 1955. The biological half-life was changed to 12 days following the 1956 proposed revisions to National Bureau of Standards

(NBS) Handbook 52. During this time, w R was assumed to be 1.7, and the critical organ was body water with a mass of 43.4 kg [Lawrence 1957].

There were ten study subjects with a tritium dose assignment at LANL. Another seven study subjects had at least one bioassay result, but no recorded dose. The majority of collective tritium exposure resulted from accelerator operations in the W building and TA-3-16. Maximum annual exposures were to physicists during Van de Graaff accelerator operations in the W Building during the

1950s. Exposures were also documented in TA-33-86. The distribution of cumulative doses was highly skewed, with arithmetic mean and median values of 14 mGy and 1.2 mGy, respectively.

Table 5-22: Tritium exposures among LANL cases and controls

Adjusted Dose Work Location (mGy) (Division- ID Period Group) 1 Job Title 550 57 1965 -196 6 0.9 GMX -4 Mechanical Fabrication technician 73901 1959 -1973 85.1 P-9 Maintenance Mechanic 79235 1976 1.5 H-1 Researcher (Staff Member) 98976 1952 0.4 P-4 Technician 100681 1966 0.1 P-9 Electrical Engineer 102753 1952 -1978 28.3 P-9 Physicist (Staff Member) 105806 1964 -1965 2.2 CMB -3 Chemist 107479 1965 0.4 GMX -9 Optical Technician 111366 1956 -1958 20.7 W-3 Physicist 128928 1954 0.4 GMX -5 Technician 1CMB=Chemistry and Materials; GMX=Explosives R&D; H=Health and Safety; P=Physics; W=Weapons Research

5.6.3.4 ORNL

A review of ORNL tritium bioassay data indicates that few ORNL study subjects were monitored for tritium exposures each year. Over 90% of the samples taken ( n=8, 959) resulted in urine concentrations less than 333 kBq∙l -1. The number of urine samples and monitored workers was greatest

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during the 1960s, which suggests a time period for most potential tritium exposures (Figure 5-7). During this time, tritium bioassay samples were assayed using liquid scintillation counting techniques. All results were reported in units of disintegrations per minute per milliliter (dpm∙ml -1) of specimen. Activity concentration results were converted to a whole-body dose equivalent according to the methods in

ICRP Publication 2 [Gupton 1969]. Equivalent doses were estimated for the individual exposure record only in cases where the instantaneous urine tritium concentrations exceeded 10 4 dpm∙ml -1 (∼167 kBq∙l -

1). Based on ICRP 2 Publication methodology, this concentration resulted in a committed dose of approximately 0.5 mGy, which was near the recording threshold of for external dosimetry using film badges.

samples workers

700 160 600 140 500 120 100 400 80 300 60 200 40 No. of No.urine samples 100

20 of No.monitored workers 0 0 1950 1960 1970 1980 1990 2000 year

Figure 5-7. Tritium bioassay samples at ORNL (1950-2000)

Information regarding ORNL tritium operations is sparse. However, available records indicate that tritium operations focused on bench-scale operations and minimal laboratory work. During the period from 1960 to 1970, the majority of job titles among monitored workers were related to laboratory work supporting biology research or isotope production. There were 13 study subjects with records of tritium urinalysis at ORNL. None of the workers were assigned estimates of equivalent dose

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by ORNL dosimetrists. All exposed subjects were laboratory workers with urine sample results collected from 1960 to 1969. Among those monitored tritium concentrations in urine ranged between ND and

102 kBq∙l -1. None of the subjects appeared to be chronically exposed and the maximum cumulative dose

received by an exposed worker was 0.32 mGy.

Work histories and facility records were evaluated to determine if additional exposure

potentials should be investigated. Early records indicated limited experimentation to develop

continuous tritium production methods was pursued by the Technical Division from 1945 to 1946 [Reed

1952]. These efforts were later superseded by the Hanford P-10 Project. Isotope production operations

involved tritium processing in the Radioactive Gas Processing Facility (Building 3033, formerly 905) from

1950 to 1990 [Kuhaida and Parker 1997]. Additionally, a tritium overexposure incident occurred in

Building 3028 during the fabrication of large (i.e., >74,000 GBq) tritium accelerator targets in 1966 [Roth

1966]. Although the incident was severe, there were no exposures to selected study subjects. The

Tritium Target Facility (building 7025) was used to construct tritium targets and analyze tritium diffusion

through metals. The facility operated from the late 1960s to 1989. Based on records review,

employment of unmonitored study subjects in Buildings 3028, 3033, or 7025 during periods of tritium

related work was unlikely.

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Table 5-23. Tritium exposures among cases and controls employed at ORNL.

Bioassay Results Adjusted in kBq∙l -1 Dose ID Period (No. of samples) (mGy) Department Job Title 850 1960 ND NA Chemistry Chemist 922 1969 -1971 ND -1.8 (2) <0.01 Chemistry Chemist 1774 1963 -1967 5.17 -101.7 (7) 0.32 Isotopes Laboratory Technician 2464 1969 0.63 <0.01 Biology Laboratory Technician 3153 1969 -1970 ND -3.5 (3) <0.01 Biology Laboratory Technician 4650 1972 0.6 <0.01 Isotopes Reactor Supervisor 5102 1969 0.23 <0.01 Biology Laboratory Technician 73316 1964 -1968 11.7 -18.3 (5) 0.09 Physics Research Staff 81282 1960 ND (1) NA Physics Physicist 83651 1969 -1973 7.8 -20.0 (3) 0.03 Health Physics Laboratory Technician Finance & Process Supervisor 84473 1983 6.5 <0.01 Materials Engineering & Millwright 85757 1969-1979 ND-6.8 (12) 0.01 Mechanical Engineering & Laborer 98981 1971 ND (1) NA Mechanical ND=not detected

5.6.3.5 SRS

The heavy water moderated reactors at SRS served to produce plutonium and tritium inventories needed for the AEC weapons program. Lithium-aluminum target assemblies (slugs) were manufactured in the 300 Area and irradiated to produce tritium gas. The lithium is enriched in 6Li and 3H

is formed by the 6Li(n,α) reaction. Gaseous separation and tritium purification took place in the 200-F and 200-H Areas. Major operations leading to tritium exposure included production reactor operations, product separations and purifications, and reservoir loading. Other tritium-related processes included laboratory research and heavy water rework; however, it is estimated that tritium releases in non-

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production areas accounted for less than one percent of total plant releases [SRS 1975]. Primary work locations for tritium exposures are shown in Table 5-24.

Table 5-24: Primary facilities for tritium exposures at SRS

Building Description Begin End Operations Operations 1

232F First tritium recovery facility 1955 1957

232H Tritium Recovery and Processing Facility 1957 NA

233H Replacement Tritium Facility (RTF) 1994 NA

234H Tritium Loading Facility 1957 NA

105C Heavy Water Moderated Reactor 1955 1986

105K Heavy Water Moderated Reactor 1964 1988

105L Heavy Water Moderated Reactor 1954 1968

105P Heavy Water Moderated Reactor 1954 1988

105R Heavy Water Moderated Reactor 1953 1964

420D Heavy Water Rework Facility 1952 1981

773A Savannah River Laboratory 1953 NA 1Tritium production ceased with the shutdown of all reactors in 1988. Remaining H Area facilities continue extraction from tritium-bearing targets irradiated in commercial light water reactors.

The Heavy Water Rework Unit (RW), the High Activity Moderator (HAM) facility, the Moderator

Processing Facility (MPF), and the Technical Purification Facility (TPF) were used to purify and maintain

the reactor moderator in order to ensure efficient operations and minimize the volume of waste for

storage and disposal. These operations involve work with a tritium-contaminated moderator, which

results in exposure hazards similar to reactor areas.

HTO and tritium gas are the most predominant chemical form of tritium at SRS. Less common

forms of tritium, such as stable metal tritides and organically-bound tritium, are also encountered at the

site. Doses were calculated from tritium urinalysis data and combined with external sources to report

whole-body exposures. From 1954 to 1958, urine samples were measured for tritium using a vibrating

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reed electrometer. The minimum detectable concentration (MDC) for this method was 37 kBq∙l -1, the

minimum reported dose was 0.10 mSv, and doses were reported in 0.05 mSv increments. During 1958,

SRS began analyzing tritium samples with a liquid scintillation counter (LSC). The original counting

system yielded an MDC of 37 kBq∙l -1. As LSC techniques improved, the MDC was lowered, first to 18.5

kBq∙l -1 (1981) and later (1986) to 3.7 kBq∙l -1 [Taylor et al. 1995].

Dosimetry methods changed over time, following a pattern similar to that of Hanford. Initially, tritium dose assessment was consistent with ICRP Publication 2 [Taylor et al. 1995]. Doses were calculated for tritium bioassay results >185 kBq∙l -1 [Butler 1958]. In 1971, the data indicate that doses

were calculated from tritium in urine ≥37 kBq∙l -1. In 1981, the site adopted a policy of calculating dose

for results greater than/equal to the detection limit. In 1986, the site adopted the methods

recommended by ICRP Publication 30, although the change in biological half-life and mean beta energy

probably occurred before 1981. Like Hanford, the implementation of ICRP Publication 30 resulted in

reduction in estimated doses by approximately 60% due to the change in dose calculation factors. Unlike

other internal emitters, historic intakes of tritium have not been re-evaluated by SRS using newer

metabolic models [Taylor et al. 1995].

There were 249 SRS study subjects with evidence of potential tritium exposure. Of these, there

were 68 workers with at least one positive bioassay result. There were 4,814 person-years of potential

exposure, with 376 person-years of recorded doses greater than zero. After restricting the data to

positive values, the mean and median doses were 3.19 mGy and 0.73 mGy, respectively. The maximum

cumulative dose from tritium exposure among SRS study subjects was 28.7 mGy and the collective dose

attributable to tritium was 0.21 person-Gy. Most of the collective dose (51.7%) was accrued among

operations staff, (i.e., reactor and separations operators) followed by maintenance workers (31.9%), and

technical support (15.9%); however, the highest average dose (4.9 mGy) was observed in technical

support staff (e.g., laboratorians, scientists, technicians).

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Table 5-25. Tritium exposures among cases and controls employed at SRS

Unadjusted Adjusted Exposure Dose Dose ID Period (mSv) 1 (mGy) Job title 2043 1972 -1975 0.15 0.06 Separations Foreman 3546 1961 0.25 0.1 Laboratory Shift Supervisor (chemist) 4076 1958 -1984 12.95 5.18 Laboratory Shift Supervisor (chemist) 4767 1969 -1978 1.55 0.62 Health Physics Supervisor 5273 1964 -1976 1.05 0.42 Reactor Supervisor 5505 1973 -1978 0.25 0.1 Laboratory Area Supervisor (chemist) 5520 1984 0.05 0.02 Laboratory Technician 5959 1966 7.3 2.92 Engineer 58390 1975 0.15 0.06 Engineer 58918 1971 -1982 0.45 0.18 Security 59427 1956 4.4 1.76 Instrument Foreman 60402 1961 0.2 0.08 Engineer 60471 1959 -1983 12.05 4.82 Reactor Shift Supervisor 62509 1977 -1978 2.4 0.96 Separations Operator 63152 1971 -1982 0.2 0.08 Shift Supervisor GS 63324 1979 0.45 0.18 Instrument Foreman 63955 1973 0.15 0.06 Fire Equipment Operator 63974 1961 -1981 10.35 4.14 Reactor Shift Supervisor 64472 1958 -1980 46.95 18.78 Reactor Operator 64783 1955 -1958 5.2 2.08 Maintenance Mechanic 65043 1957 0.7 0.28 Maintenance Engineer 65072 1962 -1991 50.7 20.79 Instrument Mechanic 65290 1980 -1983 0.45 0.18 Operator 65835 1961 -1984 18.85 7.54 Reactor Foreman 66061 1984 0.15 0.06 Clerk 66432 1968 -1977 5.95 2.38 Laboratory Technician 66553 1966 -1972 71.8 28.72 Technical Assistant 66717 1958 -1983 40.3 16.12 Heavy Water Operator 66743 1958 -1980 24.4 9.76 Reactor Operator 67110 1964 -1984 6.1 2.44 Auxiliary Operator 67867 1986 -1987 0.1 0.1 Clerk 68000 1958 -1968 16.1 6.44 Auxiliary Operator 68365 1972 -1975 0.35 0.14 Driver 68432 1978 0.15 0.06 Power Operator 68461 1972 0.6 0.24 Maintenance Mechanic 68499 1973 -1978 0.5 0.2 Power Operator

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Unadjusted Adjusted Exposure Dose Dose ID Period (mSv) 1 (mGy) Job title 68676 1970 -1984 4.7 1.88 Auxiliary Operator 68761 1958 -1977 7.75 3.1 Instrument Mechanic 68883 1972 -1975 1.95 0.78 Utility Mechanic/Painter 69016 1971 -1977 1.6 0.64 Separations Shift Supervisor 69392 1957 -1963 4.5 1.8 Auxiliary Operator 69586 1960 -1964 16.55 6.62 Maintenance Mechanic 70008 1966 5.8 2.32 Separations Operator 70122 1981 0.1 0.04 Storekeeper 70345 1971 -1972 0.65 0.26 Clerk 70523 1960 -1976 20.65 8.26 Maintenance Mechanic 70692 1977 -1981 0.55 0.22 Instrument Mechanic 70728 1960 -1979 20.15 8.06 Auxiliary Operator 70755 1958 -1987 28.95 11.67 Auxiliary Operator 71033 1981 -1987 1.4 0.68 Mechanic 71260 1973 -1974 2.55 1.02 Reactor Operator 71355 1961 -1973 0.8 0.32 Laborer 71456 1956 -1965 6.7 2.68 Auxiliary Operator 71624 1974 -1975 0.45 0.18 Instrument Mechanic 71721 1974 -1978 0.95 0.38 Maintenance Mechanic 71987 1966 0.7 0.28 Utility Operator 72461 1973 -1977 3.25 1.3 Instrument Foreman 72655 1970 -1972 3 1.2 Rigger 72858 1965 -1982 7.15 2.86 Reactor Shift Supervisor 72870 1974 -1980 8.45 3.38 Instrument Mechanic 80798 1969 -1970 4.6 1.84 Maintenance Mechanic 82503 1990 0.05 0.05 Clerk 87569 1960 1.3 0.52 Reactor Clerk 91350 1956 -1975 38.45 15.38 Instrument Mechanic 99824 1956 0.55 0.22 Reactor Operator 102655 1965 0.6 0.24 Physicist 131592 1973 -1975 1.95 0.78 Instrument Mechanic 218677 1996 -1999 0.03 0.03 Unknown

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5.6.4 Discussion

Tritium dose assessment has been performed via urinalysis and is subject to uncertainty due to variations in body metabolism. The kinetics of HTO in the body follow that of body water, except for a small portion that is bound to organic molecules [Hill and Johnson 1993]. Urine concentration data were used to estimate whole-body doses based on standard assumptions of certain variables. For example, most dose calculations employed a default assumption for the biological half-life of HTO that was approximately equal to 10 days. However, the normal excretion half-time for tritium in body water has been observed to vary by a factor of two for adults [Hill and Johnson 1993; Inkret et al. 1999]. Various metabolic models have been preferred over the years. For example, metabolic models for HTO have historically considered a single component characteristic of the retention of body water [Hill and

Johnson 1993]. However, recent models have included an additional component for tritium that is organically bound following HTO uptake [Inkret et al. 1999].

Similar to the uncertainties associated with plutonium, there is limited information regarding the RBE of tritium for carcinogenesis. Currently, the ICRP recommends a wR of unity for tritium beta radiation [ICRP 1992]. However, Straume and Carsten have summarized tritium RBEs for carcinogenesis and suggest wR ranges between 1 and 2 when related to orthovoltage x-rays as the comparison radiation

[Straume and Carsten 1993], and ranges between 2 and 3 when compared to gamma rays [Straume

1993]. Largely based on the earlier work by Straume and Carsten [1993], Kocher et al. [2005] suggested

the preferred radiation effectiveness factor for low dose and low dose rate tritium exposure is

lognormally distributed with a median value of 2.4 ( 95% CI: 1.2, 5.0).

In summary, the RBE for tritium is relatively close to unity compared to high-LET radiations, the

number of tritium-exposed subjects is small, and significant tritium exposures were rare. Therefore,

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adding tritium doses to other sources of low-LET dose without consideration of RBE differences is not anticipated to introduce a significant bias in cumulative dose used in dose-response analyses.

5.7 Benzene exposures

Previous studies [Kubale et al. 2005; Schubauer-Berigan et al. 2007a] and available process information suggest that benzene use was limited to small numbers of workers performing a narrow range of activities in the study facilities, primarily before the 1980s. These activities included pure benzene use in the early manufacture of explosives (LANL), rubber work (PNS), and laboratory activities

(all facilities), although most non-laboratory use had ceased by the mid-to-late 1950s. Workers were also potentially exposed through the use of products containing benzene as an additive or impurity, such as such as gasoline, cements, adhesives, coatings, paint thinners and removers, degreasers, and some organic solvents. Benzene contamination varied by product and by time, but is believed to be generally less than 5% (volume basis) from the 1960s to the 1970s and less than 0.1% in most commercial products thereafter. For example, it was estimated that concentrations of 0.1% to 5% benzene was common in hexane, naphtha, and toluene in the mid- to-late 1970s, while xylene, ethylbenzene, Stoddard solvent, and mineral spirits contained ≤0.1% benzene during this period

[Williams et al. 2008].

Where benzene and benzene-containing materials were associated with tasks involving

radioactive materials (e.g. solvent extraction of plutonium in urine), controls used to mitigate radiation

hazards (e.g., containment, ventilation, PPE) also provided some protection against benzene. Many of

these controls were put into service several years prior to widespread acceptance of benzene

carcinogenicity (circa 1970s); therefore, average benzene exposures in the nuclear industry were likely

to be less than exposures in other industrial settings where benzene was present in similar quantities.

Nevertheless, some workers (e.g., painters, automotive mechanics) may have been exposed to benzene

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without concomitant radiation exposure in occupations requiring few radiation exposure controls. Thus, hazards in these occupations are expected to be similar to that found outside the nuclear industry.

Industrial hygiene measurement data for benzene is sparse or not evident in facility records.

