Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute for Occupational Safety and Health 

Review of the Santa Susana Field Laboratory (SSFL) Area IV 

Special Exposure Cohort (SEC) Petition-00093 

and the NIOSH SEC Petition Evaluation Report 

Contract No. 200-2004-03805 

Task Order No. 5 

SCA-SEC-TASK5-0066

Prepared by 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, Virginia 22182 

Saliant, Inc. 

5579 Catholic Church Road 

Jefferson, Maryland 21755 

January 2009 

Disclaimer 

This document is made available in accordance with the unanimous desire of the Advisory Board 

on Radiation and Worker Health (ABRWH) to maintain all possible openness in its 

deliberations. However, the ABRWH and its contractor, SC&A, caution the reader that at the 

time of its release, this report is pre-decisional and has not been reviewed by the Board for 

factual accuracy or applicability within the requirements of 42 CFR 82. This implies that once 

reviewed by the ABRWH, the Board’s position may differ from the report’s conclusions. Thus, 

the reader should be cautioned that this report is for information only and that premature 

interpretations regarding its conclusions are unwarranted.
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Document No. 

SCA-SEC-TASK5-0066 

Effective Date: 

Draft — January 30, 2009 

S. COHEN & ASSOCIATES:

Technical Support for the Advisory Board on 

Radiation & Worker Health Review of 

NIOSH Dose Reconstruction Program Revision No. 

Review of the Santa Susana Field Laboratory (SSFL)- 

Area IV Special Exposure Cohort (SEC) Petition-00093 

and the NIOSH SEC Petition Evaluation Report 

Page 2 of 54 

Task Manager: 

________________________ Date: ___________ 

Gregory P. Beronja 

Project Manager: 

________________________ Date: ___________ 

John Mauro, PhD, CHP 

Supersedes: 
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TABLE OF CONTENT 

Abbreviations and Acronyms ..........................................................................................................5 

1.0 Introduction..........................................................................................................................8 

1.1 Scope and Purpose of SEC Review .........................................................................8 

1.2 Technical Approach and Review Criteria................................................................8 

1.3 Organization of the Report.......................................................................................9 

2.0 Key Information and Data Presented in the SSFL Site Profile..........................................11 

2.1 Principal Operations...............................................................................................11 

2.2 Development and Testing of Nuclear Reactors .....................................................12 

2.2.1 Homogeneous Water Boiler Reactors........................................................12 

2.2.2 Sodium-Cooled Graphite-Moderated Reactors..........................................12 

2.2.3 Systems for Nuclear Auxiliary Power Reactors ........................................13 

2.3 Critical Test Facilities............................................................................................13 

2.4 Nuclear Support Operations...................................................................................13 

2.4.1 Reactor Fuel Manufacturing ......................................................................13 

2.4.2 Disassembly and Examination of Reactors and Used Reactor Fuel 

Assemblies.................................................................................................13 

2.4.3 Fabrication, Use, and Storage of Radioactive Sources..............................14 

2.4.4 Preparation of Radioactive Material for Disposal .....................................14 

2.4.5 Research on Reprocessing Used Reactor Fuel...........................................14 

2.4.6 Operation of Particle Accelerators.............................................................14 

2.4.7 Research Using Radioisotopes...................................................................14 

2.4.8 Miscellaneous Operations..........................................................................14 

2.5 Other Relevant Facilities........................................................................................15 

2.5.1 Downey......................................................................................................15 

2.5.2 Canoga Park...............................................................................................15 

2.5.3 De Soto.......................................................................................................15 

2.6 Radionuclides of Concern......................................................................................15 

3.0 Overview of the SEC Petition-00093 and NIOSH’s Evaluation Report ...........................16 

3.1 SEC Petition-00093 ...............................................................................................16 

3.2 SEC Petition Evaluation Report.............................................................................17 

3.3 NIOSH’s Response to Major Issues Raised in SEC-00093...................................18 

3.3.1 Responses to Six Issues Identified in the SEC Petition and During 

Discussions with the Petitioner..................................................................18 

3.3.2 Other Issues................................................................................................21 

3.4 NIOSH’s Conclusions............................................................................................22 

4.0 Preliminary Findings Associated with the SEC Petition ...................................................23 

4.1 Overview................................................................................................................23 

4.2 Areas Considered in the SEC Petition and Coverage Dates..................................23 

4.2.1 Areas Addressed by Petition......................................................................23 

4.2.2 Coverage Dates ..........................................................................................25 

4.3 Adequacy of the Internal Monitoring Program......................................................26 

4.3.1 Overview of Internal Monitoring Program................................................26 
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4.3.2 Monitoring for Internal Exposure of SSFL Workers was Incomplete 

and Poorly Documented for Most Years of Facility Operation.................29 

4.3.3 Insufficient Correlation between Bioassay Data and Potential 

Exposures to Specific Radionuclides.........................................................31 

4.3.4 Missing Radionuclides in Bioassay Data...................................................31 

4.3.5 No Coworker Model ..................................................................................32 

4.3.6 Lack of Source Term Data.........................................................................32 

4.3.7 NIOSH’s Interpretation of Reported Values by the Contract 

Laboratory Nuclear Science and Engineering Corp. (NSEC) may not 

be Correct...................................................................................................32 

4.3.8 Potential Difficulties Associated with Uranium Bioassay Data ................33 

4.3.9 There are Unanswered Questions Regarding the Completeness and 

Quality of Personnel Exposure Records ....................................................34 

4.3.10 Use of SSFL Site Survey Data/Source Term cannot be Regarded as 

Useful Surrogate Data for Bioassay Data in Dose Reconstruction............34 

4.4 Lack of Information with Certain Facility Incidents .............................................35 

4.4.1 Overview....................................................................................................35 

4.4.2 Sodium Reactor Experiment Coolant Failure............................................35 

4.4.3 The Sodium Burn Pit .................................................................................37 

4.5 Lack of Information on the Environmental Exposures..........................................38 

4.5.1 Overview....................................................................................................38 

4.5.2 Surrogate Use of the Time-integrated Average Yearly Gross 

Alpha/Gross Beta Stack Emission Corresponding to Years 1971 to 

1999............................................................................................................39 

4.5.3 Exposure to Onsite Water Supply Wells ...................................................39 

4.6 Assignment of External Dose ................................................................................41 

4.6.1 No Coworker Model ..................................................................................41 

4.6.2 Workers Were Unlikely to Have Been Monitored for Thermal 

Neutrons.....................................................................................................41 

4.6.3 Dosimeter Response to Low-Energy Photons ...........................................42 

5.0 Conclusions........................................................................................................................43 

6.0 References..........................................................................................................................45 

Appendix A: Petitioner/Worker Interview....................................................................................54 

 
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ABBREVIATIONS AND ACRONYMS 

ABRWH Advisory Board on Radiation and Worker Health 

or the Board 

AEC Atomic Energy Commission 

AERD Atomic Energy Research Development 

AWE Atomic Weapons Employer 

AI Atomics International 

ATR Advanced Test Reactor 

BE Beryllium 

CFR Code of Federal Regulations 

Ci Curie 

cpm Counts per minute 

D&D Decontamination and Decommissioning 

DOE Department of Energy 

DOELAP Department of Energy Laboratory Accreditation Program 

dpm Disintegrations per minute 

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000 

EPA Environmental Protection Agency 

ER Evaluation Report 

ETEC Energy Technology Engineering Center 

EU Enriched Uranium 

FR Federal Register

FWSP Federal Work Study Program 

HEPA High Efficiency Particulate Air 

HHS Health and Human Services 

ISF Interim Storage Facility 

keV Kilo Electron Volt 

L Liter 

LMEC Liquid Metal Engineering Center 

LMFBR Liquid Metal Fast-Breeder Reactor 

LMIC Liquid Metal Information Center 

LMR Liquid Metal Reactor 
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MDA Minimum Detectable Activity 

MDL Minimum Detection Limit 

MeV Mega Electron Volt 

MFP Mixed Fission Product 

mL Milliliter 

MPBB Maximum Possible Body Burden 

Mr Millirem 

NAA North American Aviation 

NIOSH National Institute for Occupational Safety and Health 

NOCTS NIOSH OCAS Claims Tracking System 

NSEC Nuclear Science and Engineering Corporation 

NTA Neutron Track Emulsion Type A 

OCAS Office of Compensation Analysis and Support 

ORAU Oak Ridge Associated Universities 

ORAUT Oak Ridge Associated Universities Team 

pCi pico Curie 

r Rem 

R&D Research and Development 

REIRS Radiation Exposure Information Reporting System 

Rem roentgen equivalent man 

RI Rockwell International 

RMDF Radioactive Material Disposal Facility 

RSRMS Radiation Safety Records Management System 

SC&A, Inc. S. Cohen and Associates 

SEC Special Exposure Cohort 

SRDB Site Research Database 

SRE Sodium Reactor Experiment 

SNAP Systems for Nuclear Auxiliary Power 

SSFL Santa Susana Field Laboratory 

TBD Technical Basis Document 

UCLA University of California Los Angeles 

UF Uranium Fluoride 
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UR Uranium Radiometric 

W Watt 

WB Whole Body 
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1.0 INTRODUCTION 

1.1 SCOPE AND PURPOSE OF SEC REVIEW 

During the meeting of the Advisory Board on Radiation and Worker Health (the Board) held in 

Redondo Beach, California, on September 4, 2008, S. Cohen & Associates (SC&A) was directed 

by the Board to perform a “paper” review of the Santa Susana Field Laboratory (SSFL) Special 

Exposure Cohort (SEC) Petition-00093 and the NIOSH SEC Petition Evaluation Report (ER) for 

said petition. 

The scope of this review addresses specific issues of concern raised in the petition and NIOSH’s 

response to these concerns, as given in the ER. SC&A reviewed hundreds of documents that 

were considered relevant to the petition. Documents reviewed include the following: 

• Documents that were referenced and/or enclosed in the petition 

• Documents referenced/cited in the ER and site profile 

• Documents contained in the NIOSH Site Research Query Database 

The purpose of this review is to provide the Board with an independent assessment of issues and 

concerns that surround the petition and NIOSH’s response and proposed methods for 

accommodating these issues/concerns. Findings identified in our review are expected to provide 

the Board with a preliminary overview of potential issues that may impact the feasibility of 

dose assessment. Following a formal, multi-step resolution process, any unresolved findings 

may then be used by the Board for determining whether radiation doses can be estimated with 

sufficient accuracy, as defined in 42 CFR §83.13(c)(1). In addition, since this review was 

limited to a paper study, there will likely be a need to follow up this report with a site visit, 

interviews with claimants and petitioners, and the retrieval and review of additional documents, 

in accordance with Board-approved procedures for the review of SEC petitions and associated 

evaluation reports. It is for this reason that this report is referred to as “preliminary” and as a 

“paper study.” 

1.2 TECHNICAL APPROACH AND REVIEW CRITERIA 

The approach used by SC&A to perform this review follows the protocols described in the draft 

report prepared by SC&A entitled, Board Procedures for Review of Special Exposure Cohort 

Petitions and Petition Evaluation Reports, Revision 1 (SC&A 2006a), and the Report to the 

Working Group on Special Exposure Cohort Petition Review (SC&A 2006b). The latter is a set 

of draft guidelines prepared by a Board-designated working group for evaluation of SEC 

petitions performed by NIOSH and the Board. The former is a set of draft procedures prepared 

by SC&A and approved by the Board for use by SC&A on an interim basis (ABRWH 2006, 

p. 132). The procedures are designed to help ensure compliance with Title 42, Part 83, of the 

Code of Federal Regulations (42 CFR 83) and implement the guidelines provided in the report of 

the working group. 
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Key review criteria identified in the report of the working group include the following: 

• Timeliness 

• Fairness 

• Understandability 

• Consistency 

• Credibility and validity of the dataset, including pedigree of the data, methods used to 

acquire the data, relationship to other sources of information, and internal consistency 

• Representativeness and completeness of the exposure data with respect to the area of the 

facility, the time period of exposure, the types of workers, and processes covered by the 

data 

The working group guidelines also recommend that NIOSH include in its SEC evaluation a 

demonstration that it is feasible to reconstruct individual doses for the cohort, including sample 

dose reconstructions. 

