ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

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Document Title:

Site Profile for General Atomics

Document Number: ORAUT-TKBS-0045

Revision: 00

Effective Date: 09/26/2008

Type of Document TBD

Supersedes: None

Subject Expert(s): Shelby L. Gubin, Kenneth Fleming, Robert Hysong, Cindy Bloom, Robert Vogel, and 

Hobert Jones

Site Expert(s): N/A

Approval: Signature on File Approval Date: 09/02/2008

Bernard M. Olsen, Document Owner

Concurrence: Signature on File Concurrence: 09/02/2008

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 09/02/2008

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 09/09/2008

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 09/26/2008

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 2 of 143

PUBLICATION RECORD 

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

09/26/2008 00 Approved new document to establish the technical basis for the 

development of a radiation exposure matrix for General Atomics. 

Incorporates formal internal and NIOSH review comments. Modified 

document to include final SEC Evaluation Report language. Section 

5.0, Estimating Exposures From Residual Radioactivity, expanded to 

address dose reconstruction methodologies for exposures received 

during the residual period. Section 6.0, Estimating Exposure From 

DOE Remediation, expanded to address dose reconstruction 

methodologies for exposures received during the remediation period. 

Includes Attributions and Annotations. Training required: As 

determined by the Task Manager. Initiated by Bernard M. Olsen.
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 3 of 143

TABLE OF CONTENTS 

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 8

1.0 Purpose and Scope................................................................................................................ 11

2.0 General Site Description and Operational History .................................................................. 13

2.1 Site Description .......................................................................................................... 13

2.1.1 Facilities .......................................................................................................... 14

2.2 Process Descriptions .................................................................................................. 23

2.2.1 Fuel Fabrication .............................................................................................. 25

2.2.2 Uranium Purification ........................................................................................ 25

2.2.3 LINAC Facility ................................................................................................. 28

2.2.4 Critical Assembly Facilities .............................................................................. 32

2.2.5 Reactor Facilities ............................................................................................ 33

2.2.6 Science Laboratory ......................................................................................... 34

2.3 Source Term ............................................................................................................... 35

2.4 Safety ......................................................................................................................... 42

2.5 Incidents ..................................................................................................................... 46

2.6 Physical Examinations – X-Rays ................................................................................ 46

2.7 Summary Operational Period Assumptions, Workdays, Work Hours, 

Work Categories ......................................................................................................... 47

2.8 Residual Contamination and Remediation Periods ..................................................... 47

3.0 Estimating Internal Exposure ................................................................................................. 47

3.1 In Vitro Data ............................................................................................................... 49

3.2 In Vivo Data ................................................................................................................ 50

3.3 Radioelement-Specific Information ............................................................................. 51

3.3.1 Uranium .......................................................................................................... 51

3.3.1.1 General ............................................................................................ 51

3.3.1.2 Uranium In-Vitro Analyses and Data ................................................. 53

3.3.1.3 Uranium In-Vivo Analyses and Data ................................................. 55

3.3.2 Thorium........................................................................................................... 55

3.3.3 Plutonium ........................................................................................................ 56

3.3.4 Strontium ........................................................................................................ 56

3.3.5 Tritium ............................................................................................................. 58

3.3.6 Mixed Fission and Activation Products ............................................................ 58

3.3.6.1 General ............................................................................................ 58

3.3.6.2 Applicability of ORAUT-OTIB-0054 to Experimental 

High-Temperature Gas-Cooled and ReducedEnrichment Research and Test Reactor Fuels ................................. 59

3.3.6.3 Internal Exposure Assessment of Mixed Fission and 

Activation Products via ORAUT-OTIB-0054 Guidance...................... 60

3.3.7 Thoron ............................................................................................................ 61

3.4 Other Radionuclides ................................................................................................... 61

3.5 Internal Dosimetry Summary ...................................................................................... 62

4.0 Estimating External Exposure ................................................................................................ 62

4.1 Dose Record Format .................................................................................................. 65

4.2 Film Badge Correction Factors ................................................................................... 67

4.3 Occupationally Required Medical X-Rays ................................................................... 67
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4.4 Miscellaneous Information on External Exposure ....................................................... 68

4.5 External Dosimetry Summary ..................................................................................... 68

5.0 Estimation of Exposure from Residual Radioactivity ............................................................... 68

5.1 Introduction ................................................................................................................. 68

5.2 Estimation of Internal Exposure from Residual Radioactivity ...................................... 69

5.2.1 Introduction ..................................................................................................... 69

5.2.2 Thorium........................................................................................................... 70

5.2.3 Uranium .......................................................................................................... 71

5.2.4 Plutonium ........................................................................................................ 71

5.2.5 Mixed Fission and Activation Products ............................................................ 75

5.2.6 Tritium ............................................................................................................. 75

5.2.7 Strontium ........................................................................................................ 76