Therefore, exposure surrogates were constructed using available information on processes involving benzene and individual work histories. The approach used was similar to that in a previous study

[Schubauer-Berigan et al. 2007a], whereby “ exposure scores” were developed based on combinations of qualitative ratings of exposure intensity and frequency. Task-based scores were then accumulated for each study subject over the time under observation. In general, four steps were used to estimate these scores:

1. A literature search and site records review were performed to collect information needed to identify

and characterize benzene-related tasks and exposure potentials. Site-specific data were preferred

for characterization, followed by information from similar industry settings obtained from published

articles and reports. The site records review included records from: NIOSH collections from previous

studies, the Information Bridge - DOE Scientific and Technical Information Database maintained by

the DOE Office of Scientific and Technical Information (OSTI; http://www.osti.gov/bridge/ ), and the

Hanford Declassified Document Retrieval System (HDDRS;

http://www2.hanford.gov/ddrs/index.cfm ).

2. Values for exposure level (intensity), frequency, and duration were assigned to each task-specific

exposure scenario identified in Step 1.

3. Potentially exposed workers were identified using available employment information on job

assignment (e.g., job title, division, and department) and work location in each study facility (e.g.,

area, building, and room).

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4. Workers were linked to exposure scenarios and scores were calculated by an algorithm that uses the

information from Steps 1-3 and any modifiers to account for temporal trends in exposure.

5.7.1 Records collection and abstraction

The roster of individual study subjects excluding case status was matched to personnel records obtained from previous studies and from each study site to abstract relevant employment information.

Primary data sought were job titles, work locations, job descriptions, and functional organizations for the time period of interest.

Various documents were obtained to identify the presence, location, and use of benzene at each

study site. These records included: 1) available industrial hygiene measurements; 2) process

descriptions; 3) organizational structures; 4) various memos, procedures, reports and miscellaneous

documents related to site operations as well as safety and health; 5) environmental remediation

documents; 6) chemical/hazard inventory documents; and 7) facility design documents. Information was

abstracted that supported the spatial and temporal identification of work activities involving benzene

use.

5.7.1.1 Mihlan data

In the mid-to-late 1990s, NIOSH researchers received a set of electronic files pertaining to an

assessment of chemical exposures to workers who were employed at LANL, ORNL, Hanford, and SRS

[Mihlan 1997]. The assessment was conducted by researchers from the University of North Carolina,

Chapel Hill, who were examining the risk of multiple myeloma from ionizing radiation exposure and

other chemical and physical agents [Wing et al. 2000]. The files comprised an industrial hygiene

database that was populated with monitoring information collected during extensive searches of site

records. In developing the database, preference was given to entering results for agents of interest in

the epidemiologic study, which included aromatic hydrocarbons, among others. These searches and the

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subsequent database development were performed under the direction of Gary J. Mihlan. The monitoring data (hereafter referred to as the Mihlan data), were used in the current study to identify potential sources of benzene exposure and identify affected workers.

Of the 30,125 industrial hygiene sample observations in the Mihlan data, 213 were benzene related (0.7%). Most of the benzene records (78%) originated from measurement data abstracted from the ORNL Industrial Hygiene Workplace Sampling Data Sheets (Green Sheets). A summary of the information is provided in Table 5-26.

Table 5-26. Benzene monitoring information found in Mihlan data [Mihlan 1997]

Hanford LANL ORNL SRS Observations 4 29 175 5 Period covered 1981 -1982 1954 -1974 1973 -1990 1978 -1983 Range (ppm) 0.058 -1.0 ND -27.5 ND -949.1 ND -20.0 ND=not detected

5.7.2 Estimates of exposure level, frequency, and duration

Qualitative exposure levels were assigned to activities identified with potential benzene exposure based on the assessment of available records. Most industrial hygiene data were qualitative; however, there were a limited number of quantitative results that proved useful for judging exposure levels. Typically, quantitative benzene results were reported in units of parts per million (ppm) or in milligrams per cubic meter of air (mg∙m -3), where 1.0 ppm was assumed equivalent to 3.2 mg∙m-3 (at

25⁰C and 1 atmosphere). Few observations appeared to be adjusted for worker stay times, thus most

results were assumed to be instantaneous levels.

NIOSH recommended exposures levels (RELs), OSHA PELs, and threshold limit values (TLVs)

recommended by the ACGIH provided information on the relative severity of benzene exposures

measured in the workplace (Table 5-27). Values for TWA concentration assumed up to a 10-hour

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workday during a 40-hour work week. The short-term exposure limit (STEL) was defined as a 15-minute

TWA exposure that should not be exceeded at any time during a workday.

Table 5-27. Recommended benzene exposure levels

Exposure limit (ppm) Source TWA STEL Reference NIOSH REL 0.1 1 NIOSH Pocket Guide [NIOSH 2007] OSHA PEL 1 5 OSHA 1910.1028 [OSHA 2008] ACGIH 0.5 2.5 ACGIH TLVs  and BEIs  [ACGIH 2004]

Exposure levels of high, medium, low, very low, and none were assigned with reference to the

NIOSH STEL (i.e., 1.0 ppm or approximately 3.2 mg∙m-3). These values are shown in Table 5-28.

Table 5-28. Benzene exposure levels

Exposure Level Numeric value s Relevance High > 1.0 > NIOSH STEL Medium 0.5 -<1.0 > ½ NIOSH STEL < NIOSH STEL Low 0.1 -<0 .5 < ½ NIOSH STEL Very Low 0.01 -<0.1 above background None <0.0 1 minimal potential for exposure above background

Exposure levels were assigned using expert judgment of exposure potentials by job and task. In general, worker exposure levels were determined using available monitoring data and a hierarchical approach to worker categories, whereby a “performer” is presumed to receive greater exposure than a

“supervisor”. A description of the general worker categories in the order of exposure potential is shown in Table 5-29.

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Table 5-29. Worker categories

Category description Hanford Examples Performer Worker performing the -Painter, 1955 (Painting 1940-1969) exposure activity -Separation Process Operator, 1957, Building 234-5Z, (Recuplex) Supervisor Supervisor of the worker -Shift Supervisor, 1946, 300 Area (Frost Testing) performing the activity -Area Supervisor, 1950, Building 234-5Z (Plutonium Shaping Operations) Maintenance Worker providing maintenance -Pipefitter, 1947, 300 Area (Technical Radiochemistry Lab) and repair support services for -Steamfitter, 1952, Building 234-5Z (Plutonium Shaping an exposure activity Operations) Other support Worker providing support -Engineer Hygiene, 1952, 300 Area, Org Code HD, (Cold other than maintenance and Chemical Semi-works) repair for an exposure activity -Rad Monitor, 1954, 200W Area (Plutonium Finishing Plant (Z Plant) Laboratory) Building Worker co-located in a building -General Clerk, 1959, Building 327 (Radiometallurgy Occupant with an exposure activity Building) -Admin Clerk, 1972, Building 234-5Z (Waste Treatment and Americium Recovery Facility)

Expert judgment was also used to estimate values for exposure duration in hours per day (hrs∙d-

1). Exposures were further modified to account for fractionation by multiplying by an exposure frequency coefficient calculated as a fraction of periods per 240-day working year (i.e., daily=1, weekly=0.2, monthly =0.05, and quarterly=0.02).

5.7.3 Exposure score algorithm

A set of task-specific exposure scores were assigned to each study participant based on employment records. A task-specific exposure score is the product of exposure level ( Li), duration ( Di), frequency ( Fi), and year fraction ( Yi) assigned to the ith task. The value of Y i is defined as the ratio of the days worked in a task to the days available in the year, thus the annual exposure score (AES) for the ith study subject is sum of exposure scores from n tasks conducted within the jth year weighted by employment time. The coefficient ti, represents the set of mitigating factors used to reduce average

benzene exposures over time. The time-dependent weights used by Henn et al. [2007] were selected as

default values, although values of ti may be modified based on available task-specific information. The

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cumulative exposure score (CES) is the sum of annual exposure scores between the years of first and last exposure (i.e., 1, 2… k years, where k=year of last exposure). Alternatively, the CES for the lth study

subject is mathematically expressed as:

k n

CESl = ∑t j ∑Li FiDi Yi (5-3) j=1i = 1

The CES is adjusted for the study design, whereby the duration in the year of last exposure was based on the time of the attained age of the index case, subtracting any exposure lag period.

5.7.4 General Assignments

Efforts were made to define task-specific exposure potentials and associate tasks to individual workers. Nevertheless, details were often unavailable to precisely estimate the time of the task, the potential for benzene exposure, or the presence of specific personal performing the task. Therefore a set of default assignments were developed that focused on operational, research and development, and laboratory task classification. In constructing default values, the following general assumptions were

applied:

1. A hierarchy in exposure potential existed such that operational activities were associated

with the highest potential for exposure, followed by developmental activities.

2. Performers had the highest potential for exposure.

3. Maintenance workers were exposed less frequently than performers, but exposures

involved higher concentrations because of the potential to breach containment.

4. Supervisors performed periodic oversight of activities and had similar but less frequent

exposures compared to performers.

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5. Some benzene-related processes (e.g., rubber molding) may involve elevated exposures to

building occupants although generally these persons were removed from the activity.

Workers in this category were assumed to have a very low exposure potential.

6. Support personnel working nearby the work activity were to have been infrequently

exposed.

The list of default values are shown in Table 5-30 and Table 5-31. The values were assigned based on a general knowledge of modifying factors such as PPE use, engineering controls, or other hazard mitigation techniques used across sites. In the event that site-specific exposure information is available or the assumptions used to develop the default values are not applicable, then values were adjusted accordingly on a site basis. For example, overall exposures levels were thought to be higher than default values in PNS areas involving non-radioactive rubber vulcanizing and molding activities.

Table 5-30. Default assumptions for calculating exposure scores for operations and developmental activities involving known or suspected of benzene exposure.

Task Type Worker Category General Exposure Duration Frequency Level (hrs ∙d -1) Operational Performer Low 8 Daily Maintenance Med 1 Weekly Supervisor Low 1 Daily Building Occ Very Low 8 Daily Other near Low 1 Monthly Developmental Performer Low 4 Daily Maintenance Med 1 Weekly Supervisor Low 1 Daily Building Occ Very Low 4 Daily Other near Low 1 Monthly

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Table 5-31. Default Time-dependent Weighting Factors (TWF) [2007]

Time -period TWF

(ti) <1950 1 1950s 0.7 1960s 0.5 1970s 0.3 1980s and thereafter 0.2

It was also evident that work in some laboratories and tasks involving painting, paint removal, gasoline dispensing, and automotive repair shared similar benzene exposure potentials across sites.

Therefore standardized exposure scores were developed for these activities, which are discussed in the following sections.

5.7.4.1 Fuel dispensing and automotive repair

Gasoline sold in the U.S. is a complex mixture of volatile hydrocarbons that includes about 1-4% benzene [Verma and des Tombe 2002]. The concentration of benzene in gasoline has not varied markedly over the years since the removal of lead-containing anti-knock compounds in the 1970s, which led to a slight increase in aromatics (including benzene) for antiknock purposes [Karakitsios et al. 2007].

Diesel fuel also contains benzene, but at much lower concentrations (0.02%). Workers tasked with

refueling operations can be vulnerable to relatively high concentrations of gasoline vapors. Vehicle

mechanics are also at risk of benzene exposures from vehicle emissions, contact during repairs to fuel

systems, and the widespread use of gasoline used as a solvent for degreasing [Capleton and Levy 2005;

Hotz et al. 1997; Hunting et al. 1995; Popp et al. 1994]. Preventative measures implemented over the

last decade, such as garage ventilation, vapor capture systems, and automated dispensing systems, have

reduced benzene concentrations to levels that are typically well below the NIOSH REL in most

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occupational settings [Periago and Prado 2005]. Nevertheless, benzene exposures in fuel station attendants and vehicle mechanics are still elevated compared to the general population [Karakitsios et al. 2007]. Inhalation is the most likely pathway, although dermal exposures may be likely in certain settings (e.g., hand washing, parts cleaning, and opening fuel systems). Mechanics appear particularly susceptible to dermal exposures while working on gasoline engines and, especially, on fuel tanks

[Nordlinder and Ramnas 1987].

5.7.4.1.1 Exposures during fuel dispensing

Williams et al. [2008] reported benzene eight-hour TWA concentrations in workplaces involving gasoline ranging from 0.005 to 0.24 ppm taken from six studies conducted from the late 1970s to the early 2000s. Periago and Prado [2005] reported mean TWAs during refueling at European service stations of 0.23 ppm (range 0.08-0.50) in 1995, 0.075 ppm (range 0.04-0.14) in 2000, and 0.05 ppm

(range 0.01-0.18) in 2003. Compared to gasoline, benzene exposures from working with diesel fuels is relatively low. For example, Williams et al. [2008] reported benzene concentrations ranging between

0.001 and 0.008 ppm in studies of jet fuel exposures conducted in the late 1990s and early 2000s.

Data on measurements made in earlier years are sparse. Parkinson [1971] examined benzene concentrations at nine UK service stations and found mean concentrations ranging from 0.3 to 2.4 ppm from individual samples ranging from 0.2 to 2.9 ppm. NIOSH studies conducted in the mid-1970s found airborne benzene concentrations ranging from non-detectable to 0.32 ppm in some U.S. service stations

[Runion 1977]. Similarly, McDermott and Vos [1979] reported on TWA exposures in seven U.S. service stations that were examined in 1976; all locations reported average levels less than 0.11 ppm, with individual samples ranging from ND to 0.84 ppm.

Some industrial hygiene measurements were made at automotive work locations at Hanford and SRS. Benzene samples collected while dispensing gasoline at Hanford on three different days in

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1981 indicated TWA exposures ranging from <0.031 ppm benzene to 0.085 ppm benzene [Mihlan 1997].

A similar survey of gasoline station attendants at SRS in 1978 resulted in average and peak benzene concentrations of 0.27 and 0.74 ppm, respectively [Croley 1978].

5.7.4.1.2 Mechanic exposures

Site-specific exposure information was not available; therefore, exposures were characterized using information published in open literature. Javelaud et al. [1998] reported a shift- averaged (± SD) benzene concentration of 0.15 ppm ± 0.47 from personal air samples collected from automobile mechanics ( n=65) employed in 23 European garages. Two mechanics (3.1%) were exposed to benzene in

excess of 1 ppm (2.4 and 2.9 ppm). Handwashing with gasoline during the week of air sampling was

reported in five car mechanics (7.7%). In another study, workshift area air sampling conducted in 26

civilian and four army automotive garages in the mid-1990s resulted in a median shift-averaged benzene

concentration of 0.01 ppm (95% CI: ND, 0.14) [Hotz et al. 1997]. Popp et al. [1994] collected air samples

from automotive workplaces in areas where workers (n=20) were performing various tasks. They reported average and maximum benzene concentrations of 0.81 and 4.1 ppm, respectively. Laitinen et al. [1994] conducted personal air monitoring and collected blood samples from Finnish automobile mechanics (n=10) in 1992. Their study reported maximum short-interval benzene exposures in the breathing zone of 1.36 ppm and 3.7 ppm, for unleaded and leaded fuels, respectively. Eight-hour TWA concentrations ranged from 0.1 ppm to 0.48 ppm. Benzene concentrations from stationary samples in garages were below 0.2 ppm. Analyses of biological markers in blood suggested that air sampling results tend to greatly underestimate actual benzene exposures. Laitinen et al. [1994] attributed this bias (up to eight-fold for some tasks) to the high frequency of dermal exposures in most tasks examined. Only two of the 10 mechanics under observation lacked significant dermal exposures during the study period.

They concluded that benzene exposure through the skin was the main exposure pathway for these

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mechanics, which accounted for up to 80% of their benzene uptake. Nordlinger and Ramnas [1987] examined benzene level in seven garages in Sweden. Over 100 area air samples were conducted, resulting in average eight-hour TWA concentration ranging from 0.13 ppm to 2.13 ppm. Concentrations were generally higher in small garages compared to larger ones. Benzene levels were also elevated in winter months compared to the summer. Breathing zone samples were highest during gasoline tank draining (14.4 ppm), followed by cylinder head dismantling (8.9 ppm) and fuel injection adjustment (2.2 ppm).

5.7.4.1.3 Assignments

Workers dispensing fuels and performing automotive repair were assumed to be exposed to fuel vapors resulting in above background (albeit low) exposures to benzene. Exposure levels to maintenance personnel working on dispensing systems may have been higher for shorter periods of time, although it is expected that maintenance on fuel dispensing systems would be infrequent and maintenance workers covered many other facilities (Table 5-32).

Table 5-32. Annual Benzene exposure scores for gasoline dispensing and automotive repair.

Worker category Exposure level Duration Frequency Score (hrs ∙d -1) Performer (mechanic, attendant ) 0.5 6 1 3 Supervisor (mechanic, attendant ) 0.5 1 1 0.5 Building Occupant 0.05 6 1 0.3

5.7.4.2 Painters

Paints, lacquers, varnishes, and thinners are complex mixtures of potentially hazardous substances comprising pigments (organic and inorganic), binders (resins), additives (e.g., surfactants,

driers, and biocides) and solvents. Occupational exposures to these substances have been determined

to be carcinogenic (IARC Class 1) based primarily on the evidence of increased risk of cancers of the lung,

and of the urinary bladder in painters and workers manufacturing paint products [IARC 2010]. The

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existing evidence on hematopoietic cancers in painters was far less compelling and deemed inadequate by IARC for evaluating leukemia risk.

Inhalation is the predominant exposure route, followed by dermal absorption to a much lesser extent. Exposures stem mainly from volatile solvents, although the mechanical removal of paint (e.g., chipping and scraping) can result in inhalation of pigments and fillers. The use of PPE can substantially reduce exposures. Nevertheless, widespread use of respirators and gloves is not common among painters, especially in earlier years [IARC 2010]. Solvent exposure measurements in painting occupations have varied widely by agent and setting. Agents most frequently reported are formaldehyde, naphtha, toluene, trichloroethylene, and xylene. Average airborne benzene concentrations from paints and paint products are typically less than 0.1 ppm since the 1980s [Williams et al. 2008]. Studies reporting on benzene concentrations were sparse and most were related to spray painting in the automotive repair industry [Jayjock and Levin 1984; Uang et al. 2006; Vitali et al. 2006; Winder and Turner 1992]. There were two studies that appear most relevant to painting in the study facilities. Mølhave and Lajer [1976] reported on exposure to benzene (55 ppm) and trichloroethylene (91 ppm) during construction painting.