SC&A’s implementation of the SEC Review process includes the following steps: 

(1) Conduct a critical review of the petition and relevant reports, as well as documents and 

data that are enclosed and/or referenced in the petition/reports. 

 

(2) Identify additional issues/concerns that emerged from SC&A’s document review, which 

are independent of those stated in the petition. 

(3) As part of the SEC review, develop a preliminary technical position for issues identified 

in the petition, as well as SC&A’s independent findings. 

SC&A’s draft report with its preliminary findings will subsequently undergo a multi-step 

resolution process, including interviews with site workers and others who can provide insights 

into the issues of concern. Resolution includes a transparent review and discussion of draft 

findings with members of the Board’s working group, petitioners, claimants, and interested 

members of the public. This resolution process is intended to ensure that each finding is 

evaluated on its technical basis in a fair and impartial manner. A final report will then be issued 

to the full Board for deliberation and a final recommendation. 

1.3 ORGANIZATION OF THE REPORT 

Following this introduction, Section 2.0 of this report provides summary data contained in the 

SSFL Site Profile. The site profile specifies relevant background information and methods to be 

used by NIOSH for the reconstruction of internal and external doses, and includes brief site 

profile summaries of materials and quantities processed, facility descriptions, and proposed 

methods for dose reconstruction. 
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Section 3.0 of this report summarizes specific concerns and issues raised in the SEC Petition00093, as well as NIOSH’s ER of the petition. In the ER, NIOSH provided responses to the 

petition’s concerns, along with the conclusion that dose reconstruction is not feasible for SSFL 

workers for the years 1955 through 1958. 

As a result of our review of the petition, NIOSH’s evaluation of the petition, the SSFL Site 

Profile, and other documents, SC&A identified a total of five major findings (that in some cases 

are supported by more detailed findings), which are cited in Section 4.0 of this report. A 

discussion is provided for each finding that serves to explain the technical basis for our concern. 

• Subsection 4.1: Which areas (Area IV, Canoga Park, DeSoto, and Downey) should be 

considered in the SEC Petition and when did operations begin at Area IV 

• Subsection 4.2: Concern whether the internal monitoring program was sufficiently robust 

to estimate exposures before and after January 1, 1959 

 

• Subsection 4.3: Lack of information related to the potential exposures associated with 

facility “incidents” (Sodium Reactor Experiment, Sodium Burn Pit, etc.) 

• Subsection 4.4: Lack of information on the environmental exposures (surrogate data, 

drinking water, etc.) 

• Subsection 4.5: Justification for assignment of external dose estimates is not provided 

and there is no coworker model for external exposures (personnel records are also of 

concern) 

Section 5.0 provides concluding comments regarding the impacts of our findings on dose 

reconstruction for SSFL workers.
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2.0 KEY INFORMATION AND DATA PRESENTED IN THE SSFL SITE 

PROFILE 

The most current site profile for the Santa Susana Field Laboratory (SSFL) consists of six 

Technical Basis Documents (TBDs) that were issued at various times (see References for 

specific dates). These TBDs will be referenced throughout this report and include the following: 

• ORAUT-TKBS-0038-1: Introduction 

• ORAUT-TKBS-0038-2: Site Description 

• ORAUT-TKBS-0038-3: Occupational Medical Dose 

• ORAUT-TKBS-0038-4: Occupational Environmental Dose 

• ORAUT-TKBS-0038-5: Occupational Internal Dose 

• ORAUT-TKBS-0038-6: Occupational External Dose 

Throughout this report, individual TBDs are referenced simply by number. For example, 

ORAUT-TKBS-0038-1 will be identified as TBD-1. Collectively, the six TBDs of the site 

profile are intended to provide core information, data, and guidance that are intended to assist in 

the dose reconstruction of individual workers who may have been exposed to internal and 

external occupational radiation at SSFL as stated in ORAUT-TKBS-0038-1: 

. . . The purpose of this document is to provide a site profile that contains 

technical basis information for evaluation of the total occupational dose for 

EEOICPA claimants who were employed at the Atomics International (AI) facility 

as described above.

2.1 PRINCIPAL OPERATIONS 

SSFL consists of a total of 2,850 acres and is located in the Simi Hills of Ventura County, 

approximately 30 miles northwest of downtown Los Angeles, California. Based on ownership 

and operations, SSFL is divided into four administrative and operational portions—Area I, 

Area II, Area III, and Area IV. DOE operations are conducted in a 290-acre westernmost 

administrative and operational portion designated as Area IV. 

SSFL was initially established in 1947 by North American Aviation (NAA) to meet the 

requirements for a field test laboratory to static-fire large rocket engines; however, it also met 

NAA’s need for a nuclear research facility. As a result, Area IV was established in 1953 at 

SSFL as a nuclear research and development (R&D) facility. Since then, SSFL has housed both 

nuclear development and rocket development groups, although in distinct and separate locations. 

Atomic Energy Research Development (AERD) also conducted operations in SSFL-Area IV. In 

December 1955, the nuclear development and rocket development groups were transformed into 

separate divisions—Atomics International (AI) and Rocketdyne. 

Two distinct AI groups were housed in Area IV and supported by DOE. One focused on 

development of civilian nuclear power, and the other was a center of excellence for research and 

testing of non-nuclear components related to liquid metals. These two groups were referred to as 

AI and Liquid Metal Engineering Center (LMEC), respectively. Nuclear R&D activities in 
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Area IV increased rapidly from 1953 into the late 1960s, and then began to decline. AI was 

eventually merged into Rocketdyne in 1984 as a result of this decline. 

The LMEC was created in 1966 as a government-owned and contractor-operated organization; 

its purpose was to provide development and non-nuclear testing of Liquid Metal Reactor (LMR) 

components and to establish the Liquid Metal Information Center (LMIC) for the Atomic Energy 

Commission’s (AEC’s) Liquid Metal Fast-Breeder Reactor (LMFBR) program. The LMEC was 

renamed Energy Technology Engineering Center (ETEC) in 1978 to reflect DOE’s desire to 

broaden its mission beyond the LMFBR program. 

Several corporate mergers and organizational changes occurred over the years. In 1967, NAA 

merged with Rockwell Standard to become North American Rockwell. In 1973, the corporate 

name changed to Rockwell International (RI). Rockwell International with AI and Rocketdyne 

continued to exist as independent divisions until 1984, when AI was absorbed by the Rocketdyne 

division. The Boeing Company purchased RI in 1996, and Rocketdyne is now a division of 

Boeing. 

Before the remaining research activities ended in 1998, three primary types of operations were 

conducted at Area IV: (1) development and testing of nuclear reactors, (2) nuclear support 

operations, and (3) non-nuclear energy R&D. 

2.2 DEVELOPMENT AND TESTING OF NUCLEAR REACTORS 

Between 1953 and 1980, several nuclear reactors were built, tested, and operated in Area IV. 

These included both nuclear reactors and critical test assemblies. Nuclear reactor programs 

focused on the development and operation of homogeneous water boiler-type reactors, sodiumcooled graphite-moderated reactors, and uranium-zirconium hydride reactors. 

2.2.1 Homogeneous Water Boiler Reactors 

The water boiler reactors were operated in Buildings 4073 and 4093. The water boiler reactors 

used a 93% enriched uranyl sulfate solution held in a critical configuration in a spherical vessel. 

Rather than actually boil, the neutron and gamma flux caused radiolytic decomposition of water 

into hydrogen and oxygen in the form of tiny bubbles, which gave the impression of boiling. 

Area IV contained two water boiler reactors. 

2.2.2 Sodium-Cooled Graphite-Moderated Reactors 

The Sodium Reactor Experiment (SRE) facility consisted of 12 structures, including the reactor 

building, office buildings, and support structures. Eight structures were directly involved in 

operations with radioactive materials: 

(1) Reactor Building (Building 4143) 

(2) Component Storage Building (Building 4041) 

(3) Temporary Hot Waste Storage Building (Building 4686) 

(4) Site Service Building (Building 4163) 
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(5) Cold Trap Vault (Building 4695) 

(6) Liquid Radioactive Waste Vault (Building 4653) 

(7) Interim Radioactive Waste Storage Area (Area 4654) 

(8) Intermediate Contaminated Storage Area (Area 4689) 

2.2.3 Systems for Nuclear Auxiliary Power Reactors 

The Systems for Nuclear Auxiliary Power (SNAP) program operated from 1956 to 1971 to 

support the development and testing of small reactors designed to provide power for research 

missions in space. The SNAP reactors were uranium-zirconium hydride reactors that used fully 

enriched (93%) uranium dispersed in fuel rods containing zirconium hydride. Seven SNAP 

reactors were tested and operated in Buildings 4010, 4024, 4028, and 4059. 

2.3 CRITICAL TEST FACILITIES 

Several programs used critical test facilities (i.e., low-power reactors) in Area IV. Use of these 

low-power reactors began in 1954, and continued until 1974. The critical test facilities included 

SNAP development test facilities, which were housed in Buildings 4373, 4012, 4019, and 4024. 

Critical test facilities supporting the development of civilian nuclear power included Buildings 

4009 and 4100. 

2.4 NUCLEAR SUPPORT OPERATIONS 

Starting in 1956, several operations were conducted in Area IV to support nuclear programs. 

These included the manufacture, management, and disassembly of fuel for reactor operations, as 

well as the operation of nuclear waste management facilities for offsite disposal. 

2.4.1 Reactor Fuel Manufacturing 

As part of the nuclear reactor development work performed for the government, three different 

reactor fuel manufacturing operations occurred at the SSFL in Buildings 4003, 4055, and 4064. 

The first operation was the assembly of fuel elements for the SRE, the second was a plutonium 

fuel manufacturing facility, and the third was a uranium carbide fuel pilot plant. There was also 

a Fuel Storage Facility, used to store the Special Nuclear Materials (enriched uranium and 

plutonium) used to make the fuels. 

2.4.2 Disassembly and Examination of Reactors and Used Reactor Fuel Assemblies 

During reactor test operations, it was often necessary to examine reactor fuel assemblies and 

other test specimens to determine how they were performing. This involved handling and 

examining highly radioactive items, for which the Hot Lab operated in Building 4020 from 1959 

to 1990. The Hot Lab was a 16,000-ft2

 facility containing four large hot cells with remote 

manipulators and cranes, a mock-up area, an operating area, and decontamination areas. 

Construction was completed in 1959, and the facility was used until 1990. 
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2.4.3 Fabrication, Use, and Storage of Radioactive Sources 

Operations at SSFL required many instruments for detecting and measuring radioactivity, and 

these instruments were calibrated periodically using known quantities and types of radioactivity 

called sources, which were sealed containers that included small measured quantities of 

radioisotopes. Sources were also used for some forms of radiography, irradiation testing, and 

other applications. Sources were manufactured in the Hot Lab at SSFL, and used in various 

facilities at SSFL and elsewhere. Approximately 140,000 Ci of radioactive material (primarily 

Pm-147) were fabricated into sources at the Hot Lab. They were stored in secured locations and 

used under carefully controlled conditions. 