5.2.8 Other Radionuclides ....................................................................................... 81

5.2.9 High-Temperature Gas-Cooled and Reduced-Enrichment 

Research and Test Reactor Irradiated Fuel Materials 

Information ...................................................................................................... 81

5.2.10 Ingestion ......................................................................................................... 81

5.3 Estimation of External Exposure from Residual Radioactivity ..................................... 81

5.4 Summary .................................................................................................................... 82

6.0 Estimation of Exposure from DOE Remediation ..................................................................... 82

6.1 Source Term ............................................................................................................... 84

6.2 Safety ......................................................................................................................... 84

6.3 Estimation of Internal Exposure from Remediation ..................................................... 85

6.3.1 Ingestion ......................................................................................................... 87

6.4 Estimation of External Exposure from Remediation .................................................... 87

7.0 Attributions and Annotations .................................................................................................. 88

References ........................................................................................................................................ 91

ATTACHMENT A, FUEL FABRICATION PROCESS ....................................................................... 106

ATTACHMENT B, URANIUM (AND THORIUM) SCRAP PROCESSING ........................................ 119

ATTACHMENT C, ALLOWABLE QUANTITIES OF RADIOACTIVE MATERIAL AT 

GENERAL ATOMICS ......................................................................................... 123

ATTACHMENT D, LABORATORY AND FACILITY THROUGHPUT INFORMATION 

FOR 1972 .......................................................................................................... 126

ATTACHMENT E, EXAMPLES OF ACTIVITIES AND AIR MONITORING IN THE 

PLUTONIUM LABORATORY ............................................................................. 129

ATTACHMENT F, LINAC RADIOACTIVE MATERIAL PRODUCTION 

INFORMATION .................................................................................................. 131

ATTACHMENT G, RADIOLOGICAL LABORATORY HAZARD CLASSIFICATION 

SCHEME ............................................................................................................ 133

ATTACHMENT H, INSTRUMENTATION ON HAND AND DESIRED AS OF 

SEPTEMBER 1960 ............................................................................................ 134

ATTACHMENT I, SUMMARY OF SOME GA ACCIDENTS AND INCIDENTS ................................ 137

ATTACHMENT J, HTGR AND RERTR IFM INFORMATION ........................................................... 139
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ATTACHMENT K, HOT CELL FACILITY REMEDIATION INFORMATION ...................................... 140
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LIST OF TABLES 

TABLE TITLE PAGE

2-1 Radiation-producing equipment in the Science Laboratory as of April 1971 ........................... 40

2-2 Summary information for selected ICA forms ......................................................................... 41

2-3 Radionuclides in specific locations in the Laboratory .............................................................. 42

2-4 Radiological monitoring from December 15, 1960, through December 31, 

1964....................................................................................................................................... 43

3-1 Bioassay codes ...................................................................................................................... 49

3-2 Uranium isotopic half-lives and specific activities ................................................................... 52

3-3 Uranium mixtures specific activity and isotopic fractions ........................................................ 52

3-4 Estimated recycled uranium activity fractions for internal dose reconstruction ........................ 53

3-5 HLC alpha sampling results ................................................................................................... 57

3-6 Isotopic composition by weight and activity percent of unirradiated plutonium 

grades .................................................................................................................................... 57

3-7 Urine activity fraction for bioassay codes PF, MFP and β or gross beta ................................. 60

3-8 Fission and activation product decay times for GA facilities and/or activities .......................... 60

4-1 Dosimeter type, period of use, exchange frequency, MDL, and potential 

annual missed dose ............................................................................................................... 64

4-2 External penetrating dose summaries for 1960 through 1964 ................................................ 65

4-3 DOS column code .................................................................................................................. 66

4-4 Exposure periods for monitoring devices for 1961 .................................................................. 66

4-5 MON column codes ................................................................................................................ 66

4-6 Neutron dose correction factors ............................................................................................. 67

5-1 Controlled area inhalation intake rates during the residual period .......................................... 72

5-2 Uncontrolled area inhalation intake rates during the residual period ....................................... 74

5-3 HLC alpha sampling results and intakes ................................................................................ 77

5-4 HLC beta sampling results and intakes .................................................................................. 78

5-5 Sitewide bounding controlled area inhalation intake rates during the residual 

period ..................................................................................................................................... 79

5-6 Sitewide bounding alpha sampling results and intakes ........................................................... 80

5-7 Sitewide bounding beta sampling results and intakes ............................................................ 80

5-8 Residual radiation exposure summary ................................................................................... 82

6-1 HCF air sampling results from 1965 through 1968 ................................................................. 85

6-2 HCF debris sampling results and remediation period alpha intakes ....................................... 86

6-3 HCF debris sampling results and remediation period beta intakes ......................................... 87

C-1 Summary of licensed source terms ...................................................................................... 124