The benzene exposure reportedly originated from turpentine used for thinning, equipment cleaning, and hand washing [Molhave and Lajer 1976], which are likely practices in study facilities, especially in the earlier years. Some shipyard painters working in the confined space of large shipholds in the early

1970s were exposed to levels of up to 11 ppm benzene, 88 ppm toluene, and 538 ppm xylene [Mikulski et al. 1972]. However, benzene was not detected in work areas where total toluene and xylene concentrations did not exceed 100 ppm (Table 5-33).

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Table 5-33. Benzene air samples in painting tasks

Sampling type Results (ppm) Task and duration Year n mean Max imum Reference indoor painting personal, 1 h 1974 41 55 289 [Molhave and Lajer 1976] spray painting personal, 1 h 2006 39 0.4 1 NR [Uang et al. 2006] confined space area, 1 h 1972 62 9 11 [Mikulski et al. 1972] spray painting personal, 4 -7 h 1992 3 0.3 0.3 [Winder and Turner 1992] varnishing personal, 4 h 2006 8 3.1 16.6 [Vitali et al. 2006] spray painting personal, short -term 1984 5 0.6 1.1 [Jayjock and Levin 1984] NR=Not reported; ppm=parts per million 1The result of three plant averages weighted by the number of personal samples in each plant. 2Author stated “At least six air samples” for each area sampled. Results shown apply to the group where the sum of toluene and xylene exceeded 200 ppm.

Painting was conducted routinely throughout all study facilities using solvent-based paints and thinners intended for industrial applications. Organic-based solvents in coating products for industrial applications contain aromatics (e.g., benzene, toluene, xylene, trimethylbenzene) in concentrations that have varied widely by product and over time. Heavy naphthas, toluene, and benzene were common solvents during the 1930s. Substitutes for aromatic hydrocarbons, including turpentine, decaline and tetraline, were more prevalent in later decades. Although benzene use was widely discontinued during the 1950s, some legacy products and commercially available benzene substitutes contained appreciable amounts of the toxic chemical. For example, documentation was available indicating that a paint removal product (Lightning Paint Remover) used at LANL in 1954 contained up to 32% benzene [LANL

1967]. It was reported that commercial toluene, a widely used benzene substitute in solvent-based paints, contained about 10% benzene in the 1950s and 1960s, around 0.5% in the 1970s and fell to <

0.01% thereafter [IARC 1999].

Benzene measurements collected during painting operations in the study facilities were sparse.

A 1943 survey at PNS reported benzene airborne concentrations in excess of acceptable limits by two- to-four fold (i.e., 75 ppm at time of survey) while using paint solvents and thinners [Kroger 1943]. A

1956 survey of benzene, toluene, and xylene levels during the application of a protective coating in SRS

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Building 234-H reported that levels were “within recommended atmospheric concentrations” [Croley and Schroder 1956]. At that time, recommended levels were: 35 ppm benzene, 200 ppm toluene, and

200 ppm xylene. Painters wore airline respirators during this application. The Mihlan data included some benzene data from breathing zone samples that were collected from ORNL painters. These data were from limited sampling ( n=5 observations) in Building 3063 during painting operations that spanned one week in December, 1986. All sample results were below detection [Mihlan 1997].

Benzene exposures were assumed moderate during early painting operations. Pure benzene was not expected to be available or used in painting activities at any study facility. Exposures are assumed to decrease over time according to default TWFs. It was further assumed that exposures were limited to job assignments as painters, woodworkers, cabinet makers, and their immediate supervision

(Table 5-34).

Table 5-34. Annual benzene exposure scores for painters and painter supervisors

Worker category Exposure level Duration Frequency Score (hrs ∙d -1) Performer 0.5 6 1 3 Supervisor 0.5 1 1 0.5

5.7.4.3 Laboratory workers

Many laboratory analytical procedures were known to involve occasional use of small quantities of benzene, especially before the widespread use of benzene substitutes by the late 1960s. Industrial hygiene records were sparse although there were indications of reagent grade benzene in laboratory inventories and limited air sampling data from surveys conducted to determine benzene exposure potentials. There was evidence of benzene use in all study facility laboratories. For example:

• A 1949 Hanford industrial hygiene survey of benzol (benzene) concentrations reported average

concentrations over 100 ppm during laboratory (Bldg. 726) glassware washing operations [Adley

1950a; Gill 1950].

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• Benzene is listed in the Hanford chemical inventory for the Radiometallurgy Building, Post

Testing laboratory (Building 327) from 1953 to 1976 and the Biophysics Laboratory Building,

Radiochemistry Laboratory (Building 329) from 1952 to 1976 [Klem 1990].

• Benzene sampling was conducted in 1960 in a LANL laboratory supporting the CMB-6 Plastics

Group in the Sigma Building [LANL 1967] and in 1962 for chemical extraction by J-11 workers

(Radiochemistry Lab). In the latter case, benzene concentrations from seven air samples ranged

from 50 to 80 ppm during a 15-minute period in the general area of an extraction operation that

lasted about one hour. No other details were available [LANL 1967].

• In the 1970s, small quantities of reagent grade benzene solutions were found during industrial

hygiene surveys in the main analytical laboratories (i.e., 772-F and 773-A) of SRS [Mihlan 1997].

It is likely that these reagents were also used in the earlier periods of operations. In addition,

chemists in the TNX Building Engineering laboratory (677-G) used benzene in experiments

during vitrification piloting under the SRS Waste Solids Program (1980s to mid-1990s) [Palmiotto

1982].

• In 1973, the Carbon Technology Group in the Metals and Ceramics Division (4508) and the Bio-

Organic Group in the Analytical Chemistry Division (4500S) were recognized as users of “sizable”

volumes of benzene at ORNL [Bolton et al. 1974]. Air concentrations ranged from ND to 30.0

ppm benzene, although all eight-hour TWA samples were below 1.0 ppm.

• During a 1977 inspection to uncover any benzene at PNS, safety staff found approximately one

liter of pure benzene in the Quality Assurance Laboratory. The benzene was controlled under

laboratory procedures that included hooded ventilation. Immediately following the inspection,

the benzene was replaced by a less hazardous solvent [McDonough 1977].

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Arguably the most pervasive application of benzene was in laboratory solvent extraction procedures. In particular, benzene was a common diluent of TTA, which is a chelating agent used for the solvent extraction of plutonium and other metals. Hanford chemists used TTA-benzene to extract plutonium from urine samples for radioactive assay (1946 to mid-1960s). The solution was typically prepared by dissolving 5 g of 2-thenoyltrifluoroacetone in 100 ml of benzene. Exposures were likely during preparation of TTA, TTA analyses, and the cleaning of laboratory glassware. INL (1950s-1962) and

SRS (1952-1959) adopted Hanford’s TTA solvent extraction procedures for their early plutonium bioassay programs. LANL also used TTA for plutonium analysis from 1957 to 1963 [Campbell et al. 1972];

however, the TTA solution was prepared using toluene as the diluent [Milligan et al. 1958].

Nevertheless, there were reports on other LANL radiochemistry procedures involving TTA-benzene

extractions [Farr et al. 1953; Kleinberg and Smith 1982; Potratz 1954]. Similarly, ORNL chemists did not

use TTA extraction for plutonium bioassay, but methods using TTA-benzene for extracting other metals

(e.g., thorium, zirconium, and hafnium) were evident in the early 1950s [Swartout et al. 1951]. By the

late 1950s, ORNL researchers preferred xylene as the diluent, if applicable, because of its higher boiling

point, lower flammability, and reduced toxicity compared to benzene [Moore 1958].

A combination of engineering controls, rigorous procedures, and the use of small quantities

greatly reduced benzene exposures in study facility laboratories. Because of its toxicity and flammability,

benzene was used sparingly in procedures confined mostly to bench scale processes performed under

hooded ventilation. Many of the laboratory procedures involved chemical extraction of radioactive

compounds; therefore, additional controls were implemented to reduce radiological hazards.

Nevertheless, there was evidence of significant airborne benzene concentrations from some early

laboratory procedures. Laboratory exposures were often characterized as infrequent and involving high

concentrations over short periods of time. Assignments were limited to laboratory workers and their

immediate supervisors. Scores were modified using default TWFs to account for the reduction of

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benzene use and continued hazard mitigation over time. The general scores shown in Table 5-35 cover all laboratory work except for early Hanford plutonium bioassay procedures described in the following section.

Table 5-35. Annual benzene exposure scores for general laboratory workers and supervisors

Worker category Exposure level Duration Frequency Score (hrs ∙d -1) Performer 1 2 0.2 0. 4 Supervisor 1 1 0.05 0.0 5

5.7.4.4 Hanford

There was little evidence of significant benzene exposures from operations at Hanford other than the general uses previously described for automobile servicemen, painters, and laboratory workers. A notable exception was exposure to workers in Buildings 706 (1946-1953), who developed and implemented plutonium bioassay procedures using TTA in benzene. During this time, significant and potentially acute benzene exposures were indicated, but limited to less than ten Hanford laboratory workers assigned to the facility.

5.7.4.4.1 Bioassay Laboratory (<1954)

First developed in 1945 at the Crocker Radiation Laboratory in Berkeley, CA [Parker and Healy

1945], TTA extraction methods for plutonium bioassay were put into full-scale use at Hanford in early

1946 as a means to improve plutonium recovery and increase urinalysis sensitivity [Healy 1946].

Operations began in the old Bioassay Laboratory (Building 706), which was a one-story frame building that consisted of the main laboratory, a fluophotometer room, office, and locker rooms. The facility was initially staffed with four “non-technical” workers without prior experience in chemistry [Healy 1946] and grew to six laboratory workers and three office employees by 1950 [Gill 1950].

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In general, the early TTA analysis methods involved three successive additions of benzene solution to each separatory container (glass funnel), cleaning the container in benzene, and rinsing the container with a solution of benzene in alcohol and hot water. The latter operation resulted in significant releases of benzene vapor into the laboratory atmosphere [Gill 1950]. Two laboratorians conducted bench-top TTA-extractions using hooded ventilation. Glassware cleaning was performed with rubber gloves to prevent dermatitis, although mechanical ventilation was not recommended until 1950

[Gill 1950]. Breathing zone benzene concentrations were greatest during glassware cleaning, whereby measured concentrations ranged from 97 to 170 ppm during the 30-minute operation in 1950 [Gill

1950]. The highest concentrations (370 – 1000 ppm) were observed in 1953 while pouring discarded benzene solutions into a sink [Perry 1953]. Other TTA activities resulted in airborne levels that were typically less than 30 ppm. Most activities were short in duration.

Urinalysis procedures were implemented by the medical department to assess acute benzene exposure in bioassay laboratory workers, whereby a depressed ratio of organic sulfate to inorganic sulfate in urine was used as a marker for benzene absorption [Kammer et al. 1938]. Sulfate ratios that were less than normal values in unexposed individuals (i.e., ∼0.8) indicated formation of conjugated sulfate metabolites following benzene assimilation. Samples collected in early 1950 from three of five affected workers yielded sulfate ratios ranging from 0.61 to 0.75 [Adley 1950a]. These results prompted procedural changes that substituted tributyl phosphate (TBP) for benzene as a glassware cleanser; however, benzene continued to be used as a TTA diluent. All sulfate ratios in samples collected from workers immediately following the procedural changes were normal [Adley 1950b]. Additional urine samples were collected following the high benzene concentrations reported in 1953. The results from these samples were also within the normal range, suggesting that benzene uptake was minimal even though airborne benzene concentrations were still elevated. The analyst attributed the low benzene absorption to the short duration of vapor exposure during the pouring operation. Nevertheless, the

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potential for acute benzene exposure was evident while working with benzene in the laboratory prior to

1953.

A detailed investigation into substitutes for benzene was conducted in 1953 [Silker and

Schwendiman 1953]. Both xylene and Deobase (acid-purified “deodorized” kerosene) were identified as

suitable replacements; however, evidence confirming the removal of benzene from bioassay procedures

was not available until the late 1960s. Nevertheless, there was no evidence of persisting high benzene levels in the laboratory after 1953, which was the likely result of improved exposure controls following the 1953 industrial hygiene survey. Furthermore, a new Bioassay Laboratory (Bldg. 747) with improved engineering controls was operational by May of 1954 [Parker 1954]; therefore, significant laboratory benzene exposures were less likely after this time (Table 5-36).

Table 5-36. Benzene exposure scores for plutonium bioassay in Building 706 prior to 1954.

Task Worker category Exposure level Duration Frequency Score (hrs ∙d -1) TTA -benzene performer 100 0. 50 1 50 TTA -benzene supervisor 10 0. 5 0.2 1

5.7.4.5 INL

Available records suggested that benzene was used sparingly at INL. For example, solvent use was examined in the 1970s for a study of nonradioactive waste disposal methods. Researchers used procurement records to derive estimates of the solvent wastes that were being generated at INL, primarily in CFA, CPP, TRA, NRF, and TAN areas. Based on the 1970 purchases, annual solvent use exceeded 500,000 liters; however, less than 40 liters (<0.01%) were attributed to benzene [Commander

1971]. The report did not specify benzene use, although laboratory use in small quantities as a diluent was likely. Likewise, early experimentation of benzene as a diluent for TBP-based metal recovery at the

ICPP may have occurred, although operational records indicating its use could not be found. It is likely that ICPP procedures involving benzene (if existing) were limited to pilot studies that were discontinued

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early on in ICPP development, thus exposures would have been limited to a few personnel in laboratory settings during the early 1950s.

NIOSH obtained a large database of Industrial Hygiene monitoring data from INL in 1995. The database, known as the INL PASAIR database, consisted of information on personal and area air sampling conducted at the site from 1989 to 1995. There were 18,471 observations in PASAIR; however, only 118 observations provided some information on benzene exposures. Most benzene air monitoring was related to drilling operations in support of environmental characterization (n=40). Samples were also available for automotive repair occupations (n=23), construction trades ( n=9), and laboratorians

(n=7). All monitoring results indicated eight-hour TWA benzene concentrations that were far below the

NIOSH REL. Some of the benzene information in PASAIR is described below:

• PASAIR contained limited benzene exposure information from eight-hour TWA air samples taken

while monitoring workers in automotive repair areas at CF-665, IF-1012, CPP-655, and TAN-609

between 1989 and 1995. All sample results were reported as less than detectable, although the

detection limit was not stated.

• Benzene monitoring was conducted in CF-625A during laboratory work involving use of benzene

for preparation of laboratory standards. All eight-hour TWA results were reported as less than

detectable. In 1991, three air samples were collected during the packaging of laboratory wastes

at IF-603.

• Several benzene air samples were obtained during well drilling and soil sampling work in

support of environmental characterization. All benzene results were non-detectable, suggesting

that benzene contamination levels in soil and groundwater were too low to provide a credible

source of potential occupational benzene exposure.

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In summary, there was no evidence of large-scale benzene use at INL. Based on the lack of known industrial applications and the available (albeit sparse) evidence on benzene exposures, it was concluded that INL benzene exposures were limited to tasks covered by the general assignments, i.e., laboratory work, painting, gasoline dispensing and automotive repair.

5.7.4.6 LANL

Information on benzene use is sparse but sufficient to indicate exposures were likely in some occupations dating back to the late 1940s. For example, the 1949 annual report from the Health Division

(H-Division) stated that a several Los Alamos Laboratorians handled benzol [Shipman 1950]. From the report, it is evident that benzene hazards were acknowledged by Health Division staff in the early years and that the use of the product necessitated some level of monitoring and control. However, the report did not specify which occupations were likely to be affected, or the purposes of benzene use during this time. Information on the locations of early benzene activities was also lacking although related reports discussed sulfate in urine analyses for some workers exposed to benzene at S-Site [Shipman 1949a;

Shipman 1949b; Wipple 1948]. None of these reports provided Information on airborne levels, bioassay results, or controls in place; however, it was suggested that the number of exposed workers was small

(<10) and that acute exposures were not evident from the sampling results.

Some early industrial hygiene investigations examined benzene in a few work divisions including: Chemistry and Metallurgy (CMB-3, CMB-6, CMF-9, and CMF-13 Groups); Nuclear Rocket

Propulsion Division (N-1 and N-5 Groups); Experimental Physics Division (P-1, P-9, P-14 and P-15

Groups); Field Testing (J-11 Radiochemistry); and Interaction of Explosive Metals Division (GMX-2, GMX-

3 and GMX-11Groups) [LANL 1967]. There was evidence of laboratory use of benzene in small quantities in a variety of laboratory activities (e.g., solvent extractions, as a reagent, as a rubber solvent for slide preparation). LANL scientists investigated the use of TTA-benzene for plutonium recovery processes at

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DP West in 1947; however, its use was abandoned by 1948 because of poor compatibility with most

LANL plutonium solutions [Christensen and Maraman 1968]. Site records did not reveal evidence of benzene in bulk quantities, although some production activities in the 1950s may have used liter quantities based on descriptions in industrial hygiene records. Lists of toxic chemical inventories maintained in Stores and Warehousing (SP Groups) suggested that annual benzene use ranged from 200 kg in 1972 to 20 kg in 1978. Environmental documents did not indicate large releases of benzene from

LANL operations although there was mention of annual benzene releases of <200 kg in later years

[Farenbaugh 1981; Gunderson et al. 1985]. For example, a 1981 letter stated that benzene releases from laboratory ventilation was about 90 kg∙y -1 [Farenbaugh 1981].

5.7.4.6.1 Cleaning Leak Detectors (1944-1962)

An activity performed by Group P-1 that was identified as” cleaning leak detector machines”

was monitored by industrial hygiene for benzene concentrations from the early 1950s to 1962 [LANL

1967]. In 1954, monitoring results were reported to be above tolerance. Similarly, a survey in 1959

reported that two workers cleaning detectors were exposed to benzene concentrations that were

several times tolerance, although the workers wore respirators [Shipman 1959]. Monitoring results in

1962 showed that benzene levels were reduced to below tolerance levels because of ventilation

improvements. Assuming that the tolerance level was equivalent to the ACGIH limit for benzene that

existed at the time of monitoring, the concentrations were >35 ppm in 1954, > 25 ppm in 1959, and <25

ppm in 1962. A detailed description of leak detectors or the actual tasks involved in cleaning the

detectors could not be found; however, it was evident that benzene was used as a solvent during this

time and was likely to have been used in times prior. It is also possible that that this activity continued

after 1962 and that benzene use as a degreaser by P-1 workers was not limited to leak detectors.