2.4.4 Preparation of Radioactive Material for Disposal 

The operation of nuclear reactors generates radioactive waste and other radioactive material that 

must be disposed of offsite. Other operations at the SSFL (fuel fabrication, reactor and fuel 

examination, etc.) also generated radioactive waste. Radioactive waste was prepared for disposal 

at the Radioactive Material Disposal Facility (RMDF) with support at the Interim Storage 

Facility (ISF) in Building 4654. 

2.4.5 Research on Reprocessing Used Reactor Fuel 

The Hot Cave in the Engineering Test Building supported licensing of nuclear fuel reprocessing. 

The used fuel assemblies from nuclear reactors contain unused fissionable material, fissionable 

transuranic products (mainly plutonium), and fission products. Rockwell developed a process to 

make a partial separation of used fuel, removing part of the fission products so that the material 

could be used again as reactor fuel. The experiments used up to one kilogram quantities of 

unirradiated uranium and thorium, and up to 100-g quantities of highly irradiated materials. 

2.4.6 Operation of Particle Accelerators 

Rockwell operated a Van de Graaff generator in Building 4030, bombarding tritium targets with 

deuterons to produce neutrons. A second Van de Graaff generator was operated for neutron 

activation of materials. 

2.4.7 Research Using Radioisotopes 

Some of the research at the SSFL required the use of special radioisotopes. For these tests, small 

quantities of specially prepared radioisotopes were brought to the SSFL, used in laboratories 

under carefully controlled conditions, and then either returned to the vendor or stored safely 

when reuse was required. 

2.4.8 Miscellaneous Operations 

Two of the facilities at SSFL, the Conservation Yard and the Sodium Disposal Facility (also 

referred to as sodium burn pit), were not intended for use with radioactive materials, but both 
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were inadvertently contaminated. The site profile states that both areas have been remediated 

and that no residual contamination has been detected. 

2.5 OTHER RELEVANT FACILITIES 

There are three other operations that are addressed in the site profile; the Downey facility, the 

Canoga Park facility, and the De Soto facility. Each of these facilities is described below. 

NIOSH considers these facilities as individual covered areas and does not include them in SEC 

Petition-00093. This issue will be further reviewed below. 

2.5.1 Downey 

The Downey facility, located in Downey, California, included AEC-funded activities performed 

in a small portion of a large building from 1947–1955. The AEC activities included mainly 

paper studies, R&D, and engineering studies. The R&D activities involved the use of a 2-MeV 

Van de Graaff generator, a small-scale radiochemical laboratory, a neutron counting room, and a 

construction area with a small 0.5 W teaching reactor. 

2.5.2 Canoga Park

Activities at the Canoga Park, California, facility occurred in the Vanowen Building from 

approximately 1954–1960. Activities that had been performed at the Downey facility were 

moved to the Canoga Park facility at the end of 1955. The primary activities performed at the 

Vanowen Building included design, development, and operation of small aqueous fuel reactors; 

fuel development; and radiochemistry, and beryllium machining is believed to have occurred. 

2.5.3 De Soto 

Radiological operations occurred at the De Soto facility from 1959 to the mid-1990s. Nuclear 

fuel material and other radioactive materials were used in Buildings 101 and 104 (referred to as 

001 and 004, respectively, prior to 1984) from 1959–1983. Building 104 was used at a muchreduced level until the mid-1990s. The nuclear operations conducted in these buildings included 

the Advanced Test Reactor (ATR) fuel fabrication and supporting activities; a Gamma Radiation 

Facility; and a mass spectroscopy (Helium Laboratory). 

2.6 RADIONUCLIDES OF CONCERN 

Workers at SSFL were engaged in many process operations and maintenance activities that had 

the potential for external and internal exposures to a host of radionuclides shown in Table 2-3 of 

TBD-2.
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3.0 OVERVIEW OF THE SEC PETITION-00093 AND NIOSH’S 

EVALUATION REPORT 

3.1 SEC PETITION-00093 

SEC Petition-00093 qualified on October 19, 2007. The petition requested that NIOSH consider 

the following class: “All employees who worked in all areas of SSFL-Area IV during the time 

period from 1955 to the present (which incorporated the post-1987 remediation period).” 

The petition provided information and affidavit statements in support of the petitioner’s belief 

that accurate dose reconstruction over time is impossible for the SSFL-Area IV workers in 

question. NIOSH deemed the following information and affidavit statements sufficient to 

qualify SEC-00093 for evaluation: 

As identified in Item F.1 of the SEC-00093 Form B and discussed in the July 30, 

2007 call, the petitioner discussed the sodium burn pit, the lack of internal 

monitoring data, and indicated that no records were kept. The petitioner also 

discussed a Tiger Team report indicating that the report detailed “inadequate air 

monitoring” and that “no internal monitoring was done.” 

The information and statements regarding the lack of pre-1959 internal monitoring data qualified 

the petition for further consideration by NIOSH, the Advisory Board on Radiation and Worker 

Health, and the Department of Health and Human Services. NIOSH determined that the time 

period from 1955 to 1965 should qualify for the purposes of the SEC petition evaluation, which 

includes the employment period in the original SEC-00093 petition. 

Additional or more specific information that was included in the petition, but not mentioned 

above, includes the following: 

• Faulty HEPA filters associated with the radiation storage buildings 

• Improper dosimeter badges 

• No record of radionuclides released during the partial meltdown of the SRE 

• Evidence of spills incompletely documented 

• Monitoring wells placed upgradient of the site 

• No soil or water sampling for tritium and a general lack of characterization of the site 

• Radiological survey showing tons of radioactive waste stored in boxes by 

Building 143—high levels of Cs-137 and Co-60 

A summary of the key elements defined in SEC-00093 include the following: 

• Issue #1: SRE Incident and Release of Core Gases. The basis for this issue being 

identified was a concern that monitoring records do not exist for this incident and that 

there was a release of core gases. 
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• Issue #2: Radiation Badges. This issue was based on a Tiger Team report indicating 

“inadequate radiation badges.” 

• Issue #3: Tritium Plumes. The petitioner was concerned that there may have been 

employee exposure to the tritium plumes through the consumption of drinking water. 

• Issue #4: Uranium Fires. The petitioner references many uranium fires and cites two 

incidents of a sodium explosion, and it appears to express concern that there was 

insufficient or no monitoring associated with these and similar incidents. 

• Issue #5: Air Monitoring. The petitioner was concerned that there was insufficient air 

monitoring, and NIOSH agrees with this prior to 1958. 

• Issue #6: The Sodium Burn Pit. The concern with the pit is that there was no monitoring 

of the pit and no records were maintained. 

 

3.2 SEC PETITION EVALUATION REPORT 

On February 6, 2008, NIOSH issued its SEC Petition Evaluation Report (ER) for SEC-00093. 

Section 1.0 of the ER states that this report evaluated “. . . the feasibility of reconstructing doses 

for all employees who worked in any area of Santa Susana Field Laboratory-Area IV during the 

time period from January 1, 1955 to December 31, 1965 . . .” [Emphasis added.] 

Under 42 CFR § 83.13(c)(1), the feasibility to reconstruct doses includes the following: 

. . . radiation doses [that] can be estimated with sufficient accuracy if NIOSH has 

established that it has access to sufficient information to estimate maximum 

radiation dose for every type of cancer for which radiation doses are 

reconstructed, that could have been incurred in plausible circumstances by any 

member of the class, or if NIOSH has established that it has access to sufficient 

information to estimate the radiation dose of members of the class more precisely 

than an estimate of the maximum radiation dose. [Emphasis added.] 

The ER responded to specific concerns and issues raised in the SEC Petition, as summarized 

above in Section 3.1, and concluded the following: 

. . . Based on its full research, NIOSH modified the petitioner-requested class to 

define a single class of employees for which NIOSH cannot estimate radiation 

doses with sufficient accuracy. The NIOSH-proposed class includes all 

employees of the Department of Energy (DOE), its predecessor agencies, and 

DOE contractors and subcontractors who were monitored while working in any 

area of Area IV of the Santa Susana Field Laboratory for a number of work days 

aggregating at least 250 work days from January 1, 1955 through December 31, 

1958, or in combination with work days within the parameters established for one 

or more other classes of employees in the SEC. [Emphasis added.] 
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The ER (p. 11) states the “NIOSH-proposed class includes all employees who were monitored or 

should have been monitored for internal radiological exposures while working in all areas of 

Area IV.” 

While the original petition requested that all employees be considered, NIOSH has restricted the 

employees included in the class to either those who were monitored or should have been 

monitored. This inconsistency needs to be addressed and further instruction should be provided 

as to who is covered. 

Furthermore, the actual proposed addition to the class restricts the addition to those employees 

who were actually monitored, rather than to the potentially broader class of those “who were 

monitored or should have been monitored.” NIOSH normally includes this broader definition in 

its SEC class additions, and it is unclear why NIOSH has proposed a narrower addition to the 

SEC class in this case. Specifically, while the practice of issuing badges to workers who entered 

controlled areas is documented for the most part, the issue of environmental doses and doses due 

to exposure to the sodium burn pit and other areas outside buildings needs to be taken into 

account in the class definition. 

Data and information employed by NIOSH in its evaluation are cited in Section 4.0 of the ER. 

The recommendation to approve the SEC status of this modified class was based on NIOSH’s 

conclusion that it could not estimate internal exposures with sufficient accuracy during the 

period from 1955 through 1958 (which is a period NIOSH has stated as having limited internal 

monitoring data). NIOSH states that this timeframe corresponds with the date after which an 

established bioassay program existed at SSFL and after which sufficient internal monitoring has 

been identified. 

Sections 5.0 and 6.0 of the ER provide summary descriptions of SSFL processes, SSFL 

monitoring practices, and available monitoring data. These data closely parallel information 

contained in the six TBDs that define the SSFL Site Profile and provide the technical basis for 

Section 7.0 of the ER. 

While Sections 7.1, 7.2, and 7.3 of the ER address the generic feasibility of internal and external 

dose reconstruction for SSFL workers, Section 7.4 addresses specific issues and concerns 

identified in the SEC petition, as summarized below. 

3.3 NIOSH’S RESPONSE TO MAJOR ISSUES RAISED IN SEC-00093 

3.3.1 Responses to Six Issues Identified in the SEC Petition and During Discussions with 

the Petitioner 

Response to Issue #1: SRE Incident and Release of Core Gases

NIOSH states that several documents were reviewed in the preparation of the ER (Lochbaum 

2006; Hart 1962) beyond what was reviewed as part of the site profile. It should be noted that 

there are at least two key documents (AI 1959 and 1961) that were prepared related to this 

incident that were not reviewed as part of the site profile or ER. 
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NIOSH concludes the following based on the reports reviewed: 

Although there is not a consensus on the exact amount of gaseous radioactive 

materials released to the environment following the incident, all pertinent 

scientific reports conclude that a significant amount of fission products were 

released into the primary sodium coolant, with a percentage of that inventory 

being released to the reactor’s cover system, and subsequently released into the 

atmosphere through reactor building vent systems and from gaseous storage 

tanks. The type and range of releases to the environment following the fuel 

damage run from less than 1 Ci of iodine-131 (Christian, 2005) to a conservative 

upper bound estimate of approximately 3000 Ci of iodine-131 (Lochbaum, 2006). 

Based on documented stack releases during the incident, AI concluded that since 

no iodine-131 was detected in cover gas, only about 28 Ci of the noble gases 

krypton-85 and xenon-133 were released from the stacks to the environment 

(Rutherford, 2005). 

Based on their review of claims in NOCTS, NIOSH concluded that personnel monitoring exists 

for members of the proposed class (both internal monitoring and external monitoring) during the 

timeframe of the SRE event. NIOSH also notes that some air monitoring measurements from the 

reactor area and stack monitoring also exist. 