D-1 Science Laboratory 1972 annual inventory information ........................................................ 126

E-1 Plutonium Laboratory workplace air sample results.............................................................. 129

F-1 LINAC radioactive material production information............................................................... 131

G-1 Classification of isotopes according to relative radiotoxicity per unit of activity ..................... 133

G-2 Radioisotope laboratory classification .................................................................................. 133

H-1 Instrumentation on hand and desired as of September 1960 ............................................... 134

I-1 Summary of some GA accidents and incidents .................................................................... 137

J-1 Radionuclide content of HTGR and RERTR IFM, September 30, 2003 ................................ 139

K-1 Sequence of D&D approach ................................................................................................. 140

K-2 Concentrations of radionuclides found in soil, asphalt, and concrete before 

remediation of the HCF ........................................................................................................ 141
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LIST OF FIGURES 

FIGURE TITLE PAGE

2-1 Site map, 1977 ....................................................................................................................... 15

2-2 HCF floor plan about 1960 ..................................................................................................... 19

2-3 Recovery Operations Area ..................................................................................................... 24

2-4 Fuel Fabrication Facility layout, 1974 ..................................................................................... 26

2-5 Fuel Fabrication Facility Mezzanine, 1974 ............................................................................. 27

2-6 Uranium Purification Area ...................................................................................................... 28

2-7 Uranium Purification Area ...................................................................................................... 29

2-8 LINAC Complex ..................................................................................................................... 30

2-9 Plan view of TRIGA Reactor Facility ...................................................................................... 34

2-10 Site map showing location of Science Laboratory .................................................................. 35

2-11 Floor plan of Science Building A ............................................................................................. 36

2-12 Floor plan of Science Building C ............................................................................................ 36
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ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission 

ANL-E Argonne National Laboratory-East 

AWE Atomic Weapons Employer 

cm centimeter 

Ci curie 

CIRGA Critical Isotope Reactor, General Atomic 

cph counts per hour 

cpm counts per minute 

D&D decontamination and decommissioning 

DAC derived air concentration 

DHHS U. S Department of Health and Human Services 

DOE U.S. Department of Energy 

DOL U.S. Department of Labor 

dpm disintegrations per minute 

EBOR Experimental Beryllium Oxide Reactor 

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000 

ESTES Engineering Scale Tritium Extraction System 

FFF Fuel Fabrication Facility 

fpm feet per minute 

ft foot 

g gram 

GA General Atomics 

gal gallon 

GM Geiger-Müller 

HCF Hot Cell Facility 

HEPA high-efficiency particulate air 

HEU highly enriched uranium 

HLC high-level cell 

hr hour

HTGR high-temperature gas-cooled reactor 

ICA Isotope Committee Approval 

ICRP International Commission on Radiological Protection 

IFM irradiated fuel material 

IMBA Integrated Modules for Bioassay Analysis 

in. inch 

IPC internal proportional counter 

IREP Interactive RadioEpidemiological Program 

keV kiloelectron-volt, 1,000 electron-volts 

kg kilogram 

kVp applied kilovoltage; peak kilovoltage 

kW kilowatt 
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L liter 

LINAC linear accelerator 

LLC low-level cell 

LLD lower limit of detection 

lpm liters per minute 

m meter 

mA milliampere 

MC metallurgy cell 

mCi millicurie 

MCi megacurie 

MDL minimum detection level 

MeV megaelectron-volts, 1 million electron-volts 

MFP mixed fission product 

MGCR Marine Gas-Cooled Reactor 

mi mile 

min minute 

ml milliliter 

MPC maximum permissible concentration 

mR milliroentgen 

mrem millirem 

NCRP National Committee on Radiological Protection and Measurements 

NERVA Nuclear Engines for Rocket Vehicle Applications 

NIOSH National Institute for Occupational Safety and Health 

NRC U.S. Nuclear Regulatory Commission 

NTA nuclear track emulsion, type A 

OCAS Office of Compensation Analysis and Support 

ORAU Oak Ridge Associated Universities 

PAEC potential alpha energy concentration 

PAEE potential alpha energy exposure 

PIC personal ionization chamber; pocket ionization chamber 

pCi picocurie 

POC probability of causation 

ppm parts per million 

psi pounds per square inch 

QF quality factor 

R roentgen 

RAM Radiation Area Monitor 

RERTR Reduced-Enrichment Research and Test Reactor 

RWP Radiation Work Permit 

SEC Special Exposure Cohort 

SNAP Space Nuclear Auxiliary Power 

SNM special nuclear material 

SRDB Ref ID Site Research Database Reference Identification (number) 

STSF Sub-Critical Time of Flight Spectrum Facility 

SV Sorrento Valley 
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TCF Thermionic Critical Facility 

TLD thermoluminescent dosimeter 

TRIGA Training, Research, Isotopes, General Atomics 

U.S.C. United States Code 

USASI United States of America Standards Institute 

WA Work Authorization 

W watt 

WL working level 

WLM working level month 

yr year 

α alpha 

β beta 

γ gamma 

μCi microcurie 

μg microgram 

§ section or sections 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 11 of 143

1.0 PURPOSE AND SCOPE

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

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

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

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction. 