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5.7.4.6.2 Plastic Explosives Preparation with Benzene (≤1957)

Tasks performed by the GMX Division Group 3 (GMX-3) prior to 1958 that involved “B-P Mixer” or “Kettle” operations in TA-9 in Buildings 34, 45, 46, and 64 may have resulted in benzene exposures.

Apparently the activity involved the synthesis of the explosive compound trinitrostilbene. Records showed that workers dispensed benzene from a container into a hand-carried vessel used to transport the solvent to the work site. The contents were then poured into a chemical mixing container. Personal air samples taken during pouring activities showed concentrations between 10 ppm and 90 ppm, although TWA values were not reported. General area samples were also collected on one day reporting concentrations ranging from 0 to 30 ppm [LANL 1967]. Although airborne levels exceeding the NOISH

REL were evident, a 1956 report suggested that “organic synthesis” procedures in Building 64 of TA-9 were conducted by workers while wearing chemical cartridge respirators [Shipman 1956].

5.7.4.6.3 Cleaning and Degreasing Bomb Parts (≤1959)

Records suggested that benzene was used to clean weapon components on an open lab bench top prior to 1959. Affected groups were CM, CMR-5, and CMP-5. Monitoring data were not available, but the survey was performed at a time when any benzene activities were being evaluated for proper controls. No other parts-cleaning activities involving the use of benzene were identified [Mihlan 1997].

5.7.4.6.4 Cleaning Lithium Metal (1971-1974)

Cleaning lithium metal was evaluated by LANL industrial hygiene staff IH in 1971 (used by A-1) and 1974 (used by P-9). Lithium metal cleaning in TSL-2 (TA-35) by A-1 was reported as an “abnormal procedure” and resulted in benzene concentrations in the breathing zone that were less than 5.0 ppm, with a maximum concentration of 50 ppm directly above the open container. The task duration was reported to be less than one minute. The last known use of benzene was for cleaning lithium metal samples for accelerator experiments in the Van de Graaff building (TA-3, Room 45) in 1974. This work was conducted by the P-9 Group under hooded ventilation and consumed benzene in small amounts

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(100 ml). Breathing zone samples were not detectable although short-term concentrations under the hood approached 25 ppm [Mihlan 1997].

Table 5-37. LANL task specific benzene exposure score assignments

Task Worker category Exposure Duration Frequency Score level (hrs –d-1) Chemical Leak Detectors Performer 35 2 0.2 14 Chemical Leak Detectors Supervisor 35 0.5 0.05 0.9 Plastic Explosives Prep Performer 50 0.5 1 25 Plastic Explosives Prep Supervisor 10 0.5 1 5 Plastic Explosives Prep Building Occupants 1 0.5 1 0.5 Parts Cleaning Performer 5 1 1 1 Parts Cleaning Supervisor 1 1 0. 2 0. 2 Cleaning lithium metal Performer 5 1 0. 2 1 Cleaning lithium metal Supervisor 1 1 0. 2 0. 2

5.7.4.7 ORNL

Searches of available plant records did not reveal information on benzene use other than small quantities in some laboratory procedures dating back to the early operations period. However, the

Mihlan data provided some insight into other uses of benzene at ORNL that continued into the mid-to- late 1970s. The database contained industrial hygiene information for ORNL from 1972 to 1990 that was abstracted from the facility’s Green Sheets [Mihlan 1997]. The original survey sheets were not available.

TWA values (if any) were not identified. Of 26,698 records in the database, 167 observations were related to benzene measurements. There were 155 (93.4%) breathing zone results and 12 of unidentified sampling type. Air concentrations ranged from ND to 949.1 ppm benzene. Detection levels for breathing zone samples ranged from 0.008 to 2.5 ppm. There were 125 (74.8%) measurements above detection, with mean and median air concentrations of 19.4 and 2.5 ppm benzene, respectively.

The highest benzene level (949.1 ppm) was measured in 1977 and was identified as a breathing zone sample for a machinist in the research and fabrication shops (Dept 3016). Further information on the

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work activity was not available and there was no evidence of followup samples taken as a result of the high level, although elevated trichloroethylene was also indicated (397.7 ppm) as a coexposure.

The sample sheets also provided some information on department, job title, and activity (Table

5-38 and Table 5-39), with department information being most complete (97%), followed by job title

(59%), and activity (41.3%). The majority of observations pertained to laboratory workers (55.7%) or work conducted in laboratory departments (94.0%). The remainder of samples were for painters ( n=5), stores workers ( n=4), and machinists ( n=1). Information on actual tasks involving benzene use was sparse; however, a brief description of work activities was available for 69 observations (Table 5-39).

Some of the recorded descriptions were ambiguous; therefore, other available information and expert

judgment were used to clarify meaning (if practical). In lieu of supporting information, the activity is shown as recorded (e.g., “routine use”). The most frequently observed activity ( n=37) involved the mixing, blending, and grinding of carbon pastes used in mold preparation in the metallurgy laboratory

(Dept. 3470) from 1973 to 1976. Additional information on this task could not be found. However, based on the similar activities conducted during this time, the work was likely to be limited to bench-scale and pilot operations that were conducted by a few laboratory professionals using appropriate ventilation and PPE.

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Table 5-38. ORNL benzene samples ( n=167) by department and job title

Dept Description Buildings Job titles n % 3016 research shops & fabrication 3024, 3022, 1000, 2518 machinist 1 0.6 0 3027 Unknown Unknown lab worker 3 1. 80 3063 maintenance 2610, 3022, 2518 painter 5 2.99 3068 I&C section 3500 lab worker 1 0.6 0 3142 Stores 1000, 4500 clerk 4 2. 40 3370 chemical technology 4500 lab worker 21 12.57 NR 1 0.60 3380 ecological sciences 2001 lab worker 5 2.99 3390 analytic chemistry 4500 lab worker 25 14.97 NR 24 14.37 3420 Chemistry 4500 lab worker 5 2.99 3470 metallurgy 2000, 4500 lab worker 5 2.99 NR 40 23.95 3490 health physics 4500 lab worker 5 2.99 3675 target preparation 3025 lab worker 3 1.80 4455 Biology 9027 lab worker 13 7.78 NR 1 0.6 0 NR NA NA Lab worker 3 1.8 NR 2 1.2 Dept=Department; NR= not recorded; NA=not applicable

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Table 5-39. ORNL benzene samples ( n=69) by work activity and department

Activity Dept n % Range (ppm) Coal liquefaction research 3370 1 1.45 <0.18 #226 benzene exp 3370 1 1.45 <0.29 Testing organic vapor monitors NR 1 1.45 <0.1 Solvent extraction procedures 3390 (6) 7 10.14 0.4 – 30 4455 (1) density measurements 3390 3 4.35 1.5 – 2.8 ASTM test 3390 1 1.45 1.0 mix & fill benzene container 3470 3 4.35 6.1 – 79.5 Routine use 3390 1 1.45 0.02 Grinding, mixing and blending carbon paste 3470 37 53.62 0.3 – 90 transferring benzene 3390 9 13.04 <1.0 – 8.1 used as a solvent 3390 (2) 4 5.8 <1.0 – 2.5 3370 (1) NR (1) used to prepare sol ution 3390 1 1.45 <2.5 Dept =Department; NR=not reported; ppm=parts per million in air

5.7.4.8 PNS

Industrial use of benzene was limited to a few tasks, mostly involving painting and working with rubber products prior to 1950. Because of its toxicity and flammability, controls to minimize exposures to benzene were evident by the late 1940s. These controls included benzene substitutes whenever possible, safety personnel involvement in assigning protective measures, and monthly medical examinations for those workers who handled benzene [Murphy 1947]. In early 1956, the receipt, storage, and issue of benzene required the approval of the Executive Officer. This approval was obtained only if a less hazardous substitute could not be found. Furthermore, benzene was authorized only in quantity for immediate use [Carter 1956]. By 1958, procurement of pure benzene had ceased [Gordon

1958], although benzene continued to be used sparingly in laboratories and was present in trace amounts as a component of other materials such as paint, petroleum products, and other solvents.

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5.7.4.8.1 Paint thinner or solvent (pre-1950s)

Given the nature of shipbuilding and overhaul, a wide variety of paint application methods, coating materials, and solvents were used by Shop 71 workers. The main shop was located in Building

27, which managed most brush and spray painting activities throughout the shipyard. Varnishes and lacquers were applied in mechanically ventilated paint booths located in Building 64. Spray painting was also conducted in booths located in Bldgs. 74, 75, and 92. The Field Paint Shop (Bldg. 165) was responsible for dispensing paints used for ship construction. Paint lockers were located throughout PNS

(Bldgs. 27, 60, 64, 75, 79, 92, 150, 165, 174, and 178). In years prior to 1945, many of these paint lockers

were overcrowded and not equipped with mechanical ventilation. The lockers housed several solvents

such as benzene, carbon tetrachloride, and gasoline during this time.

There was evidence of significant benzene exposures from some paint solvents and thinners

during the 1940s. It is also possible that pure benzene was used in some applications. In 1943, average

benzene air concentrations ranging from 20 to 400 ppm were measured in Building 174 (Shop 71,

painters) over a period of several days while conducting routine painting operations [Fuller 1943; Kroger

1943]. The measurements were taken in response to worker complaints of solvent vapors while painting

in “Paint Locker 174”. During the course of their survey, the industrial hygienists recommended the

elimination of several substances based on their toxicity; however, some paints and thinners remained

in service that contributed to the elevated benzene exposures. For example, about one gallon (3.78 l) of

“Paint Remover, Formula #72” was used in a 10-minute bench-top operation three times each day

resulting in benzene airborne concentrations ranging from 200 to of 400 ppm in some work areas.

Residual benzene at the workbench resulted in area concentrations ranging from 20 to 100 ppm [Fuller

1943].

In the case of Building 174, the safety staff noted that work area ventilation was inadequate. It

was also noted that a lack in housekeeping and other poor work practices by affected staff (four to eight

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painters working at any time) were contributing factors. As a consequence, the industrial hygiene staff recommended improvements to work procedures and the use of mechanical ventilation for all painting activities in the building. In particular, it was recommended that bench operations using benzene should be conducted in ventilated paint booths or discontinued entirely [Kroger 1943]. There was no evidence that solvent vapors in Building 174 were representative of conditions in other painting locations.

5.7.4.8.2 Rubber work (pre-1950s)

A potential source of significant benzene exposure involved rubber vulcanizing (cementing) conducted in Building 154 by the transportation group (Shop 02) prior to 1950. Vulcanizing activities reportedly consumed approximately one gallon of benzene each work week. It also seems that the choice of vulcanizing solvent was not limited to benzene; other solvents included methyl ethyl ketone

(MEK), trichloroethylene, and gasoline, among others. The task involved the manual transfer of solvent

from a small container to a cloth, which was then hand-rubbed against the rubber to create a tacky

surface for cementing. There was no indication of PPE use while performing the task. The highest

exposure occurred in the Vulcanizing Room, which was described as a small room that was not equipped

with mechanical ventilation [Kroger and Gray 1943]. In 1943, the PNS medical staff measured benzene

airborne concentrations during routine rubber vulcanizing activities [Kroger and Gray 1943]. These

samples suggested that benzene vapor concentrations ranged from 75 to 360 ppm in occupied areas

during vulcanizing; however, short stay times (i.e., <2 hrs∙d-1) resulted in average eight-hour TWA concentrations of less than 40 ppm. Because airborne concentrations were approaching control levels, mechanical ventilation was recommended for further use of the vulcanizing room. By 1947, rubber vulcanizing had been discontinued in Building 154 and moved to a larger room in Building 158 that was equipped with mechanical ventilation.

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5.7.4.8.3 General use as a solvent

Benzene may have been used without authorization or in less than optimal situations because of its superior performance as a solvent and degreaser. For example, records were found that indicated benzol was used in 1947 by Shop 56 pipefitters and insulators [Murphy 1947]. Industrial hygienists discovered five one-gallon cans of benzol during a routine inspection of the Pipecoverers’ and Insulators’

Annex (Building 56). Four cans were stored in a cabinet and one opened can was found on a work bench.

The Shop 56 workers were using the benzol as an adhesive solvent and for parts-washing without the knowledge of safety personnel. There were no controls in place at the time of use and approvals had not been obtained as required. Although early PNS workers in several trades may have similarly used benzene, it is not likely that unauthorized use was pervasive, especially after 1950, when more stringent controls were put into effect.

Table 5-40. PNS task specific benzene exposure score assignments prior to 1950.

Task Worker c ategory Exposure level Duration Frequency Score (hrs ∙d -1) rubber work performer 200 2 0.2 80 rubber work supervisor 20 1 0.2 4 early painters performer 60 6 0.2 24 early painters supervisor 6 1 0.2 1.2 general solvent use maintenance 1 1 0.2 0.2

5.7.4.9 SRS

The search of available records did not reveal evidence of industrial processes requiring large scale use of benzene. Moreover, benzene was remarkably absent from a comprehensive assessment of

SRS chemical exposures conducted for a previous epidemiologic study [Hickey and Cragle 1985]. In that assessment, Hickey and Cragle [1985] examined exposure potentials existing from 1952 to 1984 for workers involved all facets of SRS operations including: heavy water production, separations, tritium production, reactor operations, fuel and target fabrication, and laboratory operations. Benzene

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exposures appear to be mostly limited to fueling and maintenance of gasoline engines, painting, maintenance using solvents with trace benzene amounts, and laboratory work. Nevertheless, there were some small source terms (other than gasoline) identified in site industrial hygiene records and environmental reports that merited further examination.

5.7.4.9.1 F and H Separations

The process of chemically recovering uranium and plutonium following irradiation was conducted in two main operating areas (F and H). Separations in F Area utilized the PUREX process, whereby irradiated fuel was dissolved in nitric acid and then extracted by tributyl phosphate (TBP) contained in a hydrocarbon diluent, typically kerosene or n-dodecane (n-paraffin). In a report on environmental dose reconstruction, two n-paraffin tanks (Tanks 21 and 22) located in F-221 were listed as benzene point sources resulting in “the highest benzene emissions in 1985” [Till 2001]. This report prompted exposure assignments to separation area workers in the previous study by Schubauer-Berigan et al. [2007a], although no additional information on the origin of benzene in these tanks was uncovered. Moreover, similar sources were not found in H Area, although solvent extraction operations in that area also used TBP-Kerosene. There were no indications of elevated benzene air concentrations in occupied spaces adjacent to the F-221 tanks, nor were data available on benzene concentrations in F or H Areas. Although kerosene contains benzene in trace amounts (0.02%), its use as a diluent was not likely to result in appreciable benzene concentrations in occupied areas. Therefore, benzene exposures were not assigned to SRS separation area workers in the current study.

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5.7.4.9.2 Defense Waste Processing Facility (1970-1996)

The Defense Waste Processing Facility (DWPF) was designed to process high-level wastes by bonding radioactive materials to silicon via vitrification. Research and development of the DWPF was conducted in the CNX/TNX area (600 Area, buildings 677-G, 678-G, 678-T) beginning in the 1970s.

Construction of pilot facilities began in 1983. The operation resulted in the build-up of benzene vapor in the process tank headspace, which presented an exposure hazard (albeit low) to workers assigned to the DWPF. Surveys conducted during testing (1980s) and active operations (1990s) suggested benzene concentrations in occupied areas were generally less than 1.0 ppm during DWPF operations.

Benzene was also a by-product of processes involving the In-Tank Precipitation Facility (ITPF) that operated intermittently from 1994 to 1996. The ITPF was a pretreatment operation for the DWPF and was designed principally to remove fission products (primarily cesium and strontium) from high- level wastes prior to vitrification. Benzene, phenol, biphenyl, and phenylboric acid resulted from the radiolytic decomposition of tetraphenyborate salts used in the precipitation of 137 Cs. Shift-averaged benzene concentrations from samples taken during development (1970s) and operations (1990s) of the

ITPF suggested negligible exposures to workers, although instantaneous measurements reached a maximum of 15.0 ppm near the tank opening.

5.7.5 Descriptive Results

Of 1,816 subjects selected at least once as a case or control, only 357 workers (24.0%) were assigned a positive benzene exposure score. All other subjects were assumed to be unexposed. Most subjects were exposed while employed at PNS ( n=93, 32.9% of PNS workers) followed by Hanford ( n=73,

14.3%), ORNL (n=68, 32.1%), INL ( n=60, 12.9%), SRS ( n=37, 12.0%), and LANL/Zia (n=29, 17.6%). Three

workers were exposed in multiple facilities. Among the exposed, most were laboratorians ( n=188,

51.1%), followed by automotive workers ( n=75, 20.4%), maintenance staff (n=55, 15.0%), and painters

(n=45, 12.2%). Eleven workers were exposed in two task categories. Also among those exposed,

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cumulative scores without exposure lag ranged from <0.03 to 283.9. The distribution of scores was highly skewed with mean and median values of 9.5 and 1.2, respectively. The mean and median years of exposure were 7.3 and 4.5 years, respectively, and ranged between <1.0 and 53 years. Without an exposure lag, there were 70 exposed cases (19.7%).

Most benzene exposures occurred between the 1950s and 1970s, which is consistent with facility production schedules and the employment histories of the cases and controls (Figure 5-8).

However, the highest individual benzene exposures were observed in the mid 1940s and mid 1950s, because of increase benzene availability and the absence of exposure controls relative to current work environments (Figure 5-9). These early exposures were mostly restricted to PNS, where some workers began their shipyard careers in nonradiological work assignments prior to 1958.

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10 exposed subjects exposed

0 1935 1945 1955 1965 1975 1985 1995 2005 year

Figure 5-8. Subjects assigned benzene exposures by exposure year

5 4.5 4 3.5 3 2.5 2 1.5

Average expsosurescore 1 0.5 0 1935 1945 1955 1965 1975 1985 1995 2005 year

Figure 5-9. Average exposure score among subjects by exposure year

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5.8 Final exposure assignments for Case and Controls

The output from the exposure assessment was used to assign exposures to each of the cases and controls under prescribed exposure lags. Bone marrow dose estimates from all facilities were summed by radiation type over the period under observation to create a set of raw cumulative dose values for each case and control ( n=1,845). There are five exposure variables describing the cumulative

radiation dose (i.e., gamma, tritium, WRX, neutron, and plutonium) and one variable that is used to

describe benzene exposures. Three additional exposure variables were created to facilitate dose-

response modeling. First, the primary exposure of interest is dose from low-LET radiations; therefore, a

variable for low-LET was created by summing gamma, tritium and WRX doses. Two categorical variables

were also developed for sensitivity analyses that describe low-LET dose and maximum plutonium urine

bioassay results (Table 5-41).