Section 4.4.2 of this report presents our concerns related to the robustness of the internal 

monitoring of staff who may have worked in the SRE area. The potential that all staff may not 

have been properly monitored, combined with the seriousness of this event, suggests that it may 

be necessary to develop an exposure model for this incident or conclude that exposures cannot be 

properly evaluated. 

Response to Issue #2: Radiation Badges

NIOSH’s response to this issue is captured in the following statement: 

The Tiger Team Report cited by the petitioner does not state that the dosimeters 

were inadequate, but that they were not Department of Energy Laboratory 

Accreditation Program (DOELAP) accredited, and it specifically focuses on the 

D&D period versus the period being evaluated in this report. As it relates to the 

program being discussed in the Tiger Team report, it was common practice (and 

is noted as an option in Section 7 of the DOELAP Administrative Standard DOESTD-1111-98) for smaller programs to be exempted from DOELAP accreditation, 

contingent upon using a National Voluntary Laboratory Accreditation Program 

(NVLAP)-accredited commercial service (DOE, 2006). However, this is pertinent 

only beginning in 1986 when the DOELAP requirements were implemented. 

Therefore, as discussed previously, the period cited in the petition is after the 

period of the evaluation and does not impact this evaluation report. 
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We concur with NIOSH’s response, assuming the petitioner was not referring to some other 

concern than the adequacy of the badges. As noted, SC&A has not interviewed the petitioners 

for this preliminary report. 

Response to Issue #3: Tritium Plumes

NIOSH states that tritium has never been detected (>1,000 pCi/l) in any of the water supply 

wells. They further state that the primary supply wells (WS-5, WS-6, WS-12, and WS-13) were 

in Areas I, II, and III at SSFL, while one well (WS-7) was in Area IV and contributed a small 

percentage of the water supply. NIOSH has proposed to bound this exposure by assuming that 

workers consumed water from a shallow monitoring well that has had the highest tritium 

concentration. NIOSH points out that this well has never been a source of drinking water and is 

downgradient of the source of tritium (Building 4010). NIOSH also stated in the evaluation 

report that since 1991, the mean concentration of tritium in this well has been 2,940 pCi/l. 

Through some modeling and migration and decay assumptions, NIOSH concluded that workers 

could have consumed water with 30,000 pCi/l of tritium in the 1950s and 1960s. 

The above assumptions seem to be claimant favorable, except for the fact that more recent 

Boeing reports (Boeing 2003, 2004, 2005, and 2006) have documented tritium concentrations as 

high as 117,000 pCi/l in the monitoring wells (not the drinking water wells). Therefore, the 

above modeling should be reviewed in light of this more recent information. Concentrations for 

the period January 1, 1959, to December 31, 1965, need to be provided or inferred by a 

scientifically defensible method in order to estimate tritium dose for this period. 

Response to Issue #4: Uranium Fires 

NIOSH has concluded that there are data available for the various uranium fires that would allow 

dose estimates to be bounded and, based on our review of this information (Alexander 1967a, b, 

c, and d; Badger 1960 and 1961; Begley 1976; Klostermann 1961; Loba 1959, 1960, 1961a, 

1961b, 1962, and 1970; Oldfield 1961; Mooers 1959a and 1959b, 1960, and 1961 a, b and c; 

Owens 1978; Rudkin 1964 a, b and c; Stephenson 1961, 1962, 1963a, and 1963b; Weber 1963; 

Young 1960 and 1965), we concur with this conclusion. 

Response to Issue #5: Air Monitoring

NIOSH has concurred that there is a lack of area air monitoring for the period at SSFL-Area IV 

prior to 1958, which impacts the feasibility of estimating internal radiation doses with sufficient 

accuracy for the proposed worker class during that time period. After 1958, NIOSH will base its 

dose reconstructions primarily on bioassay data. 

Our discussion on the adequacy of the bioassay program prior to and after January 1, 1959, is 

presented below in Section 4.3 
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Response to Issue #6: The Sodium Burn Pit

NIOSH concludes that while there may have been radiological contamination at the Sodium 

Burn Pit, the following facts appear to support a low exposure and ability to bound dose 

estimates: 

• It appears contaminant levels were low (56 pCi/g maximum) related to radioactive 

contamination (principally cesium-137), and that contaminants were primarily in the 

lower pond. 

• Contamination was not identified in areas outside the ponds. 

• The Sodium Burn Pit was an outdoor area not continuously occupied, nor 

continuously used. 

• Because the pit was not an operations area, exposures for individuals that may have 

intermittently occupied the Sodium Burn Pit would have been lower than exposures 

for individuals who performed operations work with source materials that were 

delivered to this location. 

• Radiological exposure controls were in place (e.g., workers were required to maintain 

a safe distance from the pits, including lined and unlined pits/ponds), because of the 

violent reactions that could occur if sodium or potassium made contact with the 

water. 

• Bioassay results were available for affected and/or associated members of the 

proposed class, including worst-case exposure scenarios for the proposed worker 

class. 

This issue is discussed in Section 4.3. 

 

3.3.2 Other Issues 

Two other issues were identified by NIOSH in preparing the ER. The first issue dealt with the 

identification of workers with blank radiation exposure record sheets in their file (a sheet with no 

entries). The second issue dealt with the monitoring of firemen from other sites who got 

involved with fires or events at SSFL Area IV. 

Response to “Other” Issue #1

NIOSH discovered through an interview with a current Radiation Safety Officer at SSFL that all 

individuals were issued a blank record sheet in his/her file called a “blue card.” If an individual 

entered into a “controlled” area, they were required to have a film badge and any exposure was 

entered into their file. This practice was corroborated by NIOSH through random personnel 

record reviews and through other reviews (Boice et al. 2006a and Boice et al. 2006b). 

SC&A acknowledges this practice. 
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Response to “Other” Issue #2

NIOSH learned through interviews and reviews of dosimetry information that firemen typically 

wore film badges when working in areas with the potential for radiological exposures. One 

person interviewed mentioned that firemen wore badges at all times. However, one fireman did 

not have monitoring records in his file. The SEC Petition ER concludes the following: 

As previously discussed and stated, the availability of personnel records for 

monitored individuals supports NIOSH’s ability to reconstruct dose with 

sufficient accuracy for those proposed worker class members. Because the 

available data also includes a representation of the maximum potential exposures 

(a bounding exposure scenario) for the proposed worker class, NIOSH contends 

that this supports the ability to bound the associated dose for all members of the 

proposed worker class, including dose associated with the exposure scenarios 

presented by/for these firemen. 

SC&A is in agreement with the above statements. 

3.4 NIOSH’S CONCLUSIONS 

The ER concluded that, based on the limited pre-1959 internal monitoring data and associated 

program or source information, there is insufficient information to support establishing a 

bounding internal exposure scenario for the proposed worker class that worked in Area IV of the 

SSFL from 1955 through 1958. NIOSH has identified post-1958 radiological internal 

monitoring program data and the original monitoring data for individuals working in the highestexposure areas at Area IV, and therefore has concluded that doses can be bounded for the class 

evaluated for the period from 1959 through 1965. 
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4.0 PRELIMINARY FINDINGS ASSOCIATED WITH THE SEC 

PETITION 

 

4.1 OVERVIEW

This section of the SEC review process identifies discrete issues of concern that may adversely 

affect the ability to estimate SSFL worker exposures. Findings presented below are grouped as 

follows: 

• Section 4.1: Which areas (Area IV, Canoga Park, DeSoto, and Downey) should be 

considered in the SEC Petition and when did operations begin at Area IV 

• Section 4.2: Concern whether the internal monitoring program was sufficiently robust to 

estimate exposures before and after January 1, 1959 

 

• Section 4.3: Lack of information related to the facility “incidents” (SRE, Sodium Burn 

Pit, etc.) 

• Section 4.4: Lack of information on the environmental exposures (surrogate data, 

drinking water, etc.) 

• Section 4.5: Justification for assignment of external dose estimates is not provided, and 

there is no coworker model for external exposures (personnel records are also of concern) 

4.2 AREAS CONSIDERED IN THE SEC PETITION AND COVERAGE DATES 

4.2.1 Areas Addressed by Petition 

The petitioner submitted SEC Petition Form B to the NIOSH Office of Compensation Analysis 

and Support (OCAS) on June 22, 2007. The petition referenced the employer as Atomics 

International, and the following locations were noted as being relevant to the petition: 

• SSFL Buildings 059, 010, and 143 

• DeSoto 101 

• The Burn Pit 

• Santa Susana Field Laboratory 

However, the petitioner submitted a revised Form B on November 16, 2007, in which the 

specific SSFL buildings, DeSoto 101, and Burn Pit were removed from the list of locations. It 

appears the petitioner was asked to remove these other operations, since the petition should focus 

on the DOE facility or AWE facility (“covered” facility) at which the class worked as per 

42 CFR Part 83.9(c)(1)(i). 

Periodically, DOE publishes a list of facilities “covered” under EEOICPA. The latest 

publication of these sites was on August 23, 2004 [Federal Register/Vol. 69, No. 162/Monday, 

August 23, 2004 (pp. 51825-51831)]. There are two sites shown on this list that could apply to 
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the facilities covered in the site profile; Atomics International in Los Angeles County, which is 

shown as a BE DOE (Beryllium and DOE) facility, and Energy Technology Engineering Center 

(ETEC), Santa Susana, Area IV. There are no listings for the associated facilities operated by AI 

in Canoga Park, De Soto, or Downey. 

 

DOE has developed a database of the covered sites, which is accessible through the web site, 

http://www.hss.energy.gov/healthsafety/FWSP/Advocacy/faclist/findfacility.cfm. This database 

includes the following “covered” facilities: 

• Area IV of the Santa Susana Field Laboratory, which is shown as a DOE facility with 

the operator, North American Aviation (NAA). 

 

• Atomics International (shown as Beryllium Vendor and with the locations of Los 

Angeles and Ventura Counties) and includes, but is not necessarily limited to, the 

following locations; Area IV of the SSFL, portions of the Downey facility, the Vanowen 

Building at the Canoga facility, and the De Soto facility. The description notes that the 

AI Division of NAA is a statutory beryllium vendor under the EEOICPA, but that the 

company also worked with radioactive materials under contract with the Atomic Energy 

Commission (AEC) at numerous locations. (It should be noted, as stated above, that the 

AI site is shown as a BE DOE facility in the most recent Federal Register notice). 

• Canoga Avenue Facility, which is shown as a DOE facility with the operator, NAA. 

• De Soto Avenue Facility, which is shown as a DOE facility with the operator, NAA. 

• Downey Facility, which is shown as a DOE facility with the operator, NAA. 

 

Therefore, the list of covered sites in the DOE database is not consistent with the list of covered 

sites identified in the August 23, 2004, Federal Register. 

The nomenclature related to these facilities is further confused by the document titles of the six 

Technical Basis Documents (TBDs) identified below, where various facility names are used in 

the titles. 

• ORAUT-TKBS-0038-1, Technical Basis Document: Atomics International – 

Introduction, Rev. 01 (ORAUT 2006a) 

• ORAUT-TKBS-0038-2, Technical Basis Document: Energy Technology Engineering 

Center – Site Description, Rev. 00 (ORAUT 2006b) 

• ORAUT-TKBS-0038-3, Technical Basis Document: Atomics International – 

Occupational Medical Dose, Vol. 3, Rev. 00 (Atomics International TBD, 2006a, 

ORAUT 2006c) 

• ORAUT-TKBS-0038-4, Technical Basis Document: Area IV of the Santa Susana Field 

Laboratory, the Canoga Avenue Facility (Vanowen Building), the Downey Facility, and 

the De Soto Avenue Facility (sometimes referred to as the Energy Technology 
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Engineering Center [ETEC] or Atomics International) – Occupational Environmental 

Dose, Rev. 01 (ORAUT 2007) 

• ORAUT-TKBS-0038-5, Technical Basis Document: Energy Technology Engineering 

Center – Occupational Internal Dose, Rev. 00 (ORAUT 2006d) 

• ORAUT-TKBS-0038-6, Technical Basis Document: Atomics International – 

Occupational External Dosimetry, Rev. 01 (ORAUT 2006e) 

The TBDs focus on Area IV of the Santa Susana Field Laboratory, but also address the Canoga, 

Downey, and De Soto facilities. Dose reconstructions that are prepared for employees of AI 

include work conducted at all four of these facilities. It is noted above that the AI Division of 

NAA operated all four of these facilities, and their internal and external dose monitoring 

programs were the same for all the facilities. 