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

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

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA. 

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007a).

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2007a):

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

conventional structures 

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

reasons 

 

1

 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 12 of 143

Under EEOICPA, employment at an AWE facility is categorized as either (1) during the DOE contract 

period (i.e., when the AWE was processing or producing material that emitted radiation and was used 

in the production of an atomic weapon), or (2) during the residual contamination period (i.e., periods 

that NIOSH has determined there is the potential for significant residual contamination after the period 

in which weapons-related production occurred). For contract period employment, all occupationally 

derived radiation exposures at covered facilities must be included in dose reconstructions. This 

includes radiation exposure related to the Naval Nuclear Propulsion Program and any radiation 

exposure received from the production of commercial radioactive products that were concurrently 

manufactured by the AWE facility during the covered period. NIOSH does not consider the following 

exposures to be occupationally derived (NIOSH 2007a):

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

conventional structures 

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

reasons 

For employment during the residual contamination period, only the radiation exposures defined in 42 

U.S.C. § 7384n(c)(4) [i.e., radiation doses received from DOE-related work] must be included in dose 

reconstructions. Doses from medical X-rays are not reconstructed during the residual contamination 

period (NIOSH 2007a). It should be noted that under subparagraph A of 42 U.S.C. § 7384n(c)(4), 

radiation associated with the Naval Nuclear Propulsion Program is specifically excluded from the 

employee’s radiation dose. This exclusion only applies to those AWE employees who worked during 

the residual contamination period. Also, under subparagraph B of 42 U.S.C. § 7384n(c)(4), radiation 

from a source not covered by subparagraph A that is not distinguishable through reliable 

documentation from radiation that is covered by subparagraph A is considered part of the employee’s 

radiation dose. This site profile covers only exposures resulting from nuclear weapons-related work. 

Exposures resulting from non-weapons-related work, if applicable, will be covered elsewhere. 

NIOSH has determined, and the Secretary of the U.S. Department of Health and Human Services 

(DHHS) has concurred, that it is not feasible to reconstruct internal radiation doses from exposures to 

thorium for General Atomics (GA) employees during the period from January 1, 1960, through 

December 31, 1969. As stated in the findings of the Secretary of DHHS, “Although NIOSH is unable 

to adequately estimate total internal exposures for members of the [SEC] class, internal dose due to 

intake of uranium can be reconstructed for exposures starting in October 1963, and tritium internal 

doses can be estimated after September 1965. NIOSH considers the reconstruction of occupational 

external radiation doses, including medical dose to be feasible for members of the class.” (Leavitt 

2007). The Secretary of DHHS has designated a single class of employees from GA for inclusion in 

the Special Exposure Cohort (SEC) (Leavitt 2007). 

The approved SEC class includes all Atomic Weapons Employer (AWE) employees who were 

monitored or should have been monitored for exposure to ionizing radiation while working at the 

following GA locations: Science Laboratories A, B, and C (Building 2); Experimental Building 

(Building 9); Maintenance (Building 10); Service Building (Building 11); Buildings 21 and 22; Hot Cell 

Facility (HCF) (Building 23); Waste Yard (Buildings 25 and 26); Experimental Area (Buildings 27 and 

27-1); LINAC Complex (Building 30); HTGR-TCF or Experimental Critical Facility (Building 31); Fusion 

Building (Building 33); Fusion Doublet III (Building 34); Sorrento Valley (SV)-A (Building 37); SV-B 

(Building 39); and SV-D (no building number) for a number of workdays aggregating at least 250 

workdays from January 1, 1960 through December 31, 1969, or in combination with workdays within 

the parameters established for one or more other classes of employees in the SEC. Although the 

SEC Petition Evaluation, Report SEC-00064 (NIOSH 2006a) focused on the inability to estimate 

internal dose from thorium during the period from January 1, 1960, through December 31, 1969, it is 

not currently feasible to estimate internal doses for other inadequately monitored GA source terms 
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(uranium before October 1963 and tritium before October 1965). Monitored exposures during the 

residual radiation and remediation periods and bounding doses for the residual and remediation 

period source terms can be used to determine the dose for GA personnel. 

This site profile provides information to reconstruct those doses for which applicable monitoring and 

source term information is available. 

Section 2.0 of this document describes the GA site and its history including information about the 

radiological processes and source terms as well as the radiological controls and monitoring practices. 