The average and median values of the low-LET exposure estimates for cases and controls by

leukemia subtype are shown in Table 5-42. Dose values, on average, were greater in cases compared to

controls; however, median doses were greater in controls compared to cases. All dose distributions

were severely right-skewed. Average doses in the CML grouping were higher than in other subgroups.

The maximum dose accumulated by a single worker from low-LET irradiation was 633 mGy, which was

assigned to a CLL case. In risk sets of the remaining subtypes, the maximally exposed individual within

the subtype was always a control

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Table 5-41. Summary description of exposure information for cases and controls (two-year lag)

Variable Statistic Cases Controls Total Subjects n 369 1,4 76 1,8 45 Gamma (mGy) Mean 24. 4 22.9 23.4 Median 3. 3 2. 8 2.9 maximum 626.3 566.3 626.3 Exposed (%) 367 (99.5) 1,458 (98.8) 1,825 (98.9) Tritium (mGy) Mean 0.2 0.2 0.2 Median 0.0 0.0 0.0 maximum 28.3 85.1 85.1 Exposed (%) 66 (17.9) 227 (15.4) 293 (15.9) WRX (mGy) Mean 2.6 3.3 3.2 Median 0.0 0.0 0.0 maximum 15.4 28.6 28.6 Exposed (%) 175 (47.4) 728 (49.3) 903 (48.9) Low -LET Mean 27.3 26. 3 26. 5 (sum of Gamma, Median 5.8 7.5 7. 2 Tritium, and WRX) maximum 633.8 574.0 633.8 Exposed (%) 367 (99.5) 1,464 (99.2) 1,831 (99.2) Low -LET dose 0-<1.0 mGy 95 (2 5.7) 355 (2 3.9) 448 (24. 3) distribution 1-<10 mGy 123 (33.2) 50 3 (34.1) 62 6 (33.9) (%) 10 -<50 mGy 101 (27. 4) 43 4 (29.4) 53 5 (29.0) 50 -<100 mGy 24 (6.5) 77 (5.2) 101 (5.5) ≥100 mGy 26 (7.1) 109 (7.4) 135 (7.3) Neutron (mGy) Mean 0.4 0. 0 0. 0 Median 0.1 0. 0 0. 0 maximum 4.9 6.9 6.9 Exposed (%) 52 (14.1) 185 (12.5) 237 (12.8) Plutonium (mGy) Mean 2.0 0. 1 0.1 Median 0. 0 0. 0 0. 0 maximum 49.1 20 .9 49.1 Exposed (%) 31 (8.4) 99 (6.7) 130 (7.0) Distribution of unmonitored 25 4 (68. 8) 1,0 37 (70.3) 1,29 1 (70.0 ) maximum 0-<1.7 mBq ∙d -1 86 (23. 3) 340 (23.0) 426 (23.0) plutonium urine 1.7 -<17 mBq ∙d -1 23 (6.2) 81 (5.5) 104 (5.6) bioassay results ≥17 mBq∙d -1 6 (1.6) 18 (1.2 ) 24 (1.3) (%) Benzene scores Mean 0.8 2.1 1.8 Median 0.0 0.0 0.0 maximum 61.0 283.9 283.9 Exposed (%) 69 (18.7) 291 (19.7) 360 (19.5)

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Table 5-42. Effects on dose (mGy) from exposure lagging of Low-LET radiation by leukemia subtype

Cases Controls Non -CLL No Lag Mean 28.1 26.0 Median 5.0 7.1 Two -year Mean 27.9 25.6 Median 4.9 6.9 Five -year Mean 27.2 24.9 Median 4.8 6.4 Seven -year Mean 26.4 24.4 Median 4.8 6.0 Ten -year Mean 24.8 23.4 Median 4.2 5.4 CLL No Lag Mean 28.8 28.7 Median 7.2 9.3 Five -year Mean 28.3 27.9 Median 6.9 8.9 Seven -year Mean 28.1 27.2 Median 6.9 8.2 Ten -year Mean 27.4 26.2 Median 6.2 7.6 Twenty -year Mean 19.5 18.6 Median 2.9 3.8 Thirty -year Mean 11.0 10.3 Median 0.4 0.3 CML No Lag Mean 37.5 30.1 Median 7.7 8.6 Two -year Mean 37.4 29.5 Median 7.7 8.5 Five -year Mean 37.0 28.9 Median 7.7 8.5 Seven -year Mean 36.2 28.1 Median 7.7 7.9 Ten -year Mean 34.3 26.6 Median 3.9 7.6 AML No Lag Mean 28.3 24.9 Median 4.5 6.9 Two -year Mean 28.1 24.5 Median 4.5 6.8 Five -year Mean 27.4 23.8 Median 4.5 6.1 Seven -year Mean 26.8 23.4 Median 4.5 5.9 Ten -year Mean 25.4 22.6 Median 4.4 5.5

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Study inclusion criteria specified that dose records were available for each study participant selected as a case or control ( n=1,816). Nevertheless, nine participants were identified with first radiation exposures occurring after the latest cutoff date. Of these nine, four workers were assigned benzene exposure. Most of the non-radiation exposed workers (n=5) were employed at PNS, where radiation work did not begin until 1959. All those found without radiation exposure were controls. None were selected to more than one risk set or to the same risk set as that of another in the group (Table 5-

43).

Table 5-43. Unexposed cases and controls, no lag period ( n=9)

Year of f irst radiation Benzene Exposed ID Primary site Cutoff Date Exposure (Y/N) 13799 PNS 1/4/1955 1959 N 14386 PNS 12/14/1959 1962 N 30129 PNS 4/24/1959 1960 Y 35000 PNS 4/1/1958 1959 N 41534 PNS 8/31/1962 1966 Y 91959 LANL 2/24/1973 1974 Y 161708 INL, HAN 12/28/1986 1987 N 178605 INL 4/23/1965 1966 Y 179682 INL 10/3/1958 1959 N

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6.0 Statistical Analysis

6.1 Methods

All statistical analyses were conducted using SAS  software [2007]. Modeling was conducted using the methods outlined by Langholz and Richardson [2010]. Simple modeling and univariate analyses were conducted by 1:4 conditional logistic regression using the PHREG procedure, whereby the rate ratio (RR) was determined by log-linear model of the form, log(RR(z;β))=zβ, where z is a vector of

explanatory variables. Previous research suggests that the exponential dose-response trend described

by log-linear models may poorly represent the relation between radiation and leukemia; therefore,

general relative risk models that represent a linear trend in radiation dose have become popular in

radiation epidemiology. Linear dose-response modeling was conducted by conditional logistic regression

using PROC NLMIXED and PROC NLP procedures, where odds ratios were used to approximate rate

ratios. Relative risk models followed the general form

ͧ ġ ∑Ĝu QĜĜ ͌͌ Ɣ ʬ1 ƍ ȕ %̾%ʭ ͙ %Ͱͥ where α is the radiation risk coefficient; βi is the coefficient for the ith covariate of 1, 2 … n covariates

indexed by multiplicative factors, Xi; and γj is the radiation weighting factor for the jth dose variable of

three linear dose covariates, D. For all models, three radiation sources were considered: 1) low-LET (D1)

irradiation comprised of the sum of gamma, x-ray, and tritium doses; 2) plutonium deposition (D 2); and

3) neutron radiation ( D 3). In this model, the dose-response is linear in radiation dose that is normalized to the effects of low-LET radiation exposures and adjusted for other potential covariates (e.g., benzene, birth cohort, and SES), which are modeled as log-linear functions of the factors of interest. Setting the

Low-LET radiation weighting factor to unity (γ1=1) allows for solutions to radiation weights for plutonium

and neutron doses that can be compared to standard values. Conversely, holding all radiation weights

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constant allows for a weighted sum of dose, which is preferred in radiation protection settings. For the latter approach, weights of 1, 10, and 20 were assumed for low-LET, neutron, and plutonium, respectively, which is consistent with values recommended by the ICRP [ICRP 2003]. Final models were replicated using the computer program PECAN in Epicure [1998]. Profile likelihood-based (PL-based) two-sided 95% confidence intervals were estimated for log-linear and linear model parameter estimates.

Effect modification was evaluated by likelihood ratio test at the 0.05 critical level of the cross- product term of the potential effect modifier and radiation dose. Factors evaluated as potential effect modifiers were: sex (male/female), race (White non-Hispanic/all others), birth cohort (tertiles), period of first hire (tertiles), facility of longest employment duration, and study group (INL group vs. LCCS).

Potential confounders were incorporated into models when causing >15% relative change in the overall dose parameter estimate. In addition to the factors that were tested as effect modifiers, exposures to benzene, neutron, plutonium, as well as SES were examined for potential confounding of low-LET risk estimates. To minimize the potential bias from selection of cutpoints for categories of birth cohort and hire date, continuous variables were constructed from terms calculated by restricted cubic spline (RCS) models, with three knots at the 10 th , 50 th , and 90 th percentile of birth date and hire date, respectively

[Harrell 2001]. Because the date of birth and the date of hire were correlated (Pearson coefficient=0.59,

P<0.001), models were restricted to include no more than one of these variables at a time. Final selection between hire date and birth date was based on a comparison of observed effect and the statistical significance of the covariates.

Baseline models were constructed for leukemia, leukemia excluding the CLL subtype, and three of the four major subtypes: AML, CML, and CLL. There were too few cases to adequately examine ALL risks separately. Modeling assumptions were further evaluated by sensitivity analyses that were conducted on baseline models. Factors considered in sensitivity analyses were:

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• Radiation weights: The selection of fixed radiation weights for the leukemogenicity of the high-LET

dose components is arbitrary. Information on the appropriate weights is equivocal. Kocher et al.

[2005] suggested a weighting scheme for fission neutrons and alpha particles with median values of

11 and 3.6, respectively. The U.S. EPA suggested that the effective weight for alpha emitters

deposited on bone surfaces or in bone is best described by uniform probability distribution between

0 and 1 [Puskin et al. 1999].

• Time since exposure: Some studies have suggested a varying time course of leukemia risk following

protracted radiation exposure [Schubauer-Berigan et al. 2007a; Shilnikova et al. 2003]. To evaluate

the effectiveness of radiation in producing leukemia risk over time, models were constructed

following division of cumulative exposure into time windows, in years, before the cutoff date.

Leukemia risks were assessed in windows of 0-<5, 5-<10, 10-<15, and 15+ years. Given its suspected

long latency, CLL risk was further assessed in time periods of 15-<20, 20-<25, 25-<30 and 30+ years.

• Alternate lag assumptions: All models were examined under alternate exposure lag assumptions of

0-, 2-, 5-, 7-, and 10 years. CLL models also included exposure lags of 15-, 20, 25, and 30 years. The

preferred exposure lag period was determined based on the lowest model deviance after

adjustment for relevant cofactors.

• Indeterminate leukemia cases: There was insufficient information on 31 leukemia cases to

definitively determine CLL/non-CLL status. By design, these cases were excluded from the main

analyses. To evaluate the effects from a potential source of misclassification, models of non-CLL

leukemia and CLL were reexamined after including indeterminate cases.

• High-dose worker effects: There is some evidence that radiation risk estimates may be heavily

influenced by highly skewed exposure data [Rosario et al. 2006; Schubauer-Berigan et al. 2007a]. To

examine these effects, the modeling results for leukemia excluding CLL under the preferred

exposure lag were determined using modified risk sets restricted to subjects who have lagged

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cumulative low-LET doses <100 mGy. The removal of cases exceeding 100 mGy also meant the

removal of matched controls, regardless of cumulative dose. Similarly, the model of the best fit

exposure window was restricted to persons in the lower 99% of the dose distribution (i.e., 1%

trimming).

• Evaluation of model non-linearity : By design, the risk per unit radiation dose was assumed to be

additive. The linearity of the dose-response was evaluated by: 1) adding a quadratic term to the

preferred model, 2) examining the risk by dose categories, and 3) fitting a restricted cubic spline

model. For cubic spline models, the 5th ,35th , 65 th , and 95th percentile of the distribution of low-LET

dose across the risk sets were used to identify the position of four knots. Model fits were judged by

Akaike Information Criterion (AIC) [Akaike 1973]. This evaluation was restricted to leukemia

excluding CLL under the preferred lag and adjusted for confounders.

6.2 Results

6.3 Univariate analysis

The results from univariate analyses of explanatory factors in categories of demography

(gender, race/ethnicity, birth cohort), employment (year first hired and SES), and exposure (low-LET radiation, plutonium bioassay, and benzene) are shown in Table 6-1 through Table 6-4. These analyses were restricted to outcomes of leukemia, (Table 6-1), leukemia, excluding CLL (Table 6-2), CLL (Table 6-

3), CML (Table 6-4), and AML (Table 6-5). Exposure categories were constructed under a two-year lag.

There were too few ALL cases (n=18) to be informative for most analyses, although those ALL analyses with sufficient cases in all categories provided results that were similar to those shown for the other subtypes (results not shown).

Similar patterns were evident across analyses, although results were highly uncertain. Deficits were observed in women and non-Hispanic Whites compared to men and other races/ethnicities,

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respectively; however, among these factors, only CML in other races/ethnicities compared to non-

Hispanic Whites was statistically significant (HR=4.1; 95% CI 1.3, 12.7). The most pronounced effect was observed across categories of birth year and year of first hire, where less risk was observed in those persons born earlier and/or hired earlier.

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Table 6-1. Univariate analysis of covariates in analysis of leukemia cases and age-matched controls

Variable 1 Description Cases (n=369) Controls (n=1,476) Total (n=1,845) Hazard Ratio (95% CI) Sex (%) Male 333 (90.2) 1,283 (86.9) 1,616 (87.6) 1.0 Female 36 (9.8) 193 (13.1) 229 (12.4) 0.72 (0.49, 1.04) Race (%) White non -Hispanic 344 (93.2) 1,426 (96.6) 1,770 (95.9) 1.0 Other 25 (6.8) 50 (3.4) 75 (4.1) 2.13 (1.27, 3.52) Birth year (%) Birth year <1920 110 (29.8) 565 (38.3) 675 (36.6) 1.0 Birth year 1920 -1927 124 (33.6) 437 (29.6) 561 (30.4) 1.53 (1.14, 2.07) Birth year >1927 135 (36.6) 474 (32.1) 609 (33.0) 1.68 (1.20, 2.37) Hire year (%) Hire year <1953 117 (31.7) 549 (37.2) 666 (36.1 1.0 Hire year 1953 -1959 131 (35.5) 441 (29.9) 572 (31.0) 1.43 (1.07, 1.91) Hire year > 1959 121 (32.8) 486 (32.9) 607 (32.9) 1.20 (0.88, 1.62) SES (%) Professional 81 (22.0) 304 (20.6) 385 (20.9) 1.0 Intermediate 50 (13.6) 237 (16.1) 287 (15.6) 0.79 (0.53, 1.17) Skilled non -manual 36 (9.8) 176 (11.9) 212 (11.5) 0.76 (0.49, 1.17) Skilled manual 140 (37.9) 538 (36.5) 678 (36.8) 0.98 (0.72, 1.34) Partly skilled 21 (5.7) 66 (4.5) 87 (4.7) 1.20 (0.68, 2.05) Unskilled 41 (11.1) 155 (10.5) 196 (10.6) 1.00 (0.65, 1.50) Low -LET dose 0-<1.0 mGy 95 (25.8) 353 (23.9) 448 (24.3) 1.0 distribution 1-<10 mGy 123 (33.3) 503 (34.1) 626 (33.9) 0.90 (0.65, 1.23) (%) 10 -<50 mGy 101 (27.4) 434 (29.4) 535 (29.0) 0.85 (0.61, 1.18) 50 -<100 mGy 24 (6.5) 77 (5.2) 101 (5.5) 1.14 (0.67, 1.88) ≥100 mGy 26 (7.1) 109 (7.4) 135 (7.3) 0.87 (0.52, 1.42) Max plutonium unmonitored 254 (68.8) 1,037 (70.3) 1,291 (70.0) 1.0 urine bioassay 0-<1.7 mBq ∙d -1 86 (23.3) 340 (23.0) 426 (23.1) 1.04 0.78, 1.38) results (%) 1.7 -<17 mBq ∙d -1 23 (6.2) 81 (5.5) 104 (5.6) 1.17 (0.70, 1.89) ≥17 mBq∙d -1 6 (1.6) 18 (1.2) 24 (1.3) 1.36 (0.49, 3.26) Benzene (%) Unexposed (score=0) 300 (81.3) 1,185 (80.3) 1,485 (80.5) 1.0 Low 0 - median exposed 32 (8.7) 148 (10.0) 180 (9.8) 0.85 (0.55, 1.26)

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Table 6-2. Univariate analysis of covariates in analysis of non-CLL cases and age-matched controls

Variable 1 Description Cases (n=264) Controls (n=1,056) Total (n=1,320) Hazard Ratio (95% CI) Sex (%) Male 236 (89.4) 910 (86.2) 1,146 (86.8) 1.0 Female 28 (10.6) 146 (13.83) 174 (13.2) 0.74 (0.48, 1.12) Race (%) White non -Hispanic 247 (93.6) 1,017 (96.3) 1264 (95.8) 1.0 Other 17 (6.4) 39 (3.7) 56 (4.24) 1.83 (0.98, 3.27) Birth year (%) Birth year <1920 71 (26.9) 377 (35.7) 448 (33.9) 1.0 Birth year 1920 -1927 87 (33.0) 312 (29.6) 399 (30.3) 1.54 (1.08, 2.22) Birth year >1927 106 (40.1) 367 (34.8) 473 (35.8) 1.79 (1.20, 2.68) Hire year (%) Hire year <1953 80 (30.3) 376 (35.6) 456 (34.6) 1.0 Hire year 1953 -1959 95 (36.0) 315 (29.8) 410 (31.1) 1.46 (1.03, 2.07) Hire year > 1959 89 (33.7) 365 (34.6) 454 (34.4) 1.17 (0.82, 1.66) SES (%) Professional 61 (23.1) 218 (20.6) 279 (21.1) 1.0 Intermediate 39 (14.8) 188 (17.8) 227 (17.2) 0.73 (0.46, 1.14) Skilled non -manual 30 (11.4) 131 (12.4) 161 (12.2) 0.81 (0.49, 1.32) Skilled manual 99 (37.5) 364 (34.4) 463 (35.1) 0.97 (0.68, 1.41) Partly skilled 13 (4.9) 44 (4.2) 57 (4.3) 1.07 (0.52, 2.06) Unskilled 22 (8.3) 11 (10.6) 133 (10.1) 0.70 (0.40, 1.18) Low -LET dose 0-<1.0 mGy 74 (28.1) 264 (25.0) 338 (25.6) 1.0 distribution 1-<10 mGy 80 (30.3) 367 (34.7) 447 (33.9) 0.76 (0.53, 1.10) (%) 10 -<50 mGy 69 (25.1) 299 (28.3) 368 (27.9) 0.81 (0.55, 1.19) 50 -<100 mGy 21 (7.9) 53 (5.0) 74 (5.6) 1.38 (0.77, 2.41) ≥100 mGy 20 (7.6) 73 (6.9) 93 (7.1) 0.96 (0.53, 1.69) Max plutonium unmonitored 186 (70.5) 747 (70.7) 933 (70.7) 1.0 urine bioassay 0-<1.7 mBq ∙d -1 58 (22.0) 236 (22.4) 294 (22.3) 0.99 (0.69, 1.38) results (%) 1.7 -<17 mBq ∙d -1 15 (5.7) 59 (5.6) 74 (5.6) 1.03 ((0.54, 1.82 ≥17 mBq∙d -1 5 (1.9) 14 (1.3) 19 (1.4) 1.43 (0.46, 3.74) Benzene (%) Unexposed (score=0) 217 (82.2) 841 (79.6) 1,058 (80.2) 1.0 Low 0 - median exposed 19 (7.2) 104 (9.9) 123 (9.3) 0.70 (0.4, 1.15) 1Exposure categories (low-LET, plutonium, and benzene) are under a 2-year lag.