In addition, TBD-1 (p. 6) states, “The name Atomics International (AI) is used to represent all 

[all meaning the four facilities; Area IV, Downey, Canoga Avenue, and De Soto Avenue] of 

them unless more specific location information is warranted.” Therefore, when the petition 

refers to the facility as AI and it is shown as a covered DOE facility, it is easy to understand why 

there would be some confusion as to what locations are covered. 

Given the above, we believe NIOSH should re-assess the SEC Petition being only applicable to 

Area IV at SSFL. 

4.2.2 Coverage Dates 

The initial date of coverage is dependent on whether all four facilities are covered in the petition, 

or if just Area IV of SSFL is covered. 

If all four facilities are covered in the petition, the initial year of coverage should be 1947 or 

1948. TBD-2 (p. 9) states, “The SSFL was initially established by North American Aviation 

(NAA) in 1947 to meet the requirements for a field test laboratory to static-fire large rocket 

engines, but it also met the NAA’s need for a nuclear research facility.” The first year of 

coverage shown for the Downey facility in the DOE database is 1948, which is also noted as the 

first year of coverage in TBD-1 (ORAUT 2006a) (p. 6). TBD-2 (p. 27) discusses AEC-funded 

activities taking place at Downey between 1948 and 1955. 

If only Area IV of SSFL is covered in the petition, it is not clear whether the initial date of 

coverage is 1953, 1954, or 1955. The initial date of coverage for Area IV is shown as 1955 in 

42 CFR Part 83.9 and the DOE “covered facilities” database. However, there are conflicting 

references in the TBDs as to when operations began at Area IV. TBD-2 (p. 9) states, “Area IV 

was established at the SSFL in 1953 as a nuclear research and development facility.” Later in 

this same document on the same page, it is stated that “Nuclear R&D activities in Area IV 

increased rapidly from 1953 into the late 1960s, then declined.” Figure 2-3 in TBD-2 also shows 

that Area IV was established in 1953 to provide nuclear-related research. TBD-2 (p. 13) states, 

“Between 1954 and 1980, several nuclear reactors were built, tested, and operated in Area IV.” 
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Given the above, we believe NIOSH should re-assess the time periods of interest to the SEC 

Petition. 

4.3 ADEQUACY OF THE INTERNAL MONITORING PROGRAM 

4.3.1 Overview of Internal Monitoring Program 

SC&A evaluated internal monitoring data from 136 claimant files with dose reconstruction 

overviews. Our evaluation included identifying the number of workers that were monitored in a 

given year. The results of this evaluation are presented graphically below. The percentage of 

workers monitored and the total number monitored are shown in Figures 4.3.1-1 and 4.3.1-2 for 

the period of 1955–1995 and 1956–1966, respectively.1

 The conclusions we have drawn in 

evaluating this information are presented below. 

0%

5%

10%

15%

20%

25%

30%

35%

40%

1955 1960 1965 1970 1975 1980 1985 1990 1995

Year

% Workers Internally Monitored

Figure 4.3.1-1: Percentage of Workers Monitored, 1955–1995 

 1

 Figures showing the actual number of workers internally monitored were redacted given Privacy Act 

concerns. 
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0%

5%

10%

15%

20%

25%

30%

35%

40%

1956 1957 1958 1959 1960 1961 1962 1963 1964 1965 1966

Year

% Wokers Internally Monitored

Figure 4.3.1-2: Percentage of Workers Monitored Internally from 1956–1966 

4.3.1.1 1948–1958 

TBD-5 describes the internal dosimetry (bioassay) program at SSFL, with key excerpts presented 

below. As part of the ER, NIOSH located personal and area monitoring data in the NIOSH Site 

Research Database (SRDB) and in its NIOSH OCAS Claims Tracking System (NOCTS), which 

have been used to estimate doses to individual employees in the proposed class. NIOSH 

identified only limited amounts of internal personnel monitoring data for pre-1959 exposures, 

which is consistent with its finding that an SSFL routine bioassay program was not initiated until 

August 1958 (Kellehar 1966). 

 

The figures above indicate that there was essentially no internal monitoring taking place prior to 

1959, which is consistent with NIOSH’s findings in the SEC ER. 

4.3.1.2 1959–1965 

Entry into the bioassay program was based on job assignment, but monitoring procedures were 

generally invariable across occupations. However, firefighters appeared to be monitored more 

frequently than other workers. By the early 1960s, the bioassay program “normally” consisted of 

urinalysis for personnel whose work assignments involved “potential exposure to radioactive 

materials.” The frequency of sampling during the 1960s was slightly higher than in any other 

time period and varied from one to four per year, depending on the nature of the employee’s 

work, past exposure history, etc. Special bioassay sampling, consisting of more frequent urine 

testing, was in place very early (1960), but “only when gross internal contamination” was 

suspected (Hart 1979). 
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The figures above indicate that while the number of personnel that were internally monitored 

was increasing during this period, it was not until roughly 1963 when the percentage of workers 

monitored “somewhat” stabilized above 25%. Hence, there is some question whether the 

bioassay data collected and analyzed from 1959 through 1963 was sufficient for internal dose 

reconstruction or to develop a coworker model for internal dose reconstruction.

4.3.1.3 1966–2004 

In 1970, standards for bioassay sampling were published (Staszesky 1970). Work in areas where 

unencapsulated radioactive material was present required baseline and termination urine 

samples. A new baseline could be required for a change in job assignment. For new operations, 

a “pilot” bioassay program consisting of weekly urine samples could be required until a pattern 

was established. Regular work in these areas required a quarterly routine urine sample, but 

monthly samples could be required in a case of high exposure potential. Periodic fecal samples 

and in-vivo counts could also be required. Employees who periodically performed work in these 

areas were subject to semiannual urine samples. Personnel who frequently entered these areas, 

but did not perform hands-on work, such as project engineers, industrial engineers, etc., provided 

annual routine urine samples. 

SC&A has observed no cases where workers had a significant increase in the monitoring 

frequency after 1970. This is consistent with the data, which show that overall internal 

monitoring frequencies did not change over time (Figures 4.3.1.3-1 and 4.3.1.3-2). 

By the mid-1970s, the definition of who was included in the routine bioassay monitoring 

program had changed to “personnel whose work assignments potentially expose them to 

respirable-sized radioactive aerosols” (Hart 1979). By the late 1980s, the criterion was 

“personnel whose work assignments potentially expose them to radioactive aerosols” (Tuttle 

1989). Quarterly urine sampling was the norm through the 1980s (Hart 1979, 1980a, 1980b, and 

1980c; Eggleston 1983 and 1984; Tuttle 1985, 1986a, 1986b, 1986c, 1988a, 1988b, and 1989). 

In the mid-1970s, fecal sampling was used, but “only when gross internal contamination” was 

suspected (Hart 1979). Using the concept of a Maximum Possible Body Burden (MPBB), an 

excretion rate was determined by radionuclide that would indicate that one MPBB had been 

received. For several years prior to 1968, the policy was to restrict employees from work in 

potential airborne areas until their body burden was less than 25% of the MPBB (Alexander 

1968a). Starting in January 1968, ETEC imposed a restriction from work in areas with potential 

airborne exposure (or in some cases, from all radiation areas) if the bioassay results indicated the 

receipt of 50% or more of the MPBB. The restriction remained in place until two consecutive 

bioassay samples indicated that the remaining deposition was less than 25% of the MPBB 

(Staszesky 1970). 

The figures above indicate that the internal monitoring seems to have stabilized during this 

period and decreased in later years as operations ceased. 
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4.3.2 Monitoring for Internal Exposure of SSFL Workers was Incomplete and Poorly 

Documented for Most Years of Facility Operation 

In Section 5.2 (p. 10) of TBD-5, the evaluation of the SSFL internal monitoring program is 

prefaced with the following statement: 

Early 1960s AI documents describe all the elements of a comprehensive radiation 

safety program, including a laboratory with bioassay capability. ... [Emphasis 

added.] 

This statement, however, was tempered by numerous admissions in subsequent sections of the 

TBD. A sampling of statements suggesting deficiencies and data limitations include the 

following: 

 Page 13: 

Specific radionuclides could be determined “where required” (Lang 1960). 

Some detail has been found on early urinalysis methods. In addition to the inhouse laboratory capability, bioassay services were contracted to the following 

vendors: [Eight vendors are listed.] 

 

...Information on the periods during which ETEC used these laboratories was not 

found. [Emphasis added.] 

 In addition to the eight contract laboratories, ETEC had its own in-house laboratory, which 

analyzed urine for uranium content by fluorometric method. Exposure to uranium may have 

existed in various states of enrichment up to 93%. 

 Page 14: 

Due to its higher specific activity, EU activity could be determined by counting.... 

No specific information on sensitivities for the in-house laboratory was 

obtained.... [Emphasis added.] 

 Page 15 (Regarding thorium): 

No details of early thorium analyses were recovered... 

 Page 15 (Urinalyses for period 1967–1974): 

Partial documentation on bioassay methods from 1967 through 1974 was found. 

...These documents are believed to refer to services offered by UST [one of the 

eight contract laboratories]. [Emphasis added.] 
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 Page 17 (Urinalysis for period 1975–1988): 

The following analytical methods were taken from a series of annual reports.... 

The measurement “type” in parentheses appears in many personnel bioassay 

records. The detection limits should have improved over the years. However, a 

listing was not found. ... [Emphasis added.] 

 Page 18 (On the method for analyzing mixed fission products): 

Mixed fission products were precipitated from a basic oxalate media. ... Alkali 

metals such as 137Cs did not precipitate. In addition, volatile fission products 

such as 1-131 were lost. ... [Emphasis added.] 

 Page 19 (On in-vitro methods for individual radionuclides): 

Although fecal sampling was mentioned as both a routine and a special bioassay 

method in site documents, little detail has been found about the analytical 

methods used. ... [Emphasis added.] 

 Page 20 (On the use of whole-body counting for monitoring workers): 

... whole-body counting for fission or activation products was apparently not part 

of the routine bioassay program at ETEC. Between 1975 and 1988, only 25 

counts on 25 individuals were summarized in annual reports. ...All WBCs were 

reported positive for 137Cs. Ten counts were performed in 1977 and 15 were 

performed in 1979. [Emphasis added.] 

 Page 20/21 (On the use of chest counting): 

In 1967, the first chest (lung) counts for uranium using a medical system were 

performed at UCLA. The 186-keV gamma ray from the decay of 235U was used to 

quantify the amount of EU in the lung ... Calibration of this system was crude;... 

[Emphasis added.] 

Starting in 1968, Helgeson Nuclear Services provided lung counting services. ... 

The results were reported in milligrams of 235U ± 2 sigma... 