Sections 3.0 and 4.0 discuss internal and external dose assessment, respectively. Section 5.0 

provides information for assigning dose during the residual radiation period beginning in 1970, after 

the cessation of operations under the U.S. Atomic Energy Commission (AEC), and extending through 

1995. Section 6.0 provides information for assigning dose during the DOE remediation period from 

1996 through 1999. 

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

justification, or clarification of the associated information, are presented in Section 7.0. 

2.0 GENERAL SITE DESCRIPTION AND OPERATIONAL HISTORY

General Dynamics Corporation acquired property in La Jolla, California, in 1956 where the John J. 

Hopkins Laboratory for Pure & Applied Science – later named the General Atomic Division of General 

Dynamics facility – was built. The information that follows applies to a period of AEC radiological 

operations at GA in La Jolla, California, from January 1, 1960, to December 31, 1969. This document 

assumes that the residual contamination period was from January 1, 1970, through December 31, 

1995. DOE remediation activities are listed by the DOE Office of Worker Advocacy and are assumed 

to run from January 1, 1996, through December 31, 1999. Although beryllium exposure is not 

considered in this part of the program, it is noted that the covered period for beryllium began a year 

before the period for radiation exposure. 

The GA source term includes depleted, normal, and highly enriched uranium (typically greater than 

20% and up to 93.5% 235U by mass) in the form of scrap and reactor fuel both in a fresh form and in 

an irradiated form that includes uranium as well as activation and fission products. Thorium was used 

in fuel fabrication work. Plutonium work also occurred on the site. Four linear accelerators (LINACs) 

and several nuclear reactors were also available. Sealed sources were used for instrument 

calibrations and other work. Byproduct materials (atomic numbers 3 to 83) and tritium were also 

used. 

According to a company pamphlet, General Dynamics Corporation established General Atomic 

Division on July 18, 1955, in San Diego, California, to develop commercial applications for nuclear 

technology. GA was also known as John Jay Hopkins Laboratory for Pure and Applied Science; Gulf 

General Atomic; Gulf General Atomics Company; Gulf Energy & Environmental Systems; and GA 

Technologies. At one point they were held in partnership between Gulf Oil Corporation and Scallop 

Nuclear, and the partnership was known as Valley Pin Associates. 

2.1 SITE DESCRIPTION 

The current GA main site, a 60-acre complex on Torrey Pines Mesa in La Jolla, California (GA 1998), 

is the facility location that is covered by EEOICPA. The main site is approximately 300 ft above sea 

level, 1 mi from the Pacific Ocean, and 13 mi northwest of downtown San Diego. The GA site is in the 

center of Torrey Mesa Science Center, a 304-acre industrial park. The majority of GA buildings are 

on top of the mesa, but some work [e.g., at the Fuel Fabrication Facility (Building 37)] and the Fusion 

Energy Facility (also known as DIII-D National Fusion Facility) occurred (and some still occurs) at the 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 14 of 143

base of the mesa to the east of the central set of site buildings. Radiological work at facilities on the 

mesa included the HCF (Building 23) activities; Training, Research, Isotopes, General Atomics 

(TRIGA) reactor facilities activities; LINAC work; and neutron studies. In addition, research and 

development, including metallurgical and chemistry work, was occurring in the Science Laboratory 

and Experimental Buildings. The Fuel Fabrication Facility and the HCF were both in operation by the 

end of the 1950s. The HCF included support for high-temperature gas-cooled reactor (HTGR) and 

TRIGA reactor fuel research and development. Beginning in October 1966, a Plutonium Laboratory 

that was equipped with gloveboxes and high-efficiency particulate air (HEPA) filtration exhaust 

systems was activated for the first time in the HCF (Gurren 1966a). General Dynamics/Convair work 

at the Kearny Mesa Site in San Diego, California, which is not listed as a covered facility by the DOE 

Office of Worker Advocacy, is not specifically considered in this site profile, but is mentioned because 

of the possibility that dosimetry records for some workers might include exposures from both the San 

Diego site and the La Jolla site. Work on the Experimental Beryllium Oxide Reactor (EBOR) project 

might have occurred both in California and at the Idaho National Laboratory, and dosimetry records 

from both sites might be needed to reconstruct dose for these workers. 

GA possessed and used radioactive materials under several licenses and its radiological programs 

were regulated by the Federal Government and the State of California. 

2.1.1 Facilities

Due to facility changes at GA over approximately 50 years of site operations, Figure 2-1 was 

considered to be the best figure available to indicate locations of buildings on the site that would have 

been related to EEOICPA work. 

The following is a list of facilities that have been identified from available documents. This list is not 

considered to be a comprehensive listing of all the GA facilities and their associated radiological 

operations, but rather provides an overview of facilities and identifies most radiological operations 

onsite. 