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Table 6-3. Univariate analysis of covariates in analysis of CLL cases and age-matched controls (excl. indeterminate cases)

Variable 1 Description Cases (n=74) Controls (n= 296 ) Total (n= 370 ) Hazard Ratio (95% CI) Sex (%) Male 68 (91.9) 263 (88.9) 331 (89.5) 1.0 Female 6 (8.1) 33 (11.2) 39 (10.5) 0.70 (0.25, 1.64) Race (%) White non -Hispanic 68 (91.9) 286 (96.6) 354 (95.7) 1.0 Other 6 (8.1) 10 (3.4) 16 (4.3) 2.74 (0.86, 8.41) Birth year (%) Birth year <1920 24 (32.4) 133 (44.9) 157 (42.4) 1.0 Birth year 1920 -1927 26 (35.1) 88 (29.7) 114 (30.8) 1.77 (0.93, 3.48) Birth year >1927 24 (32.4) 75 (25.4) 99 (26.8) 2.09 (1.01, 4.47) Hire year (%) Hire year <1953 25 (33.8) 119 (40.2) 144 (38.9) 1.0 Hire year 1953 -1959 26 (35.1) 97 (32.8) 123 (33.2) 1.28 (0.69, 2.37) Hire year > 1959 23 (31.1) 80 (27.1) 103 (27.8) 1.40 (0.72, 2.70) SES (%) Professional 17 (23.0) 65 (22.0) 82 (22.2) 1.0 Intermediate 10 (13.5) 34 (11.5) 44 (11.9) 1.09 (0.44, 2.63) Skilled non -manual 4 (5.4) 31 (10.5) 35 (9.5) 0.46 (0.12, 1.38) Skilled manual 24 (32.4) 120 (40.5) 144 (38.9) 0.73 (0.36, 1.47) Partly skilled 3 (4.1) 17 (5.7) 20 (5.4) 0.64 ( 0.14, 2.16) Unskilled 16 (21.6) 29 (9.8) 45 (12.2) 1.95 (0.90, 4.21) Low -LET dose 0-<1.0 mGy 14 (18.9) 58 (19.6) 72 (19.5) 1.0 distribution 1-<10 mGy 29 (39.2) 94 (31.8) 123 (33.2) 1.28 (0.59, 2.86) (%) 10 -<50 mGy 24 (32.4) 102 (34.5) 126 (34.0) 0.98 (0.45, 2.20) 50 -<100 mGy 3 (4.1) 17 (5.7) 20 (5.4) 0.75 (0.15, 2.79) ≥100 mGy 4 (5.4) 25 (8.5) 29 (7.8) 0.67 (0.17, 1.16) Max plutonium unmonitored 45 (60.8) 194 (65.5) 239 (64.6) 1.0 urine bioassay 0-<1.7 mBq ∙d -1 21 (28.4) 79 (26.7) 100 (27.0) 1.18 (0.64, 2.14) results (%) 1.7 -<17 mBq ∙d -1 7 (9.5) 19 (6.4) 26 (7.0) 1.65 (0.60, 4.21) ≥17 mBq∙d -1 1 (1.4) 4 (1.4) 5 (1.4) 1.12 (0.06, 7.69) Benzene (%) Unexposed (score=0) 58 (78.4) 239 (80.4) 297 (80.3) 1.0 Low 0 - median exposed 9 (12.2) 33 (11.2) 42 (11.4) 1.13 (0.47, 2.52) 1Exposure categories (low-LET, plutonium, and benzene) are under a 2-year lag.

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Table 6-4. Univariate analysis of covariates in analysis of CML cases and age-matched controls

Variable 1 Description Cases (n= 52 ) Controls (n=2 08 ) Total (n= 260 ) Hazard Ratio (95% CI) Sex (%) Male 47 (90.4) 180 (86.5) 227 (87.3) 1.0 Female 5 (9.6) 28 (13.5) 33 (12.7) 0.69 (0.22, 1.72) Race (%) White non -Hispanic 45 (86.5) 200 (96.2) 245 (94.2) 1.0 Other 7 (13.5) 8 (3.9) 15 (5.8) 4.1 (1.3, 12.7) Birth year (%) Birth year <1920 15 (28.9) 67 (32.2) 82 (31.5) 1.0 Birth year 1920 -1927 20 (38.5) 61 (29.3) 81 (31.2) 1.51 (0.70, 3.39) Birth year >1927 17 (32.7) 80 (38.5) 97(37.3) 0.86, 0.36, 2.09) Hire year (%) Hire year <1953 10 (19.2) 79 (38.0) 89 (34.2) 1.0 Hire year 1953 -1959 25 (48.1) 51 (24.5) 76 (29.2) 3.74 (1.71, 8.79) Hire year > 1959 17 (32.7) 78 (37.60 95 (36.5) 1.71 (0.72, 4.20) SES (%) Professional 11 (21.2) 38 (18.3) 49 (18.50 1.0 Intermediate 10 (19.2) 38 (18.3) 48 (18.5) 0.88 (0.33, 2.31) Skilled non -manual 6 (11.5) 27 (13.0) 33 (12.7) 0.77 (0.23, 2.34) Skilled manual 20 (38.5) 7 4 (35.6) 94 (36.2) 0.95 (0.41, 2.25) Partly skilled 2 (3.9) 9 (4.3) 11 (4.2) 0.77 (0.10, 3.73) Unskilled 3 (5.8) 22 (10.6) 25 (9.6) 0.46 (0.09, 1.69) Low -LET dose 0-<1.0 mGy 15 (28.9) 42 (20.2) 57 (21.9) 1.0 distribution 1-<10 mGy 12 (23.1) 71 (34.1) 83 (31.9) 0.46 (0.18, 1.12) (%) 10 -<50 mGy 14 (26.9) 70 (33.7) 84 (32.3) 0.50 (0.20, 1.20) 50 -<100 mGy 5 (9.6) 5 (2.4) 10 (3.9) 2.34 (0.60, 9.53) ≥100 mGy 6 (11.5) 20 (9.62) 26 (10.0) 0.71 (0.21, 2.22) Max plutonium unmonitored 33 (63.5) 128 (61.5) 161 (61.9) 1.0 urine bioassay 0-<1.7 mBq ∙d -1 16 (30.8) 61 (29.3) 77(29.6) 1.0 (0.51, 1.93) results (%) 1.7 -<17 mBq ∙d -1 2 (3.8) 17 (8.2) 19 (7.3) 0.47 (0.07, 1.71) ≥17 mBq∙d -1 1 (1.9) 2 (1.0) 3 (1.2) 1.91 (0.09, 20.24) Benzene (%) Unexposed (score=0) 40 (76.9) 157 (75.5) 197 (75.8) 1.0 Low 0 - median exposed 4 (7.8) 25 (12.2) 29 (11.20 0.64 (0.18, 1.76) 1Exposure categories (low-LET, plutonium, and benzene) are under a 2-year lag.

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Table 6-5. Univariate analysis of covariates in analysis of AML cases and age-matched controls

Variable 1 Description Cases (n= 147 ) Controls (n= 588 ) Total (n= 735 ) Hazard Ratio (95% CI) Sex (%) Male 129 (87.8) 510 (86.7) 639 (86.9) 1.0 Female 18 (12.3) 78 (13.3) 96 (13.1) 0.91 (0.52, 1.54) Race (%) White non -Hispanic 142 (96.6) 570 (96.9) 712 (96.9) 1.0 Other 5 (3.4) 18 (3.1) 23 (3.1) 1.12 (0.36, 2.94) Birth year (%) Birth year <1920 38 (25.9) 210 (35.7) 248 (33.7) 1.0 Birth year 1920 -1927 53 (36.1) 184 (31.3) 237 (32.2) 1.67 (1.04, 2.72) Birth year >1927 56 (38.1) 194 (33.0) 250 (34.1) 1.87 (1.09, 3.28 Hire year (%) Hire year <1953 49 (33.3) 209 (35.5) 258 (35.1) 1.0 Hire year 1953 -1959 51 (34.7) 182 (31.0) 233 (31.7) 1.22 (0.77, 1.95) Hire year > 1959 47 (32.0) 197 (33.5) 244 (33.2) 1.02 (0.63, 1.65) SES (%) Professional 34 (23.1) 131 (22.3) 165 (22.5) 1.0 Intermediate 19 (12.9) 101 (17.2) 120 (16.3) 0.72 (0.38, 1.32) Skilled non -manual 18 (12.2) 72 (12.2) 90 (12.2) 0.96 (0.50, 1.77) Skilled manual 53 (36.1) 191 (32.5) 244 (33.2) 1.07 (0.67, 1.75) Partly skilled 9 (6.1) 29 (4.9) 38 (5.2) 1.21 (0.50, 2.72) Unskilled 14 (9.5) 64 (10.9) 78 (10.6) 0.84 (0.41, 1.63) Low -LET dose 0-<1.0 mGy 43 (29.3) 158 (26.9) 201 (27.4) 1.0 distribution 1-<10 mGy 47 (32.0) 187 (31.8) 234 (31.8) 0.91 (0.56, 1.48) (%) 10 -<50 mGy 34 (23.1) 170 (28.9) 204 (27.8) 0.72 (0.43, 1.21) 50 -<100 mGy 12 (8.2) 38 (5.6) 50 (6.8) 1.14 (0.53, 2.35) ≥100 mGy 11 (7.5) 35 (6.0) 46 (6.3) 1.17 (0. 52, 2.49) Max plutonium unmonitored 105 (71.4) 438 (74.5) 543 (73.9) 1.0 urine bioassay 0-<1.7 mBq ∙d -1 28 (19.1) 117 (19.9) 145 (19.7) 1.01 (0.61, 1.65) results (%) 1.7 -<17 mBq ∙d -1 11 (7.5) 28 (4.8) 39 (5.3) 1.70 (0.77, 3.51) ≥17 mBq∙d -1 3 (2.0) 5 (0.9) 8 (1.1) 2.60 (0.53, 10.73) Benzene (%) Unexposed (score=0) 125 (85.1) 464 (78.9) 589 (80.1) 1.0 Low 0 - median exposed 10 (6.8) 65 (11.1) 75 (10.2) 0.56 (0.26, 1.09)

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6.3.1 Multivariable analyses

The results of the main analysis of low-LET bone marrow irradiation and the risk of non-CLL leukemia are shown in Table 6-6 through Table 6-10 below. These analyses were essentially repeated for each leukemia subtype, excluding ALL, and the results of those models are summarized at the end of the

results section.

The ERR per 10 mGy -1 of bone marrow dose, lagged two-years, was less than one percent in the

base model of non-CLL leukemia and low-LET radiation (Table 6-6). All model estimates were not

statistically significant at the 0.05 level. Adding plutonium dose to the model had negligible effects on

the low-LET risk estimate when solving for radiation weights; however, including neutron dose lowered

the risk estimate by over 30%. The calculated radiation weights for the high-LET radiation components

were extremely large relative to expected radiation weighting factors. This difference may be due to

problems in model specification, bias in dose estimates, or differences in other risk factors in those

persons with neutron and/or plutonium exposure. Holding the weights constant at ICRP recommended

values increased the risk estimate by 63% and nearly all (90%) of the change was attributable to

plutonium dose. Given these inconsistencies in radiation weights and that a relatively moderate effect

to risk estimates was evident when including high-LET dose contributions, subsequent models in main

analyses excluded plutonium and neutron doses.

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Table 6-6. Effects on low-LET radiation risk estimates from neutrons (n) and/or plutonium (p) exposures in leukemia excluding CLL under a two-year lag in a linear model (n=264).

Radiation weighting Adjusted Low –LET Radiation Percent c hange from factor type for: ERR 10 mGy -1 weighting unadjusted (P-value) factor model estimates Unadjusted 0.0089 (0.57) NA NA +n 0.0059 (0.70) 448 -34.2 +p 0.0091 (0.56) 122 1.7 +n and p 0.0061 (0.69) 416 (n) , 187 (p) -31.2 constant s +n 0.0098 (0.53) 10 (fixed) 9.3 +p 0.0139 (0.34) 20 (fixed) 56.3 +n and p 0.0145 (0.33) 10 (n ), 20 (p) 62.9 NA= not applicable; n= neutron; p= plutonium

Of variables tested for interaction with low-LET radiation in the pooled analyses, only race approached statistical significance (Table 6-7). The effect of facility on the pooled analyses could not be tested by design; however, restricting analyses to the LCCS group and examining facility effects showed risk heterogeneity across facilities (LRT=18.07, df=4, P=0.003). Positive dose-responses were observed in workers from PNS (ERR ∙10 mGy -1= 0.37; 95% CI: >0.076, 2.0) and SRS (ERR ∙10 mGy -1= 0.25; 95% CI:

0.044, 0.83), which influenced the overall estimate for the LCCS Group (ERR ∙10 mGy -1= 0.012; 95% CI: -

0.016, 0.066). The estimate of non-CLL leukemia risk for INL was also positive (ERR ∙10 mGy -1= 0.003;

95% CI: -0.023, 0.081), although markedly less than the LCCS Group. Nevertheless, there was no evidence of significant risk differences between study groups (LRT=0.086, df=1, P=0.769).

253

Table 6-7. Tests for interaction between radiation dose and sex, race, birth cohort, hire year, SES, and study group for Leukemia excluding CLL

Model ERR 10 mGy -1 LRT Value df P-Value Sex Female -0.0677 (fixed) 1. 551 1 0.2 13 Male 0.008 4 Race Non Hispanic White s 0. 3032 3. 069 1 0.080 Non White 0.0 055 Birth cohort YOB ≤ 1919 -0. 000 5 0.6 21 2 0.7 33 1919 < YOB ≤1927 0.02 81 1927 < YOB 0. 0036 Hire year YFE ≤ 1952 0.01 28 0.368 2 0.8 32 1952

Including sex alone in the model reduced risk estimates for non-CLL leukemia by over 27%

(Table 6-8). In contrast, adjusting for race alone increased the estimate by nearly 40%. Including hire

year by tertiles resulted in a 40% increase in the estimate compared to just under a 15% increase when

controlling for hire year using RCS. Adjustment for benzene, SES, or birth cohort resulted in less than

15% relative change in the unadjusted parameter estimate. Similar results were obtained for CML,

where final models were adjusted for race, sex, and hire year. The model of AML risk was not influenced

by sex or race but was increased by 28% when including RCS terms for birth cohort and 17% when

254

including hire date. Given the strong association between birth cohort and AML compared to hire date and AML, subsequent models were controlled for birth date using spline terms. CLL risk estimates were influenced by sex (four-fold decrease), race (two-fold increase), and birth cohort (ten-fold increase using

RCS).

Given the strong correlation between birth cohort and hire year, it is likely that both covariates were measures of the same confounding effect on the relation between radiation exposure and leukemia risk. However, unlike birth cohort, hire year is not likely to be an independent risk factor for disease (a necessary condition for a confounder); therefore, adjustment by hire year may be inappropriate despite the observed effect. Subsequent sensitivity analyses comparing model fit and confounding of radiation risk estimates by time at birth and time at hire did not reveal significant differences in fit or effect. Therefore, adjusting for either birth year or hire year provided comparable results.

Table 6-8.Tests for confounding of radiation dose effect by sex, race, birth cohort, hire year, benzene exposure, and SES in leukemia excluding CLL.

Percent change P-value of confounder ERR 10 mGy -1 from unadjusted confounder No Adjustment 0.0089 NA NA Sex 0.00 65 -27.1 0.180 Race 0. 011 37.8 0.050 Birth Cohort Spline 0.00 88 -1.3 0.023 Categorical 0.0099 11.1 0.008 Hire Year Spline 0.0 102 14.7 0.083 Categorical 0.0126 40.8 0.086 Benzene 0.00 87 -2.0 0.105 SES 0.00 79 -11.3 0.564

The results of examining the effects of different exposure lag periods and windows of exposure before the cutoff date are shown in Table 6-9 through Table 6-16. A two-year exposure lag provided the best linear model fit for non-CLL leukemia, although parameter estimates did not vary markedly over

255

the lag periods tested and were imprecise (Table 6-9). Non-CLL leukemia risk was highest in the exposure window of 5 to 10 years prior to death (ERR ∙10 mGy -1= 0.20; 95% CI:-0.032, 0.68) and decreased with increasing time prior to attained age (Table 6-10). A two-year exposure lag provided the lowest model deviance for AML (Table 6-11 and Table 6-12), compared to a five-year lag for CML (Table

6-13 and Table 6-14). Similar wave-like risk patterns were observed in all subtype analyses, although the best window fit for CML (10-15 years) was slightly earlier than AML (5-10 years). Of these models, only the risk of AML in the window 5 to 10 years preceding death was statistically significant (ERR∙10 mGy -1 =

0.76; 95% CI: (0.047, 2.7).There was little evidence of improved model fit for CLL under any exposure lag

assumption used; however, the best fit by exposure window was achieved at 25 to 30 years before

death (Table 6-15 and Table 6-16).