By 1977, two 5-in.-diameter, thin-window phoswich detectors were used, ... [to 

detect U-235] 

(SC&A notes that all chest measurements only quantified the amount or the 

activity of U-235. Without a firm understanding of the level of enrichment, the 

more important contribution of U-234 to total alpha activity cannot be 

determined.) 
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 Page 25 (On solubility type, fraction activity, and particle size per facility): 

In the absence of any measurements or studies, NIOSH guidance requires the use 

of default solubility classes and particle size values from the International 

Commission on Radiological Protection. ... With one exception, facility-specific 

solubility and particle size data for ETEC has not been found. Activity fractions 

were not available with the exception of those for limited fuel fabrication 

operations. ... Table 5-9 lists [recommended/default] this information (emphasis 

added). [Our review of Table 5-9 indicates that to date, (1) a solubility class has 

not been assigned to all radionuclides, (2) activity fractions are lacking for several 

facilities, (3) activity fractions are based on inappropriate data, and (4) activity 

fractions fail to identify select radionuclides (e.g., Na-24, radioiodines)]. 

4.3.3 Insufficient Correlation between Bioassay Data and Potential Exposures to Specific 

Radionuclides 

Two sources (NIOSH 2008, ORAUT 2006e) state that there is sufficient bioassay and other 

supporting data available for 1959 and beyond to establish an upper bound for uranium, mixed 

fission products (MFPs), Po-210, plutonium, SR-90, tritium, and thorium. Of the 37 internal 

monitoring records available to SC&A, 78% were monitored for uranium, 62% for mixed fission 

products, 0% for polonium, 24% for plutonium, 5% for strontium, 14% for tritium, and 0% for 

thorium. 

SC&A does not believe that these sources have clearly demonstrated a correlation between the 

bioassay data available and the potential exposures to specific radionuclides. They have not 

clearly defined for which workers each of the procedures were conducted, including gross alpha 

and gross beta. From a brief review of claimant files, it appears that monitoring was not routine 

for all workers handling radioactive material. Furthermore, detection limits for 1975–1988 are 

unavailable, and Table 5-9 in TBD-5 containing solubility type and fraction of activity is 

incomplete in many cases. Hence, it appears that there were significant limitations in the 

bioassay program in 1959, and perhaps for a few years beyond 1959. 

4.3.4 Missing Radionuclides in Bioassay Data 

The site profile indicates that bioassay data were available for gross alpha, gross beta, uranium, 

fission products, plutonium, thorium, Po-210, Sr-90, H-3, P-32, S-35, C-14, Pm-147, 

americium, and curium. Of the 37 internal monitoring records available to SC&A, the vast 

majority of monitoring was for uranium and mixed fission products; 78% and 62%, respectively. 

Fission products were monitored for 49% of the workers. Only a few monitoring records exist 

for the following radionuclides: 8% for cesium, 3% for beryllium, 3% for mercury, 3% for 

potassium, 0% for polonium, 24% for plutonium, 5% for strontium, 14% for tritium, and 0% for 

thorium. Potential exposure to radionuclides such as U-233 and U-234 could have occurred 

during these operations, but it may not be possible to determine this, given the limitations of 

uranium monitoring (see below). 
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4.3.5 No Coworker Model 

The site profile has not cited an internal coworker model as necessary, and the document does 

not address the use of a coworker model for those individuals potentially exposed but not 

monitored. A review of some of the claimant files indicates that some workers, who it appears 

should have been monitored, were not monitored, and those workers who were monitored were 

not monitored on a routine basis. In many cases, the dose reconstruction reports rely on 

guidance that has been developed for internal dose determinations based on other site 

information, rather than relying on site information. A coworker model would allow for more 

precise determinations of the doses the SSFL workers received. It is unclear whether there are 

data for coworker models for all the relevant radionuclides, especially for the early years 

through 1961/1962. 

4.3.6 Lack of Source Term Data 

TBD-5 states the following: 

Without bioassay or air sample data, the last resort is determination of airborne 

concentrations using source term evaluations (NIOSH 2002, p. 19). Data on the 

amount of dispersible material available does not appear to be available for 

ETEC. 

4.3.7 NIOSH’s Interpretation of Reported Values by the Contract Laboratory Nuclear 

Science and Engineering Corp. (NSEC) may not be Correct 

Section 5.3.1.2 (page 14) of TBD-5 attempts to clarify data reported by NSEC related to 

urinalyses for gross alpha, gross beta, and MFPs: 

Gross Alpha 

Shepard (1959) gave a minimum measurable concentration of 7.5 dpm/L for gross 

alpha counting. NSEC gave its minimum measurable concentration as 

0.2 cpm/mL (NSEC 1957). It is assumed that this is a typographical error and 

0.2 dpm/mL was intended. [Emphasis added.] 

Gross Beta 

Shepard (1959) gave a minimum measurable concentration of 75 dpm/L for gross 

beta counting. NSEC gave its minimum measurable concentration as 1.0 cpm/mL 

(NSEC 1957). It is assumed that this is a typographical error and 1.0 dpm/mL 

was intended. [Emphasis added.] 

Mixed Fission Products 

... for beta activity with an approximate minimum detectable amount (MDA) of 

60 dpm/sample [is assumed] (ORAU 2004, p. 27) ... NSEC gave its minimum 

measurable concentration as 2.0 cpm/mL (NSEC 1957). It is assumed that this is 

a typographical error and 2.0 dpm/mL was intended. [Emphasis added.] 
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The unsupported assumption that in all three cases a “typographical error” was, in fact, made 

may not be correct, given the discrepancies between the values reported by NSEC relative to 

those of Shepard: 

Shepard 1959 NSEC

Gross alpha 7.5 dpm/l 0.2 cpm/ml or 200 cpm/l

Gross beta 75 dpm/l 1 cpm/ml or 1,000 cpm/l

MFP 60 dpm/l 2 cpm/ml or 2,000 cpm/l

If, in fact, the NSEC data were correctly reported in the units of cpm/ml, and adjusted for 

yield(s) and counting efficiencies, MDA values (redefined in dpm/l) are likely to more than 

double. Such large differences are hard to explain and raise questions about the credibility of 

bioassay data provided by contract laboratories as a whole. Since this concern applies to the 

time period post-1958, it brings into question NIOSH’s ability to perform internal dose 

reconstructions post-1958. This issue needs further investigation. 

4.3.8 Potential Difficulties Associated with Uranium Bioassay Data 

Uranium at SSFL to which workers may have been exposed existed in various degrees of 

enrichment (i.e., 2% to 93%). Section 5.3 of TBD-5 discusses the two independent methods 

used to assess uranium in urine; the fluorometric method only identifies uranium concentrations 

in ug/ml, while the radiometric method assesses the gross alpha activity in dpm/l. Given the 

potentially wide range of specific activities of uranium defined by fluorometric urine data, all 

fluorometrically analyzed urine would also require a concurrent radiometric evaluation of that 

sample in order for these data to be useful to dose reconstruction. 

From information contained in Section 5.3 of the TBD, it is unclear whether urine samples were 

consistently analyzed by both fluorometric and radiometric methods. This is particularly evident 

from the internal monitoring records pre-1961, where there appears to be no distinction made 

between the types of urinalysis methods used. From 1961–1962, there was a transition period 

where uranium in general was monitored, as well as Uranium Radiometric (UR) and Uranium 

Fluoride (UF). After 1962, the type of urinalysis methods used was usually specified. 

It is not unreasonable to assume that for the early years, concern for the chemical toxicity of 

uranium may have limited urine bioassay to the fluorometric method. If this assumption is true, 

the absence of concurrent radiometric analysis of urine samples would severely limit the value of 

early fluorometric data. 

Thus, in the event that a fluorometric urine bioassay cannot be matched with a concurrent 

radiometric analysis, a claimant-favorable default value should be used that defines the 

enrichment level of uranium. NIOSH should further explore this issue, as it might apply to the 

post-1958 time period. 
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4.3.9 There are Unanswered Questions Regarding the Completeness and Quality of 

Personnel Exposure Records 

In Section 5.2 of the TBD, the following statements appear: 

 Page 10: 

...AI established health and safety files on each employee that contained radiation 

exposure records, injury records, and other “pertinent” data (Lang undated, 

1960). Today, personnel radiation exposure records are in the Radiation Safety 

Records Management System (RSRMS), which encompasses about 170 file 

cabinets. [Emphasis added.] 

 Page 12: 

The bioassay records in the individual files generally consist of: 

• Individual Personnel Keysort Cards (Figure B-2, Attachment B), which were 

used to track the type, frequency, and week of sample collection. ... The 

forms can be difficult to read due to the quality of the copies, and dose 

reconstructors should refer to the forms listed below for urine and fecal data. 

This form might be the only place in vivo [?] data are listed. [Emphasis 

added.] 

 Section 5.7 (page 23): 

The bioassay results from ETEC and its predecessor organizations are 

apparently not available in a computerized format. The units used to report the 

results are generally included in the hard-copy reports. ... [Emphasis added.] 

 Section 5.8 (page 23): 

No codes have been found [for excreta samples]. ... 

These statements suggest that (1) all personnel records currently exist in hardcopy form only, 

and (2) records may be of poor quality, difficult to interpret, and incomplete. 

SC&A concludes that the use of these records for dose reconstruction will require a 

comprehensive assessment regarding the quality and completeness of records contained in the 

170 file cabinets, and requires guidance to dose reconstructors for their interpretation/use. 

4.3.10 Use of SSFL Site Survey Data/Source Term cannot be Regarded as Useful Surrogate 

Data for Bioassay Data in Dose Reconstruction 

In the absence of an individual’s in-vitro/in-vivo bioassay data, the TBD provides the following 

information and guidance to dose reconstructors: 
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 Section 5.11 (page 26): 

If bioassay data are not adequate to evaluate an individual’s internal doses, dose 

reconstructors can use workplace monitoring data (NIOSH 2002). The following 

types of workplace data might be available for ETEC: breathing zone air 

samples, general area air samples, and surface contamination surveys. However, 

these data are not likely to be in individual exposure records. Data on 

respirator use are not likely to be available ... resuspension factors are not likely 

to be available. [Emphasis added.] 

 Section 5.12 (page 30): 

Without bioassay or air sample data, the last resort is determination of airborne 

concentrations using source term evaluations (NIOSH 2002, p. 19). Data on the 

amount of dispersible material available does not appear to be available for 

ETEC. [Emphasis added.] 

This “guidance” is deficient and places an unrealistic responsibility on the dose reconstructor 

because, at a minimum, the dose reconstructor will need to make judgments regarding (1) how to 

use general air sampling data when breathing zone data are not available or limited, and (2) what 

resuspension factors should be assigned. It is important that additional guidance be provided that 

ensures that claimant-favorable assumptions are applied in a consistent manner. 

4.4 LACK OF INFORMATION WITH CERTAIN FACILITY INCIDENTS 

4.4.1 Overview 

Table 2-6 in TBD-2 presents major site incidents that had significant potential for internal or 

external exposure to personnel. The two primary incidents of concern are the SRE and the 

Sodium Burn Pit. 

 

4.4.2 Sodium Reactor Experiment Coolant Failure 

TBD-2 contains a discussion of the SRE coolant failure in Section 2.2.1.1.2 and in Table 2-6. 

Section 2.2.1.1.2 provides very little information on the incident, and does not discuss any 

potential exposure information. Table 2-6 (30 pages after Section 2.2.1.1.2) provides more detail 

on the incident, but does not provide any worker exposure information. There is information in 

Table 2-6 about exposure being negligible for nearby residents (a maximum theoretical 

calculated dose of 0.06 rem to someone living in Susana Knolls, the nearest residential area at 

the time), but does not present exposure information for workers. 