Main Campus (10955 John Jay Hopkins Drive) 

 Building 1 Administration Building 

 Building 2 Science Laboratories A, B and C 

 Building 9 Experimental Building (E Building) 

 Building 10 Maintenance Building (current Health Physics offices) 

 Building 11 Service Building (early Shipping and Receiving) 

 Building 13 Technical Office Building (T.O. Bldg) 

 Building 14 Technical Office East – Building #1 

 Building 15 Technical Office East – Building #2 

 Building 21 TRIGA Building (previously contained three TRIGA reactors) 

 Building 22 TRIGA Fuel Laboratory Building 

 Building 23 HCF (also contained BeO, EBOR, tritium production, and a room where plutonium 

was handled, which was sometimes referred to as the Plutonium Laboratory) 
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Figure 2-1. Site map, 1977 (GA 1978, p. 61). 

 Building 25 Waste Yard/Incinerator 

 Building 27 Experimental Area Building #1 

 Building 27-1 Experimental Area Building #1 – Bunker 

 Building 29 Experimental Area Building #2 

 Building 30 LINAC Complex 

 Building 31 HTGR-TCF 

 Building 33 Fusion Building 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 16 of 143

 Building 34 Fusion Doublet III Building 

Facilities Off the Main Campus 

 Building 37 Sorrento Valley Building “A” (SV-A, 11220 Flintkote Avenue)

 Building 39 Sorrento Valley Building “B” (SV-B, 11220 Flintkote Avenue)

 Building (number unknown) Sorrento Valley Building “D” (SV-D, 11388 Sorrento Valley Road)

Administration Building 1 

This is an office building. No records of radiation, radioactivity, or radioactive contamination in this 

building were found. 

Building 2, Science Laboratories A, B, and C (SNM-69 1963, p. 17) 

These two-level buildings (in addition to Laboratory B) are attached end-to-end and make up the outer 

annular ring that surrounds the central library and cafeteria. (From an aerial view, these buildings 

distinctly appear as the outer ring of a bull’s-eye.) The Laboratory buildings were approximately the 

same size and approximately equal to one-third of the entire structure. They were arranged in a 

counter-clockwise direction from A to C to B beginning with the southernmost section. These 

buildings contained offices, scientific and electronic laboratories, shops, and low-level hot cells. The 

majority of the activities of the GA research in metallurgy, chemistry, and experimental physics have 

been conducted in these buildings. 

Building 9, Experimental Building (E Building) (SNM-69 1963, pp. 18 and 194) 

The Experimental Building was involved in much of the initial fuel development work that was 

conducted on the site. The building housed a drafting room, engineering and metallurgical 

development facilities, machine shops, a special materials machining room, and a hydriding room. 

The major early operation in the facility was the fabrication of TRIGA reactor fuel elements (UZrH, a 

uranium-zirconium hydride alloy). There were two radiologically controlled areas in the structure that 

were designated as Zones A (South Control Zone: A-1, A-2, and A-3) and B (Hot Suite: B-1, B-2, 

and B-3). 

All operations in which generation of particulate matter was possible were conducted in hoods or 

gloveboxes, or were serviced by individual elephant trunks from an exhaust system. Because GA had 

contracts with commercial and government clients, a major aspect of working in these areas was 

material control to minimize the chance of commingling different enrichments of special nuclear 

material (SNM). Procedural and administrative controls were in place to minimize this occurrence and 

these controls typically aided the health physics group in relation to personnel protection. There were 

two SNM storage vaults adjacent to the east wall of the facility and a Security storage area that was 

designed to store materials in Zone A for short periods. 

At a later date, a plutonium storage vault was located in the Experimental Building (SNM-69, 

Amendments 1-41, p. 156). It appears that the material was designated as being “C.4, encapsulated,” 

on the license. It is not clear when this SNM vault came into existence or what the material was 

ultimately used for, although it appears that work with thermionic cells was conducted at some time in 

the building. It appears the vault was used to store encapsulated plutonium used in the Space 

Nuclear Auxiliary Power (SNAP) 15A program, because the SNAP Plutonium Handling Room had 

been relocated from the Sorrento Valley B (SV-B) Building in October 1965 (Bold 1965a). 

There were many operations that took place in this facility. Uranium stock material had to be sized to 

less than 0.125-in. cubes. The uranium was cold-rolled into a sheet, annealed, sheared to size in a 

ventilated hood, and then sent to a storage vault. The shaped product fuel (including uranium, 

zirconium, and other additives) was either formed from the uranium-zirconium mixture being directly 

melted and cast in an induction furnace or the mixture was compacted (pressed), arc-melted, cast into 

4-in.-diameter cylinders again using arc melting, machined, canned in copper, extruded, pickled to 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 17 of 143

remove the outer copper layer, and then cut to length. Each of these operations took place in the 

Experimental Building. 