256

Table 6-9. The effect of exposure lag on the excess relative risk of non-CLL leukemia from low-LET radiation dose (n=264 cases)

No Adjustment Adjusted for race , sex , and hire year Lag (years) ERR 10 mGy -1 P-value -2LL ERR 10 mGy -1 P-value -2LL 0 0.008 0.580 849.48 0.008 0.568 840.95 2 0.009 0.531 849.39 0.009 0.522 840.87 5 0.009 0.530 849.39 0.009 0.529 840.88 7 0.008 0.572 849.46 0.008 0.575 840.96 10 0.006 0.682 849.62 0.006 0.697 841.13 Shaded row indicates selected model

Table 6-10. Analysis of radiation and non-CLL leukemia risk by exposure windows of years before the cutoff date (n=264 cases)

Exposure No Adjustment Adjusted for race, sex, and hire year window (years) ERR 10 mGy -1 P-value -2LL ERR 10 mGy -1 P-value -2LL 0-<5 -0.033 0.712 849.65 -0.026 0.786 841.20 5-<10 0.163 0.160 847.81 0.197 0.120 838.86 10 -<15 0.119 0.301 848.71 0.130 0.280 840.11 15+ 0.004 0.787 849.71 0.004 0.280 840.11 Shaded row indicates best fit

257

Table 6-11. The effect of exposure lag on the excess relative risk of AML from radiation dose ( n=147 cases)

No Adjustment Adjusted for birth date (RCS) Lag (years) ERR 10 mGy -1 P-value -2LL ERR 10 mGy -1 P-value -2LL 0 0.014 0.480 472.68 0.019 0.379 465.43 2 0.015 0.455 472.62 0.019 0.370 465.40 5 0.016 0.440 472.58 0.019 0.382 465.44 7 0.015 0.471 472.65 0.017 0.432 465.59 10 0.013 0.544 472.81 0.013 0.542 465.84 Shaded row indicates selected model

Table 6-12. Analysis of radiation and AML risk by exposure windows of years before the cutoff date (n=147 cases)

Exposure No Adjustment Adjusted for birth date (RCS) window (years) ERR 10 mGy -1 P-value -2LL ERR 10 mGy -1 P-value -2LL 0-<5 -0.043 0.729 473.06 0.044 0.804 466.15 5-<10 0.250 0.211 471.61 0.761 0.026 461.24 10 -<15 0.183 0.327 472.21 0.407 0.099 463.49 15+ 0.011 0.617 472.92 0.009 0.674 466.03 Shaded row indicates best fit

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Table 6-13. The effect of exposure lag on the excess relative risk of CML from radiation dose ( n=52 cases)

No Adjustment Adjusted for race, sex, and hire year Lag (years) ERR 10 mGy -1 P-value -2LL ERR 10 mGy -1 P-value -2LL 0 0.021 0.470 166.86 0.026 0.429 158.15 2 0.024 0.433 166.77 0.028 0.404 158.07 5 0.025 0.417 166.72 0.029 0.398 158.05 7 0.026 0.412 166.71 0.028 0.403 158.07 10 0.023 0.446 166.80 0.024 0.458 158.22 Shaded row indicates selected model

Table 6-14. Analysis of radiation and CML risk by exposure windows of years before the cutoff date (n=52 cases)

Exposure No Adjustment Adjusted for race, sex, and hire year window (years) ERR 10 mGy -1 P-value -2LL ERR 10 mGy -1 P-value -2LL 0-<5 -0.963 0.446 166.80 -0.975 0.458 158.22 5-<10 0.070 0.737 167.27 0.182 0.494 158.30 10 -<15 0.365 0.244 166.03 0.438 0.248 157.44 15+ 0.020 0.518 166.96 0.020 0.523 158.36 Shaded row indicates best fit

259

Table 6-15. The effect of exposure lag on the excess relative risk of CLL from radiation dose ( n=74 cases)

No Adjustment Adjusted for race, sex, and birth cohort Lag (years) ERR 10 mGy -1 P-value -2LL ERR 10 mGy -1 P-value -2LL 0 0.000 0.993 238.20 -0.0038 0.770 227.47 2 0.000 0.987 238.20 -0.0038 0.770 227.47 5 0.001 0.966 238.20 -0.0038 0.772 227.48 7 0.001 0.936 238.19 -0.0037 0.783 227.48 10 0.002 0.907 238.18 -0.0036 0.789 227.49 15 0.00 4 0.811 238.14 -0.0026 0.850 225.16 20 0.005 0.777 238.12 -0.0024 0.868 225.17 25 0.004 0.834 238.15 -0.0029 0.841 225.16 30 0.008 0.765 238.11 -0.0020 0.915 225.19 Shaded row indicates selected model

Table 6-16. Analysis of radiation and CLL risk by exposure windows of years before the cutoff date (n=74 cases)

Exposure No Adjustment Adjusted for race, sex, and birth cohort window (years) ERR 10 mGy -1 P-value -2LL ERR 10 mGy -1 P-value -2LL 0-<5 -0.393 0.242 236.83 -0.110 0.850 227.52 5-<10 10 -<15 -0.277 0.078 235.10 -0.261 0.133 225.30 15 -<20 -0.074 0.681 238.03 -0.071 0.710 225.06 20 -<25 0.049 0.654 238 .00 0.019 0.851 225.16 25 -<30 -0.008 0.914 238.19 -0.039 0.545 224.83 30+ 0.008 0.765 238.11 -0.002 0.915 225.19 Shaded row indicates best fit

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A summary of the preferred linear ERR models is shown in Table 6-17. In general, positive, but imprecise estimates of excess risk were observed in all adjusted models except for CLL (Table 6-17). The non-CLL ERR per 10 mGy was 0.009 (-0.014, 0.051) using a two-year lag, excluding indeterminate cases, and adjusting for sex, race, and hire year. Including indeterminate cases in the model reduced the observed non-CLL risk (ERR∙10 mGy -1 = 0.003; 95% CI: -0.15, 0.038). Among leukemia subtypes, AML risk

estimates appeared most precise (ERR∙10 mGy -1 = 0.019; 95% CI: -0.015, 0.089; two-year lag), although the highest risk was observed in CML (ERR∙10 mGy -1 = 0.029; 95% CI: <0, 0.18; five-year lag). There was

little evidence of radiation-induced CLL and risk estimates under all lag assumptions were negative. The

best linear fit for CLL used a 25 year lag and adjusted for sex, race, and birth year (ERR∙10 mGy -1 = -

0.003; 95% CI: -0.015, 0.051). Treating indeterminate cases as CLL had little effect on the CLL risk estimate (ERR∙10 mGy -1 = -0.002; 95%CI: <0, 0.047) after adjusting for confounders. All models exhibited attenuated risk in the high dose range as evidenced by increased risk after exposure trimming.

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Table 6-17. Summary of final linear modeling results for low-LET radiation and leukemia mortality

best fit exposure lag best fit exposure window ERR 10 mGy -1 (95% CI) ERR 10 mGy -1 (95% CI)

No. Lag Outcome cases 1 Adjusted for: (years) Full model Excl. >100 mGy Window Full model 1% Trim All 369 sex, race, hire 5 0.002 (-0.013, 0.029) 0.042 (-0.023, 0.14) 5-10 0.11 (-0.047, 0.46) 0.29 (<-0.034, 0.94) Leukemia year

Non-CLL 264 sex, race, hire 2 0.009 (-0.014, 0.051) 0.065 (-0.017, 0.19) 5-10 0.20 (-0.032, 0.68) 0.32 (-0.11, 1.1) Leukemia 2 year

AML 147 birth cohort 2 0.019 (-0.015, 0.089) 0.040 (-0.048, 0.19) 5-10 0.76 (0.047, 2.7) 1.1 (0.0042, 3.7)

CML 52 sex, race, hire 5 0.029 (<0, 0.18) 0.14 (-0.048, 0.62) 10-15 0.44 (<0, 2.4) 1.0 (<0, 4.03) year

CLL 1 74 Sex, race, 25 -0.003 (-0.015, 0.051) 0.043 (-0.013, 0.186) 25-30 -0.020 (NC, NC) -0.0125 (NC, NC) birth cohort

AML= Acute Myeloid Leukemia; ALL= Acute Lymphocytic Leukemia; CML= Chronic Myeloid Leukemia; CLL= Chronic Lymphocytic Leukemia; ERR=Excess Relative Risk; NC=not calculable. 1Number of cases in full models (i.e., there are fewer cases in restricted risk sets for trimmed models) 2Excluding leukemias of ambiguous subtype

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6.3.1.1 Radiation weights

The effects of including high-LET dose components in the adjusted risk model for non-CLL are shown in Table 6-18. Solving for radiation weights resulted in relatively large values for both high-LET components compared to ICRP recommended weights in the full model. Fixing weights at 10 and 20 for neutrons and plutonium alpha radiations, respectively, resulted in a 67% increase in estimated risk.

Using weights recommended by the U.S. EPA [1999] or by Kocher et al. [2005] also increased risk estimates, but by a lesser amount. These patterns were consistent in cumulative dose models; however weight solutions were reduced when using exposure windows. In models examining non-CLL risk in the

5-10 exposure window, the best fit plutonium and neutron weighting factors were zero and 16, respectively, and the relative change in the ERR estimate was approximately 3%.

Table 6-18. Sensitivity analysis of non-CLL leukemia risk including from high and low-LET radiations combined (lagged two-years), adjusted for sex, race, and hire year.

Linear ERR ∙10 mGy -1 Radiation Weights Relative Change (95% CI) plutonium Neutron (%) 0.009 ( -0.014, 0.051) 0 (fixed) 0 (fixed) NA 0.007 (<0, 0.049) 171 406 -22 0.015 ( -0.004, 0.056) 20 (fixed) 10 (fixed) 67 0.011 ( -0.012, 0.053) 1 (fixed) 10 (fixed) 22 0.012 ( -0.012, 0.055) 3.6 (fixed) 11 (fixed) 33 NA=not applicable

6.3.1.2 Dose-response linearity

There was strong evidence of a non-linear dose-response in all models. Figure 6-1 shows the results of fitting linear, linear-quadratic, categorical (0-<1.0 mGy, 1-<10 mGy, 10-<50 mGy, 50-<100 mGy, and ≥100 mGy), and RCS (four-knot) dose-response models of low-LET dose and leukemia, excluding the CLL subtype, under a two-year exposure lag. The best fit was obtained from the RCS model

(red line, AIC=844.25), which had highly significant spline terms ( P=0.002, P=0.001, and P=0.001 for terms 1, 2, and 3, respectively). The RCS shows risk attenuation at cumulative doses below 10 mGy,

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followed by a sharp increase to 100 mGy before decreasing again. Restricting the models to cumulative doses below 100 mGy improved the linear model fit in the high dose range (Figure 6-2) but the RCS still

provides the best model fit (Table-6-19) and the spline terms remained statistically significant ( P=0.005,

P=0.011, and P=0.013 for spline terms 1, 2, and 3, respectively). Restricting models to doses from 10

mGy to 100 mGy resulted in a best fit that is linear in dose (Figure 6-3). This restriction resulted in a

positive but imprecise linear ERR estimate (ERR∙10 mGy -1 = 0.35; 95% CI: -0.003, >1).

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Figure 6-2. Plot of dose-response models for low-LET dose and the relative risk of leukemia, excluding CLL and restricted to doses < 100 mGy. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots (n=244 cases)

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Figure 6-3. Plot of dose-response models for low-LET dose and the relative risk of leukemia, excluding CLL, and restricted to doses from 10 mGy to 100 mGy. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots (n=90 cases)

Table 6-19. Model fit characteristics for relation between non-CLL leukemia and low-LET radiation

Model Full Model Doses <100 mGy Dose 10 -<100 mGy (AIC) (AIC) (AIC) Linear 850.87 756.89 147.18 Linear -Quadratic 852.10 757.12 147.74 Categorical 852.18 758.08 147.57 RCS 844.74 751.47 149.80 Linear ERR ∙10 mGy -1 0. 009 (-0.014, 0.051) 0.065 (-0.018, 0.19 ) 0.35 (-0.0031 , >1) (95% CI) All models adjusted for sex, race, and hire year

Although not part of the original analysis plan, the non-CLL data were also fit to a three-piece linear spline model as a means to better characterize the dose-response [Steenland et al. 2011]. The model was first allowed to solve for the two knots used to indicate the points of departure from linearity in the low- and high-dose ranges. These knot value solutions (8 mGy, and 46 mGy) were fixed in

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subsequent models to reduce the total number of modeling parameters (Figure 6-4). The piecewise model provided the best overall fit to the data (AIC=835.9) and more precise estimates of the risk in each segment. Sex, race, and hire year were no longer confounding and were not included in the final model. The slope estimates (ERR∙10 mGy -1) were statistically significant in the first two segments: -0.68

(95% CI: -0.92, -0.33) for doses ≤8.0 mGy; 0.20 (95% CI: 0.082, 0.35), dose=8<-46 mGy; and -0.016 (95%

CI: <-0.022, 0.018), dose=46+ mGy.

Figure 6-4. Plot of dose-response models for low-LET dose and the relative risk of leukemia, excluding CLL. Red= three-piece linear spline model (knots= 8, 46); blue= restricted cubic spline (RCS), 4 knots (n=264)

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Dose response analyses were repeated by leukemia subtype and also for non-CLL in the 5-10 year exposure window. The subtype analyses revealed a similar dose-response response pattern for

AML (Figure 6-5) and CML (Figure 6-6), where attenuated risk was observed in the low and high dose ranges. CLL risk was best fit to a linear-quadratic model with a negative linear parameter estimate, although the model results were highly imprecise (Figure 6-7). In contrast, the non-CLL dose-response in models restricted to exposures in the 5-10 year window before the cutoff date (ERR∙10 -1 mGy = 0.20;

95% CI:-0.032, 0.68) was best fit to a linear model (Figure 6-8) with little evidence of risk attenuation in either extreme dose region.

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Figure 6-5. Plot of dose-response models for low-LET dose and the relative risk of AML. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=147 cases)

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Figure 6-6. Plot of dose-response models for low-LET dose and the relative risk of CML. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=52 cases)

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Figure 6-7. Plot of dose-response models for low-LET dose and the relative risk of CLL. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=74 cases).

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Figure 6-8. Plot of dose-response for low-LET dose in the 5-10 exposure window and the relative risk of non-CLL leukemia. Purple= linear; green=linear-quadratic; blue= categorical; red= restricted cubic spline (RCS), 4 knots ( n=264 cases).

6.4 Discussion

With 369 leukemia cases in a population of 105,245 U.S. nuclear workers, the current study is the largest ever conducted examining the leukemogenicity of protracted occupational exposures to ionizing radiation. The large study size improved examinations of dose-response compared to earlier occupational studies and enabled analyses to be conducted by leukemia subtype. Subsequently, nonlinear dose-responses were elucidated in all models and observed risks patterns were found to be heterogeneous across subtypes. For example, sub-linear non-CLL risks were observed at doses less than

10 mGy or greater than 100 mGy; however, a positive dose response was observed in a linear ERR model from 10 mGy to 100 mGy (ERR∙10 mGy -1= 0.35; 95% CI: -0.003, >1). Fitting the non-CLL data to a three-

piece linear spline model greatly improved model fit and resulted in a negative and significant dose-

response for dose below 8 mGy (ERR∙10 mGy -1= 0-0.68; 95% CI: -0.92, -0.33), then positive from 8 to 46

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mGy (ERR∙10 mGy -1= 0.20; 95% CI: 0.082, 0.35), and then negative for dose greater than 46 mGy (ERR∙10

mGy -1= -0.016; 95% CI: <-0.022, 0.018). This model also was absent of any confounding or effect

modification, which suggests instability of risk estimates in other models.

The cause of the non-linear dose response is unknown. Risk attenuation at higher exposures is

often observed in occupational studies, and may result from a number of causes including a depletion of susceptible persons, biologic saturation, exposure misclassification, or a healthy worker survivor bias

[Stayner et al. 2003]. A healthy worker effect is less likely because leukemia is not strongly related to lifestyle factors. Exposure misclassification, although inevitable, is believed to be reduced by the availability of individual measurement data and the exclusion of MED Era workers who are prone to highly uncertain dose estimates. With respect to low-dose observations, we note that cases with exposures of 1.0 mGy or less were typically professionals who were hired later, were older at hire, and worked fewer years prior to termination compared to cases in the higher dose categories. Thus, it is conceivable that some of these low-dose cases received exposure during employment elsewhere and the lack of this information in dose-response analyses has led to biased estimates. To examine the potential for this misclassification, non-CLL cases in the lowest dose category (n=74) were matched to

EEOICPA claimant information to uncover additional exposure information. Of non-CLL cases with assigned two-year lagged cumulative exposures below 1.0 mGy, eight workers were identified as

EEOICPA claimants. Of these, four cases had information on exposures at other facilities. Had these data been available to the current study, a change in exposure category would have resulted for all four cases.

The negative dose response in the low-dose range was not evident in previous studies of workers employed at these facilities [Schubauer-Berigan et al. 2007a; Schubauer-Berigan et al. 2005].

The addition of low-exposed cases through extended followup may be a source of the nonlinear dose- response. Restricting the current study to non-CLL cases with followup through 1990 ( n=112 cases)

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reduces the nonlinear low-dose effect and the best fit is obtained using a linear model (ERR∙10 mGy -1=

0.06; 95% CI: -0.006, 0.20). Further restricting the model to LCCS non-CLL cases (n=81 cases) completely

eliminates the negative dose-response effect in the low-dose region, although model results were

imprecise (ERR∙10 mGy -1= 0.09; 95% CI: -0.005, 0.27).

In the absence of known biases, one could interpret the results as evidence of a threshold or

protective effect. There has been considerable debate regarding the validity of the LNT Model because

the carcinogenicity of low and very low dose radiation exposure remains largely uncertain [Breckow

2006; Brenner and Sachs 2006; Chadwick and Leenhouts 2005; Charles 2006; Scott 2008; Tubiana et al.