Based on our review of claimant files, it does not appear that all workers who should have been 

monitored were monitored. NIOSH (Hughes 2008) provided SC&A with a spreadsheet 

containing information on claimants who worked at Area IV in 1959, and whether they were 

internally or externally monitored. The key finding from this spreadsheet (the spreadsheet 

cannot be provided given Privacy Act concerns) is that 42% of those claimants, who were 
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“radiation workers”, were not internally monitored while all radiation workers were externally 

monitored.2 If these radiation workers did indeed work in the SRE area, the lack of internal 

monitoring would be a concern, given the program was supposed to be fully functional in 1959. 

 

Atomics International prepared two reports on this incident; a report titled SRE Fuel Element 

Damage, An Interim Report (NAA-SR-4488)—November 15, 1959 (AI 1959), and a second 

report titled, SRE Fuel Element Damage, Final Report (NAA-SR-4488 (suppl)—1961 (AI 1961). 

The interim report (AI 1959) contains important information related to this incident that is not in 

the site profile. Examples of this information include the following: 

• During this occasion, specifically, in October 1958, the maximum radiation 

levels in the general area of the moderator coolant pump were reported to be 

about 50 Mr/hr (October 14). Below shield blocks 1 and 2, the radiation level 

was about 21 mr/hr (on October 11). (p. IV-C-9) 

• Radiation levels measured on April 18, 1959 varied from 50 to 420 mr/hr. 

...Additional measurements made 5 days later (a total of 17 days after 

shutdown) indicated no significant decay. (p. IV-C-10) [Table IV-C-6 

includes radiation levels in the Gamma Facility on various dates in August, 

September, and October of 1959. The measured radiation levels peaked on 

August 12 (2.9 r/hr) and decreased to 0.7 r/hr on October 5.] 

• Cold trapping was started during run 14. However, radiation measurements 

could not start until August 8 (due to the radiation hazard from the high 

radiation levels of Na24), at which time the dose rate, extrapolated to near the 

surface, was about 70 r/hr. It is possible that initial cold-trap dose rates, had 

they been measured, would have yielded significantly higher values. (p. IV-C12) [The radiation rates at the cold trap, shown in Table IV-C-7, range from 

63 r/hr on August 8, 1959, to 50 r/hr, with a peak of 81 r/hr on August 13.] 

• Following the termination of run 14, the fuel handling cask was used to 

inspect the fuel elements in the reactor. ... Operations directed towards 

removal of these slugs resulted in occasional radiation levels as high as 

1000 r/hr at 1 ft. from the slugs. However, the maximum total exposure 

received by operations personnel during these cask operations did not exceed 

1 rem in a single week. (p. IV-C-22). [The basis for this last statement was 

not provided in the report, and the number of personnel exposed was also not 

presented.] 

 

The final report (AI 1961) also contains information related to potential worker exposure. 

Examples of this information include the following: 

• This report discusses the distribution and management of the fission products 

during the recovery operations. During the recovery effort the objectives 

2

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were: (1) To limit personnel exposure to an average dosage rate of 

1.25 rem/quarter (5 rem/yr). (p. III-19) 

• Throughout the recovery effort the radiation exposure to each individual was 

limited to less than 5 rem/yr. It was occasionally necessary to permit the 

weekly exposure for some key individuals to reach 600 mrem per week, in 

which case the individual was not exposed to radiation during the following 

week. Such exposures required a special permit, and only 30 permits were 

issued. For the 150 persons directly involved in the work, the average 

exposure was 2 rem/yr.

This information is pertinent to reconstructing both external and internal exposures associated 

with this incident. 

4.4.3 The Sodium Burn Pit 

The sodium burn pit was built to clean nonradioactive metallic sodium and NaK from various 

scrap test components. Contamination was identified in the sodium burn pit in 1978, at which 

time monitoring of the area began and continued until 1983. Prior to 1978, no radioactivity was 

expected in this area, thus raising questions whether individuals involved in these activities 

would be considered for internal monitoring. It is unclear when the radioactive contamination 

was introduced to the burn pit or how far back in time the potential exposure to radiation exists. 

Given the violent nature of the operation, it would be expected that this operation would have 

generated an airborne hazard. Since the sodium burn pit was not expected to result in radiation 

exposures, there was no routine monitoring. But this does not mean that routine exposure can be 

ruled out. Additional information should be gathered on this site to demonstrate that 

unmonitored workers are not likely to have received sufficient dose to be of concern, or a model 

should be developed to bound the exposures that could have been received. 

The TBD also does not consider exposure to contaminated soil that has resulted from spills and 

other incidental releases. For example, a review by a U.S. Environmental Protection Agency 

(EPA) official in 1989 (Dempsey 1989) identified Building 064, the Special Nuclear Materials 

Storage Area, that had been contaminated as a result of a spill. This EPA official also had 

concerns about the validity of some, if not all, of their environmental data: 

In the Rocketdyne procedure, soils are heated in a muffle furnace for 8 hours at 

500oC. Several problems were identified: first, this temperature is sufficient to 

volatilize most man-made radionuclides of concern, including cesium-137 and 

strontium-90. Second, from the Rocketdyne procedure, soil is sieved through a 

coors crucible to obtain uniform particle size.... This procedure is a screening 

method at best and is not an accurate quantitative procedure. 

TBD-5 does not specifically address radiological incidents that may have resulted in internal 

exposures to workers’ that were unmonitored. 
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Based on information provided in TBD-5, it is unclear (1) whether workers exposed during these 

early incidents were adequately monitored for internal exposure, and (2) if workers were, in fact, 

monitored, whether their exposure records exist. A particular concern involves the potential 

exposures associated with the SRE. Worker exposures may have included a complex mixture of 

highly enriched uranium, actinides, MFPs, and various activation products, including large 

amounts of Na-24. 

4.5 LACK OF INFORMATION ON THE ENVIRONMENTAL EXPOSURES 

4.5.1 Overview 

Occupational environmental dose refers to radiation exposures received by workers while onsite 

but outside the SSFL facilities from facility discharges to the atmosphere, ambient external 

radiation originating in the facilities, and inadvertent ingestion of site-generated radionuclides. 

The environmental monitoring program was established at Area IV in May 1954, before 

construction of the first radiological facility, with emphasis on soil, vegetation, and water 

sampling. 

TBD-4 provides guidance and data for assigning occupational environmental doses for Area IV, 

and the Downey, Canoga, and De Soto sites that make up SSFL, for all years, starting with 1954 

through 1999. Due to insufficient environmental monitoring/data limitations, the TBD relied on 

the following data and applied the following assumptions in order to provide the dose 

reconstructor the means to estimate inhalation intakes that are radionuclide-specific and for all 

years of facility operations. 

Average annual gross alpha/gross beta concentrations in facility stack emissions 

were the basis for estimating potential worker environmental inhalation intakes. 

Most of the available SSFL stack emission data include annual average gross 

alpha and gross beta concentrations at the stack point of release. Years with 

data vary by facility, but gross alpha/gross beta concentration information is 

available for most years between 1971 and 1999. ... 

...Identification of specific radionuclides released from various facilities in stack 

emissions are available ... from 1988 to 1999 and were used to characterize 

radionuclide emissions for all years. ... 

...In years where data were not available, stack concentrations were assumed to 

be the average yearly gross alpha and gross beta concentrations in stack effluents 

from years 1971 to 1999, for which data were available. 

...Furthermore, the average percentage that each identified radionuclide 

contributed to the gross alpha or gross beta concentration determined from 1988 

to 1999 data was applied to each of these years to make radionuclide-specific 

stack concentration estimates.... 
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NOTICE:

...Because the stack effluent concentrations were at the point of release, a further 

reduction factor of 0.01 was taken to account for the lessened overall intake due 

to contribution from multiple, widely spaced facilities; atmospheric dispersion of 

stack effluent over the course of a year’s exposure; and building wake effects.

[Emphasis added.] 

4.5.2 Surrogate Use of the Time-integrated Average Yearly Gross Alpha/Gross Beta 

Stack Emission Corresponding to Years 1971 to 1999 

The surrogate use of the time-integrated average yearly gross alpha/gross beta stack emissions 

corresponding to years 1971 to 1999 (when stack measurements were taken) is likely to 

underestimate stack emissions for years 1954 through 1970. SC&A’s conclusion is supported by 

the steady reduction in facility operations over time, as illustrated in Figure 4.5.2-1. For 

example, nuclear reactor programs were essentially phased out in the early 1970s. 

Sections 4.6.4, 4.6.5, and 4.6.6 of TBD-4 acknowledge the lack of external dose rate monitoring 

data prior to 1974, and provide unsupported/unreferenced assumptions that were used to derive 

annual external dose estimates for a restricted number of facilities, as given in Table 4-4 of the 

TBD. 

4.5.3 Exposure to Onsite Water Supply Wells 

The SEC ER states that onsite water supply wells were the primary water source from 1949 to 

1964, which differs from Section 4.7 of TBD-4, which states, “Potable water is not a source of 

occupational radioactive material at SSFL, because the SSFL facilities used either bottled water 

from an off-site vendor (Moore et al. 1962) or the city water supply.” In addition, other 

references (Winzer 1980 and 1981; Curphey 1983) indicate that well water was a source of 

drinking water into the 1980s. 

Although the ER stated that the drinking water supply wells did not have elevated levels of 

tritium (>1,000 pCi/l) (concern with tritium given current tritium plume on site), the ER has tried 

to bound any contamination that may have existed onsite by assuming the onsite supply wells 

were contaminated with tritium at a concentration of 30,000 pCi/l. More recent Boeing reports 

(Boeing 2003, 2004, 2005, and 2006) have documented tritium concentrations as high as 

117,000 pCi/L in the monitoring wells (not the drinking water wells). Therefore, this issue 

should be re-evaluated to determine if this exposure route can be properly assessed in earlier 

years, and if so, how to assess this exposure. 
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Figure 4.5.3-1: Summary of Area IV Activities 

NOTICE

wever, th
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4.6 ASSIGNMENT OF EXTERNAL DOSE 

TBD-6 contains documentation to assist in the evaluation of occupational external doses from 

processes that occurred at SSFL. An objective of this document is to provide supporting 

technical data using claimant-favorable assumptions to evaluate occupational external doses that 

can be reasonably associated with worker radiation exposures. This document addresses the 

evaluation of unmonitored and monitored worker exposure, missed dose, and the bias and 

uncertainty associated with the monitoring of external dose. 

4.6.1 No Coworker Model 

TBD-6 has not cited an external coworker model approach as necessary, and the document does 

not address the use of a coworker model for those individuals potentially exposed but not 

monitored. It appears that individuals may have been unknowingly exposed. An October 22, 

1962, memorandum from F.H. Badger3 to the Health and Safety File regarding “Health and 

Safety Observations at RMDF” states that, “Routine smear surveys have repeatedly revealed 

significant contamination or radiation dose rates in areas usually thought to be free of radioactive 

material.” One of the examples provided was a 4 Rad/hr capsule lying in an area thought to be 

uncontaminated. 

In addition, in an April 1991 Tiger Team Assessment, the DOE noted several issues with the 

external radiation dosimetry program. The report, as an example, noted that “In 1989 and 1990, 

extremity doses were not added in to exposure records or reported to the Radiation Exposure 

Information Reporting System (REIRS).” 

Based on the poor track record of badging employees appropriately, the conclusion of no 

monitoring being equal to no work in a radiation area does not seem to be justified. The dose 

reconstructor should be able to look at a job title and determine potential exposure, and then be 

linked to a coworker model. This issue is also relevant to how NIOSH has defined the SEC from 

1955 to 1958, which at present does not include employees who “should have been monitored,” 

but is limited to employees who were monitored. This SC&A conclusion indicates a need for 

NIOSH to revisit the limitation. 