Building 10, Maintenance Building 

There was not much information found during data capture activities that described the operations 

that took place in this building. It is presumed that due to its location in proximity to Building 9 (the 

Experimental Building) that it primarily supported activities in that structure. It currently houses the 

site’s Health Physics professional staff. A radioanalytical laboratory toward the center of the building 

analyzes operational, environmental, and remediation samples. 

Building 11, Service Building (early Shipping and Receiving) (SNM-69 1963, p. 20) 

The building no longer exists, but it originally housed early administrative functions such as 

purchasing, employment, and shipping and receiving for the entire site. Shipments were received, 

surveyed, and transported from the facility to other buildings on the site for storage or use. In August 

1961, storage of radioactive items among surplus items began (Bethard 1961a). The New Delhi 

TRIGA reactor was stored here for a period. A control area was set up to segregate this material from 

other nonradioactive material. 

Building 13, Technical Office Building (T.O. Building) 

This is an office building. Documentation was found that indicated that a portion of the New Delhi 

TRIGA reactor was stored in the basement of this facility and later relocated to Building 11. No other 

records of radiation, radioactivity, or radioactive contamination in this building were found. 

Building 14, Technical Office East, Building #1 

This is an office building. No records of radiation, radioactivity, or radioactive contamination in this 

building were found. 

Building 15, Technical Office East, Building #2 

This is an office building. No records of radiation, radioactivity, or radioactive contamination in this 

building were found. 

Building 21, TRIGA Building (previously contained three TRIGA reactors) 

This building is north of the Science Laboratories and contained three TRIGA reactors. Three reactor 

licenses were found for the site (R-38, -67, and -100). Licenses R-38 and -67 mention two TRIGA 

reactors (Mark I and F), which were shown on figures as located at the site. These reactors came on 

line in 1957 and 1960, respectively. The Mark F reactor was sometimes designated as the Flash 

Irradiation Reactor (FLAIR). An additional reactor, which came online in the mid-1960s and was 

designated as Mark III, was noted in several documents, but no figure (other than a very rough 

sketch, SNM-69 1963, p. 114) has been found to identify its location. These pool-type reactors were 

designed to be inherently safe in that the fuel design eliminated the potential for a reactor to go 

supercritical. The reactor facility was used to test the various designs of the GA-designed reactors. 

Building 22, TRIGA Fuel Fabrication Facility (SNM-69 1963, p. 227) 

Amendment 61 (January 9, 1975) of SNM-696 identified the TRIGA Fuel Laboratory Building as being 

in the TRIGA Fuel Fabrication Facility. This building is east of the HCF (Building 23). It is constructed 

primarily of 7-in.-thick prestressed or reinforced concrete. (In addition to the Fuel Laboratory, the 

building also contains several nuclear material storage areas or vaults.) The Fuel Laboratory was 

divided into six separate zones: metal preparation, furnace room, hydride room, machine shop, 

assembly and inspection room, office, and locker room. All building operations that generated 

airborne radioactivity were conducted in closed equipment, hoods, or gloveboxes, or the radioactivity 

was exhausted from the area via elephant trunks. 
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Building 23, HCF (also contained BeO, EBOR, tritium production, and a plutonium handling 

room) (GA 1998, p. 18) 

The HCF was constructed in 1959 and was operated for over 30 years. It was utilized primarily to 

facilitate remote inspection of irradiated, highly radioactive HTGR and TRIGA fuel as well as 

thermionic fuel elements for space power applications (Aycock 1995). Some of the very earliest 

examinations involved UO2-BeO fuel for the Marine Gas-Cooled Reactor (MGCR), which later 

became known as EBOR. Figure 2-2 shows the layout. Approximately 12% of the building footprint 

was taken up by the hot cell (SNM-69 1963, p. 31). The hot cell was made up of three sections: the 

high-level cell (HLC), low-level cell (LLC), and metallurgy cell (MC). The HLC was used as a remotely 

operated hot machine shop to cut open and section irradiated fuel samples for inspection. The MC 

was used for analytical inspection and research activities; including the preparation of samples, which 

entailed grinding, polishing, and mounting. The LLC was used as an interface between the other hot 

cells and other areas of the building, and housed all shipping cask off-loading and packaging 

operations. Thick (up to 42-in.) high-density concrete walls were in place to minimize personnel 

exposure to penetrating radiation. 

In addition to activities that took place in the hot cells, there were four additional, notable operations 

that took place in the building: (1) development, fabrication, and inspection of UO2-BeO fuel materials 

in support of the EBOR fuel project (Rooms 108/108A) from 1960 to 1972; (2) tritium extraction 

laboratory for preparation and testing of irradiated fuel samples in Room 109 from 1969 to 1987; 

(3) an Engineering Scale Tritium Extraction System (ESTES) that was used to demonstrate the ability 

to extract tritium from New Production Reactor target compacts (Rooms 109, 119, and 120) from 1987 

to 1991; and (4) plutonium transport experiments using PuO2 (Room 116). The two rooms added to 

support EBOR were included late in 1959 during the final phase of construction to accommodate the 

Fuel Fabrication Facility (Aycock 1995). 