2006; Wall et al. 2006]. This controversy has been fueled through advances in molecular biology, where

recent observations of non-linear cellular responses following irradiation suggest that supra-linear,

threshold, or even protective effects are biologically plausible [Averbeck 2009; Brenner 2009].

Nevertheless, low-dose effects, which are most important for risk assessment, are much more difficult

to discern compared to the effects at high doses given limited statistical power and results that are

highly susceptible to bias. Therefore, it is difficult to draw conclusions about the risk deficit observed in

the low-dose range in this study and further interpretation merits caution.

Among the leukemia subtypes, positive, non-significant dose-responses were observed in

adjusted models of the myeloid leukemias but not with CLL (Table 6-20). Other studies examining the

leukemia risk from protracted exposures have reported similar results although these studies are not

completely independent [Cardis et al. 1995; Cardis et al. 2007; Muirhead et al. 2009]. Statistically

significant CML dose-responses were found in the Three-Country study (i.e., US, UK, and Canada)

conducted by Cardis et al. [1995] and the most recent update to the study of British nuclear workers

[Muirhead et al. 2009]. Likewise, statistically significant and positive radiation dose-responses for AML

and CML were reported in studies of the A-bomb survivors [Preston et al. 1994; Richardson et al. 2009].

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Consistent with previous studies of populations exposed to known leukemogens such as benzene [Finkelstein 2000; Silver et al. 2002], ionizing radiation [Richardson et al. 2009; Richardson and

Wing 2007; Schubauer-Berigan et al. 2007a; Shilnikova et al. 2003] and smoking [Richardson et al. 2008], wave-like patterns of leukemia risk with time since exposure were apparent in most models in the current study [Schubauer-Berigan et al. 2007b; Shilnikova et al. 2003]. For example, the previous study by Schubauer-Berigan et al. [2007a] estimated the non-CLL leukemia mortality risk at ERR∙10 mSv -1= 0.32

( 95% CI: <0, NC) in a linear model with doses accumulated in the best-fit 5-10 year window. Likewise,

the current study saw marked improvement in the risk estimate for non-CLL leukemia in the 5-10 year

exposure window (ERR ∙10 mGy -1= 0.20; 95% CI:-0.032, 0.68), where a linear model provided the best fit

to the data compared to linear-quadratic, categorical, and RCS models tested. Moreover, risk

attenuation in the low dose range that was observed in main models was no longer evident in the

windows-based models. Similar results were obtained from leukemia subtype analyses. The AML dose-

response (ERR∙10 mGy -1 = 0.019; 95% CI: -0.015, 0.089) was best fit using a two-year lag while the dose-

response for CML (ERR∙10 mGy -1 = 0.029; 95% CI: <0, 0.18) was fit using a five-year lag period. Both

subtypes were influenced by time since exposure, where the greatest change (40-fold) was observed in

AML risk in the 5-10 year exposure window (ERR∙10 mGy -1 = 0.76; 95% CI: 0.047, 2.7) compared to CML

(15-fold) in the 10-15 year exposure window (ERR∙10 mGy -1 = 0.44; 95% CI: <0, 2.4). This effect of time since exposure on AML risk was also apparent in the LSS; however, some have suggested this effect was mostly limited to persons who were less than 30 years old at exposure [Richardson et al. 2009].

In addition to risk attenuation in the high dose range, the depletion of susceptible persons may also explain the short latency and differing risk with time since exposure that is common to radiogenic leukemias. Nakamura [2005] has theorized that nearly all radiation-related leukemia is attributable to a small group of individuals who are predisposed to substantial numbers of preleukemic cells. In chronically exposed populations, those most susceptible to the disease present early after relevant

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exposure, and persons who are resistant (i.e., the majority) remain leukemia free and continue to accumulate dose, which is postulated to result in the perceived attenuation of risk at high doses and the

dilution of risk with increased followup. If this theory bears true, it is prudent to consider occupational

protective measures based on the risks observed in the most relevant exposure period, rather than the

entire exposure period, in efforts to better protect those who are most susceptible to radiation-induced

leukemia.

We did not observe significant rate ratio modification by the covariates tested except for

facilities within the LCCS Group. Significant risk differences between these facilities were also observed

in the previous study; however, only birth cohort (which also modified the rate ratio) was selected as a

stratifier in subsequent models [Schubauer-Berigan et al. 2007a]. These facility differences reduced the precision of risk estimates and may reflect biases within facilities that cannot be assessed or accounted for in the current study design. A meta-analytic approach that weights data prior to pooling and uses random effects models to adjust for heterogeneity across facilities may improve risk estimates [Daniels and Schubauer-Berigan 2011].

Benzene, a known leukemogen and a confounder in the previous study, was not associated with leukemia risk and did not confound results in the current study. In the absence of smoking data, SES was

used to potentially adjust for smoking-related effects or other lifestyle factors. There was no evidence of

significant risk differences by SES level or confounding of radiation risk and ultimately SES was not

included in any model. Sex, race, and birth date (AML, CLL) or hire date (all other models) were the only

confounders chosen based on a priori selection criteria.

Interestingly, birth cohort effects were strongest in subtypes that primarily affected the elderly

(i.e., AML and CLL). Most doses accumulated in the early years of facility operations; therefore, an

inverse correlation between birth cohort and dose is expected. For example, the Spearman correlation

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coefficient for birth cohort (tertiles) and low-LET dose (five categories) for AML under a two-year lag was

-0.19, P<0.001. We also noted in univariate analysis that leukemia risk increased with birth year. Levi et al. [2000] observed that the trend in U.S. age-standardized leukemia rates (per 100,000) in elderly persons (age ≥ 70 years) has risen 23% from 1960-1964 (53.4) to 1995-1997 (65.8), although analysis by subtype was not reported. Similarly increasing adult leukemia trends were observed in Denmark from

1943 to 2003 [Thygesen et al. 2009], where AML and CLL age-specific rates were increasing while CML was slightly decreasing over time (1943-2003). These differences seem to coincide with changes in diagnosis, especially for AML, which is the most prevalent cause of leukemia in this study; therefore,

AML mortality during the early period of observation may be underestimated. It is also possible that some difference in environmental risk factors may partly explain the increasing leukemia trend. For example, the NCRP (2009) has estimated that average ionizing radiation exposure to persons in the U.S. has nearly doubled from the early 1980s to 2006, primarily as a result of the growing use of medical imaging procedures. Of course, one would expect that concomitant decreases in environmental benzene levels due to use restrictions would tend to partially or wholly offset increasing risks from ionizing radiation exposure in diagnostic or interventional medicine.

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Table 6-20. Leukemia Excess Relative Risk (ERR) per 10 mGy exposure by subtype and by study

Study Overlap AML CML CLL Current Study Three -County Study 0.019 0.029 -0.00 3 (Hanford, ORNL) (95% CI: -0.015, 0.089) (95% CI: <0, 0.18) (95% CI: -0.015, 0.051) Three -Country NRRW 0.033 0.11 -0.009 Study [Cardis et Current study (90% CI: <0, 0.149) (90% CI: 0.029, 0.309) (90% CI: <0, 0.073) al. 1995] (Hanford, ORNL) NRRW Three -Country Study 0.012 0.033 <-0.019 [Muirhead et al. (90% CI: -0.012, 0.065) (90% CI: 0.004, 0.093) (90% CI: <-0.019, 0.012) 2009] LSS NA 0.043 0.064 NR [Richardson et al. (90% CI:0.027, 0.066) (90% CI:0.030, 0.137) 2009] AML= Acute Myelogenous Leukemia; CLL= Chronic Lymphocytic Leukemia; CML= Chronic Myelogenous Leukemia; LSS= Life-Span Study of Japanese Atomic Bomb Survivors NA= not applicable; NR=not reported; NRRW= National Registry of Radiation Workers (UK)

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The study extended followup of those previously studied by Schubauer-Berigan et al. [2007a;

2007b] and was expanded to include INL workers (Table 6-20). The aggregate data used to examine leukemia excluding the CLL subtype were best fit to a linear model using a two-year exposure lag, which estimated the ERR per 10 mGy absorbed dose to bone marrow at 0.009 (95% CI:-0.014, 0.051). Although confidence intervals overlapped, the current point estimate is nearly three-fold less than that obtained from the earlier study (ERR∙10 mGy -1= 0.026; 95% CI: <-0.010, 0.103). Part of the risk difference was due to lower risk in INL workers compared to the LCCS Group. Excluding the risk contribution from INL workers (ERR∙10 mGy -1= 0.0004; 95% CI: <0, 0.075) increased the risk estimate by nearly two-fold

(ERR∙10 mGy -1= 0.015; 95% CI: -0.012, 0.062). It was also noted that the former study defined exposure in terms of dose equivalent, which combined the contributions from low- and high-LET radiations using weights of unity for low-LET, 10 for neutron exposures and 20 for doses from plutonium. Adjusting the current study results to dose equivalent using the same weighting scheme and including high-LET doses reduces the relative percent change from approximately -42% to -20% (ERR∙10 mGy -1= 0.021; 95% CI: -

0.005, 0.0785); therefore, these design differences do not entirely account for the deficit observed.

Another difference from the earlier study was the current exclusion of MED Era workers; therefore, information on 123 leukemia cases hired prior to 1950 was not used in risk estimation. These earlier cases were more likely to have been exposed than workers hired after 1950, which may explain some of the remaining differences in estimates. The exclusion was intended to reduce exposure misclassification thought to have caused risk attenuation in the high dose range that was observed in the previous study; however, there was no discernable difference in risk at high doses when comparing the two studies. Future studies that include these cases may improve the overall model fit of dose- response models.

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6-21. Comparison of the results on Non-CLL leukemia risk obtained in the current study to the previous study.

Study Model Description Study Cohort Units and dose Risk Estimate Current Study Linear ERR model Hanford, INL, LANL/Zia, ORNL, PNS, ERR ∙10 mGy -1 fro m Low - 0.009 ( -0.014, 0.051) Non-CLL leukemia and SRS workers (n=105,246) LET only (n=264) under a two followup through 2005 year lag, adjusted for race, sex, and hire year

Restricted to Hanford, LANL/Zia, ERR ∙10 mGy -1 from Low - 0.015 ( -0.012, 0.062) ORNL, PNS, and SRS workers (n=200 LET only cases)

Restricted to Hanford, LANL/Zia, ERR ∙10 mSv -1 from Low - 0.021 ( -0.005, 0.0785) ORNL, PNS, and SRS workers (n=200 LET, neutron and cases) and exposure as dose Plutonium. equivalent (radiation weights of 10 and 20 for neutron and plutonium, respectively)

Schubauer -Berigan et Linear ERR model of Hanford, LANL/Zia, ORNL, PNS, and ERR ∙10 mSv -1 from Low - 0.0 26 (<-0.01 0, 0.103) al. [2007a] non-CLL leukemia SRS workers (n=94, 517) followed LET, neutron and (n=206) adjusted for through 1990 (ORNL, and LAN/Zia) Plutonium sex, benzene and 1994 (Hanford and SRS), and 1996 excluding 22 (PNS) ambiguous cases

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Diminishing summary leukemia risk estimates with continued followup have been observed in populations exposed to benzene [Rinsky et al. 2002; Silver et al. 2002] and ionizing radiation [Boice et al.

2011; Krestinina et al. 2010]. Rinsky et al. [2002] reported a decline in the SMR from 3.37 (95% CI: 1.54,

6.41) to 2.56 (95% CI: 1.43, 4.22) after adding 15 years of followup to a cohort of white male benzene- exposed rubber hydrochloride workers. Silver et al. [2002] also examined these workers using yearly followup information and found a five-fold decrease in the risk estimates from 1961 to 1996. Perhaps more noteworthy is that Silver et al. [2002] , using conditional logistic regression and exposure window analysis, found temporal patterns of leukemia risk that mimic current radiation-related risks, whereby peak risks were associated with exposures occurring five to ten years prior to the age at death of the case. In regard to radiation exposed cohorts, Boice et al. [2011] reported decreased non-CLL leukemia mortality after adding nine years of followup for the study of Rocketdyne radiation workers, as did

Krestinina et al. [2010] after adding six years of followup to the Techa River Cohort. Exposure profiles for these populations are temporally similar, where most significant exposures occurred earlier in the followup period and continued followup resulted in negligible contributions to collective dose. In the current study, vital status was extended at least nine years for all facilities. Although more cases should improve statistical power, the new cases are likely to have been hired later, thus have lower exposures relative to earlier cases (and controls) due to temporal exposure patterns in most facilities. Moreover,

new cases with significant dose are more likely to have been exposed many years prior to the onset of

disease; thus their radiation exposure is less likely to be causal given that non-CLL leukemia risk tends to

decrease with increasing time since exposure (minus any latency period). Therefore, extended followup

of nuclear worker cohorts that maintain current radiation exposures ALARA is unlikely to uncover new

information on leukemia risk and may actually dilute risk estimates. Moreover, examinations of

temporal patterns of risk appear to be critical in any future studies examining risk factors for leukemia.

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We attempted to examine the effects from low- and high-LET radiations without a priori specification of radiation weights. The results did not confirm the weighting scheme currently recommended by the ICRP; instead, there were indications that larger weights were needed for both high-LET dose components to improve model fit using maximum likelihood methods, although results were imprecise. This finding was in stark contrast to our current knowledge on plutonium leukemogenicity, which suggests that the RBE for plutonium alpha radiation is far below the ICRP recommended value of 20, and perhaps on the order of 1 to 3 [Grogan et al. 2001; Puskin et al. 1999;

Spiers and Vaughan 1989]. It is conceivable that the large weights were a result of other characteristics that were unique to the few plutonium workers or persons with measurable neutron dose, resulting in a risk profile that is different for reasons other than radiation exposure.

Remarkably, there was little effect on the radiation risk attributable to low-LET radiations with the inclusion of the high-LET dose components in adjusted models allowed to solve for radiation weights

(e.g., -22% in the adjusted non-CLL model). Therefore, the addition of the high-LET dose contribution was not informative on the overall effects on leukemia from ionizing radiation. Fixing the weights at

ICRP recommended values increased risk estimates for non-CLL leukemia by approximately 67% on a relative scale. Most of this change was attributable to the plutonium dose component; holding the plutonium weight at unity resulted in a modest increase (22%). Additional sensitivity analyses (results not shown) revealed that peak estimates of risk were obtained with plutonium weights fixed in the range of 20 to 30 in combination with neutron weights fixed between 30 and 50. Weights outside of these ranges had negligible effect on risk estimates. Therefore, using the radiation weighting factors recommended by the ICRP did not improve model fit and may have led to an overestimation of the radiation effect, based on comparisons with models that allowed for weight solutions.

Although considerable efforts were undertaken to reduce uncertainties in dose, exposure misclassification is unavoidable and may have led to biased estimates. Exposure records were

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incomplete, especially in the early periods of facility operation when doses were highest. The quality of

exposure information varied across sites and within sites over time. The quality of dose estimates also

varied by radiation type. Individualized measurements of other leukemogenic agents (e.g., benzene and

smoking) that may confound the association between radiation and leukemia were not available for this

study.

7.0 Conclusions

This case-control study of leukemia radiogenicity from protracted occupational radiation exposure is the largest ever conducted. Positive, but imprecise estimates of the dose response were observed for all leukemia outcomes except the CLL subtype, although the observed dose response showed attenuated risk in the low dose (<10 mGy) and high dose (>100 mGy) regions. Risk attenuation in the low-dose region is particularly difficult to explain given that similar results were not observed in previous studies; however, there is some evidence of exposure misclassification in the low dose range that may at least partially explain this observation. Future research should be focused to better evaluate the shape of the dose-response, particularly in the low-dose range which is critical for risk assessment purposes. Rate ratio modification was not observed in any model, although risk estimates in most linear models tested were confounded by sex, race, and hire date or birth date. This confounding disappeared in piecewise models that provided a better fit to the complex dose-response that were characteristic of full models.

Leukemia risks were also characterized by a “wave-like” function of time, where peak excess risk per unit dose was observed at certain exposure lags or windows of exposure prior to disease onset.

Marked improvements in estimates from ERR models were observed in the best fit exposure windows, where linear dose-response models prevailed over other forms tested for non-CLL, AML, and CML leukemia. These observations provided evidence of a monotonic increase in risk with dose accrued in

283

the most relevant period of exposure, where risks of radiogenic leukemia appear relatively high compared to estimates over the full observation period. These periods were dependent upon leukemia subtype, where shorter periods of latency were traits of AML compared to chronic leukemias.

It is unlikely that continued followup of this cohort will be informative on leukemia risk given that new cases are likely to have little radiation exposure and leukemia risk appears to diminish with time since exposure. Study power may be improved with the addition of other nuclear worker cohorts, although cautious pooling is warranted because of the evidence of risk heterogeneity across facilities in the current study. The inclusion of incident cases or non-underlying causes of death may also improve study power and should be considered in future analysis. Moreover, continued improvement in exposure assessment methods may greatly reduce uncertainties that may have biased risk estimates in this study.

284

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Appendix I: Contributions

Corresponding Author: Robert D. Daniels, 4676 Columbia Pkwy., R-14, Cincinnati OH 45226. Phone: (5 13) 458-7178. Fax (513) 841-4470. Email: [email protected] .

Author Affiliation: Division of Surveillance, Hazard Evaluations, and Field Studies (DSHEFS), National Institute for Occupational Safety and Health (NIOSH), U.S. Centers for Disease Control and Prevention (CDC).

Study concept: R. Daniels, S. Silver (NIOSH), M. Schubauer-Berigan (NIOSH). Originally titled “Evaluating Occupational Ionizing Radiation Exposure Standards” and submitted as a proposal for funding under the NIOSH National Occupational Research Agenda (NORA), Project ID: NORA FY08 AUR 0001 Silver.

Study design: R. Daniels with input from Dissertation Committee members

Protocol Development: R. Daniels, T. Kubale (NIOSH)

Data: collection: R. Daniels, T. Kubale

Vital status ascertainment: R. Daniels, T. Kubale, K. Waters (NIOSH), P. Laber (NIOSH), C. Gersic (NIOSH)

Exposure assessment: R. Daniels

Data management and programming: K. Waters, R. Daniels

Statistical analysis and interpretation of data: R. Daniels with input from Dissertation Committee members, and S. Bertke (NIOSH)

Report writing: R. Daniels with input from Dissertation Committee members

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