4.6.2 Workers Were Unlikely to Have Been Monitored for Thermal Neutrons 

As stated in Section 6.2 of TBD-6, “...Both fast and thermal neutrons were measured and 

recorded as whole-body (WB) dose in rem” (emphasis added). This statement is contradicted in 

Section 6.4, where it states, “...It is assumed that the dose recorded was the result of fast neutron 

exposure” (emphasis added). 

The second statement is likely to be correct, since the common practice at DOE facilities was to 

assess NTA film for tracks produced by proton recoil. It is unlikely that NTA dosimeters were 

modified and calibrated for tract analysis of thermal neutrons. [Tracks in emulsions exposed to 

3

 F.H. Badger was employed by Atomics International. His title was Analyst, Health Physics, Senior 

Health and Safety Operations. 
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thermal neutrons may be produced by nitrogen in the gelatin that captures a thermal neutron and 

releases a 0.58 MeV proton–N14(n, p) C14.] 

In the absence of empirical data involving neutron spectra for reactors and Pu fuel storage 

facilities, the lack of dosimeter calibration methods, and the relative insensitivity of NTA film to 

neutrons with less than 500 keV (or as much as 1 MeV), there remains an undefined level of 

uncertainty for recorded neutron doses. Therefore, the use of Y-12 data as surrogate values may 

not be appropriate. 

TBD-6 assumes that the NTA film effectively measured all neutron exposure received at AI, and 

does not consider correction factors for the insensitivity of NTA to neutrons at energies below 

500 keV. Actual neutron energy spectrum data are limited to a few facilities (i.e., SRE). There 

is no discussion of neutron-to-photon ratios in the site profile; however, it is mentioned as an 

option for calculating thermal neutron exposure in the ER report. 

4.6.3 Dosimeter Response to Low-Energy Photons 

TBD-6 does not discuss issues associated with the response of dosimeters to low-energy photons. 

There are statements to the effect that the dosimeter was similar in design to the Hanford 

dosimeter. The Hanford dosimeter applied a correction factor for exposure of plutonium facility 

workers to compensate for badge shortcomings. The ER indicates there are source term data 

available to bound low-energy photon dose; however, no specific information on source term is 

provided. Furthermore, there is no consideration for dose from skin contamination incidents. 
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5.0 CONCLUSIONS 

After a thorough review of SEC Petition-00093 and the large number of relevant/support 

documents, it is still unclear as to which of the four facilities should be covered in the petition 

and the initial year of coverage, which is dependent on whether all four facilities are covered in 

the petition or if it is just Area IV of SSFL that is covered. The list of covered sites in the DOE 

database is not consistent with the list of covered sites identified in the August 23, 2004, Federal 

Register. If only Area IV of SSFL is covered in the petition, it is not clear whether the initial 

date of coverage is 1953, 1954, or 1955. If it is determined that all four facilities will be 

included in the SEC Petition, the initial coverage date will need to be changed to 1947 or 1948. 

NIOSH and DOE should make this decision, based on further research. 

SC&A agrees with the findings in the SEC ER that the internal monitoring program was 

insufficiently robust to estimate exposures before January 1, 1959. Only limited amounts of 

internal personnel monitoring data for pre-1959 exposures were identified, which is consistent 

with NIOSH’s findings that an SSFL routine bioassay program was not initiated until August 

1958 (Kellehar 1966). Our review of the percentage of workers being monitored for internal 

exposure indicates that the percentage continued to increase after 1959, until it plateaued several 

years later. A determination needs to be made by NIOSH on whether the percentage of workers 

who were monitored in the several years after 1958 indicates a fully functioning internal 

monitoring program. 

In addition, monitoring for internal exposure of SSFL workers was incomplete and poorly 

documented for most years of facility operation. These deficiencies and data limitations are 

stated throughout sections of the TBD. For 1959 and beyond, SC&A does not believe that 

NIOSH has clearly demonstrated a correlation between the bioassay data available and the 

potential exposures to specific radionuclides. 

There is a lack of information concerning two major facility incidents that had significant 

potential for internal or external exposure to personnel; the SRE coolant failure and the sodium 

burn pit. TBD-5 does not specifically address radiological incidents that may have resulted in 

internal exposures to workers’ that were unmonitored and, based on our review of claimant files, 

it does not appear that all workers who should have been badged were indeed badged. As noted, 

the poor track record of badging employees appropriately indicates a need for NIOSH to revisit 

the limitation for its proposed addition to the SEC class only to employees who were monitored. 

The information in TBD-5 is unclear as to whether workers exposed during these early incidents 

were adequately monitored for internal exposure and, if workers were monitored, whether their 

exposure records exist. 

SC&A questions the data and unsupported/unreferenced assumptions that were applied in order 

to reconstruct occupational environmental doses, due to insufficient environmental monitoring 

and data limitations. This includes inhalation intake estimates that are radionuclide-specific and 

the surrogate use of the time-integrated average yearly gross alpha/gross beta stack emissions 

corresponding to years 1971 thru 1999, which likely underestimate stack emissions for years 

1954 through 1970. There are also conflicting reports concerning the exposure to onsite water 
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supply wells. This issue should be re-evaluated to determine if this exposure route can be 

properly assessed in earlier years and, if so, how to assess this exposure. 

In the absence of empirical data involving neutron spectra for reactors and Pu fuel storage 

facilities, the lack of dosimeter calibration methods, and the relative insensitivity of NTA film to 

neutrons with less than 500 keV, there remains an undefined level of uncertainty for recorded 

neutron doses. Therefore, the use of Y-12 data as surrogate values may not be appropriate. 

There is no discussion of neutron-to-photon ratios in the site profile; however, it is mentioned as 

an option for calculating thermal neutron exposure in the ER report. The ER indicates there are 

source term data available to bound low-energy photon dose; however, no specific information 

on source terms is provided. Furthermore, there is no consideration for dose from skin 

contamination incidents. Additional work is needed in this area to either better define the likely 

neutron dose or provide a plausible bound for the dose. 
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NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

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Rudkin, R.D., 1964b. Incident Report from R.D. Rudkin (Health and Safety, Atomics 

International) to R.F. Wilmes, Subject: Uranium Fire Incident A-0407: Building 1, Room 1126 

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Effective Date: 

January 30, 2009 

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NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

Rudkin, R.D., 1964c. Incident Report from R.D. Rudkin (Health and Safety, Atomics 

International) to R.F. Wilmes, Subject: Uranium Fire Incident A-0436: HQ 1, Room 1131 and 

1114 – SSFL, March 12, 1964. 

Rutherford, 2005, Sodium Reactor Experiment (SRE) Original Accident Release Data, 

memorandum to E. Crawford; Phil Rutherford; January 21, 2005; SRDB Ref ID: 19161 and 

19162 

Santa Susana Field Laboratory SEC Petition. 2007. SEC Petition on SSFL to the Office of 

Compensation and Analysis, SEC-00093 

SC&A. 2005a. Second Supplemental Review of the Mallinckrodt Site Profile Revision 1 

(Effective Date March 10, 2005), Second Supplement to SCA-TR-TASK1-0002, July 1, 2005. 

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(Effective Date March 10, 2005), Third Supplement to SCA-TR-TASK1-0002. August 16, 2005. 

SC&A. 2006a. Board Procedures for Review of Special Exposure Cohort Petitions and Petition 

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SC&A. 2006b. Report to the Working Group on Special Exposure Cohort Petition Review, 

Draft. January 16, 2006. 

Schafer, G.K., 1961. Radiological Safety Incident Report from G.K. Schafer (Health and Safety 

Department, Atomics International) to J.C. Lang, Subject: Uranium Fire Incident A-0456: 

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SEC Petition Evaluation Report, Petition SEC-00093, NIOSH, Report Submittal Date—February 

6, 2008. 

Shepard, J.L., 1959. “Quotation on Urinalysis,” letter to Atomics International, U.S. Nuclear, 

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Special Exposure Cohort Petition—Form B: June 22, 2007; OSA Ref ID: 103252. 

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for radiation workers), Atomics International, Canoga Park, California, August 12, 1970. 

Stephenson, R.M., 1961. Radiological Safety Incident Report from R.M. Stephenson (Health 

and Safety Department, Atomics International) to J.C. Lang, Subject: Uranium Fire Incident A0357: Fed Area, Room 1026 – SSFL, November 27, 1961. 

Stephenson, R.M., 1962. Radiological Safety Incident Report from R.M. Stephenson (Health 

and Safety Department, Atomics International) to R.F. Wilmes, Subject: Uranium Fire Incident 

A-0355: FEPF, Room 1153 – SSFL, March 26, 1962. 
Effective Date: 

January 30, 2009 

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

Stephenson, R.M., 1963a. Radiological Safety Incident Report from R.M. Stephenson (Health 

and Safety Department, Atomics International) to R.F. Wilmes, Subject: Uranium Fire Incident 

A-0541: Room 1059 – SSFL, March 11, 1963. 

Stephenson, R.M., 1963b. Radiological Safety Incident Report from R.M. Stephenson (Health 

and Safety Department, Atomics International) to R.F. Wilmes, Subject: Uranium Fire Incident 

A-0498: Room 1252, SNAP – SSFL, August 28, 1963. 

Tuttle, R.J., 1985. Annual Review of Radiological Controls—1982, Rockwell International, 

Energy Systems Group, Canoga Park, California. January 29, 1985. 

Tuttle, R.J., 1986a. Annual Review of Radiological Controls—1982, Rockwell International, 

Rocketdyne Division, Canoga Park, California. February 12, 1986. 

Tuttle, R.J., 1986b. Annual Review of Radiological Controls—1984, Rockwell International, 

Rocketdyne Division, Canoga Park, California. March 18, 1986. 

Tuttle, R.J., 1986c. Annual Review of Radiological Controls—1985, Rockwell International, 

Rocketdyne Division, Canoga Park, California. July 11, 1986. 

Tuttle, R.J., 1988a. Annual Review of Radiological Controls—1986, Rockwell International, 

Rocketdyne Division, Canoga Park, California. July 5, 1988. 

Tuttle, R.J., 1988b. Annual Review of Radiological Controls—1986, Rockwell International, 

Rocketdyne Division, Canoga Park, California. July 28, 1988. 

Tuttle, R.J., 1989. Annual Review of Radiological Controls—1988, Rockwell International, 

Rocketdyne Division, Canoga Park, California. May 12, 1989. 

Weber, R.K., 1963. Radiological Safety Incident Report from R.K. Weber (Health and Safety 

Department, Atomics International) to R.F. Wilmes, Subject: Uranium Fire Incident A-0360: 

Building 004, Room 424-68 – SSFL, December 20, 1963. 

Winzer, C.E., 1980. Internal Letter from C.E. Winzer (Atomics International) to “Those Listed,” 

Subject: Drinking Water Supply—SSFL, August 14, 1980. 

Winzer, C.E., 1981. Letter from C.E. Winzer (Atomics International) to Mr. John Curphey, 

District Sanitary Engineer for the State of California, Department of Health Services, Sanitary 

Engineering Section, 3704 State Street, Suite 212, Santa Barbara, California 93105, January 9, 

1981. 

Young, L.N., 1965. Incident Report from L.N. Young (Health and Safety, Atomics 

International) to R.M. Hill, Subject: Uranium Fire Incident A-0489: Building 022, AMP 

Evaporator – SSFL, May 18, 1965. 
Effective Date: 

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NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

Young, W.A., 1960. Internal Letter from W.A. Young (Solid Waste Chemistry Unit, Atomics 

International) to R.H. Beight, Subject: Uranium Fire Incident A-0006: Explosion in Room 426-

13 – SSFL, July 27, 1960. 
Effective Date: 

January 30, 2009 

Revision No. 

0 

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NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

APPENDIX A: PETITIONER/WORKER INTERVIEW 

To be provided at a later date.