The airflow throughout the facility was directed from clean areas towards areas with high potential for 

contamination. A continuous air monitor was located in the operation gallery (workspace immediately 

adjacent to the hot cells) to signal elevated air concentrations, and continuous health physics 

coverage was provided for the facility. 

Building(s) 25, Waste Yard/Incinerator 

The waste processing facility includes several buildings that are associated with waste handling, 

minimization, processing, and storage; SNM and byproduct material storage; and a gamma counting 

facility. Liquid wastes that were generated on the GA site were treated by placing them in evaporation 

pools. A low-level radioactive waste incinerator was also used to burn unirradiated industrial trash 

containing 235U. Typical materials included paper, plastic shoe covers, gloves, mops, plastic bags, 

etc. from GA laboratories and fuel fabrication facilities. Only authorized personnel were allowed into 

the fenced-off incinerator area, and they were required to wear a film badge and self-reading 

dosimeter and to submit bioassay samples for analysis (SNM-69 1963, p. 325). 

Several locations were used in the waste processing facility area to store radioactive materials 

including TRIGA hot storage, a temporary storage yard, the Butler Building, and the By-Products 

Vault. 

A gamma counting facility is also located in the controlled area for the purposes of confirming the 

amount of 235U present in containers before receipt by the waste processing facility. 

Building 27, Experimental Area Building #1 (EA 1 Laboratories) 

EA-1 was constructed between September 15, 1964, and April 7, 1965, on a site adjacent to and 

immediately west of the TRIGA Reactor Facility. (This facility was constructed to be an Annex to the 

TRIGA Reactor building, Building 21 [1]). The building was constructed of concrete masonry material 

with exterior dimensions of 41.3 ft by 112 ft with interior floor space of approximately 4,400 ft2. There 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 19 of 143

Figure 2-2. HCF floor plan about 1960 (GDC and H&N 1960, p. 89). 

were no interior load-bearing walls, so the facility could quickly and easily be modified (Hill 1997). 

The GA Neutron Activation organization occupied the building immediately after construction. This 

group established offices in the south end and northeast corner, and set up radiochemistry 

laboratories in the rest of the building. The laboratories were used primarily to prepare target samples 

for neutron irradiation and to perform postirradiation radiochemical separations. The GA Analytical 

Chemistry and Radiochemistry groups were combined into the GA Chemistry group, which then 

occupied the whole building except the laboratory in the northwest corner. Operation of the GA 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 20 of 143

Chemistry organization ended in 1996, and administrative control of the facility was assigned to the 

GA Decontamination and Decommissioning (D&D) organization (Hill 1997). 

Building 27-1, Experimental Area Building #1 – Bunker 

Construction details were not obtained from document capture, but rather from a discussion with a 

site radiochemist who was and is still employed at the GA site [1]. The facility was constructed in the 

same period as Building 21 and is immediately north of that building. Much of the structure was built 

into the mesa. During facility remediation, the southernmost walls of the structure were left in place to 

support the face of the mesa [2]. The Neutron Activation organization used the facility to conduct 

activities that did not generate much long-lived contamination. Some postirradiation examinations of 

Peach Bottom fuel elements were conducted there [1]. Later a three-unit hot cell facility was set up in 

the building to “milk” 90Y from large 50-Ci quantities of 90Sr. The facility was called the Strontium 

Facility. 

Reports mention the 90Y production facility, molybdenum facility, and chemistry laboratories. Monthly 

reports in the mid-1960s discuss the use of titanium tritide associated with the TRIGA reactor 

building. It appears that targets were irradiated in the TRIGA reactor and then the Activation Analysis 

group worked with the targets in the Experimental Area Building #1 – Bunker before they moved to 

their new location. Thermionic generators were also in these areas where the Analytical Analysis was 

conducted. 

Building 29, Experimental Area Building #2 

Phone conversation with the Lead GA D&D technician indicated that this facility was and is an office 

building and that radioactive material was never handled or processed in the facility [2]. Another 

conversation with an individual in the Site Radiochemistry group indicated that the building originally 

housed the medical offices and that Frank Bold, one of the early Health Physics Department 

managers, had an office there [1]. At present the facility is occupied by a company that conducts 

classified work. 

Building 30, LINAC Complex (SNM-69 1963, p. 18) 

The LINAC complex includes both a 17-MeV and a 45-MeV electron accelerator that can be coupled 

to provide even greater acceleration. This facility has been in continuous operation since September 

1958. For the majority of accelerator experiments, the LINAC was not used in conjunction with the 

critical assemblies. The console and administrative office areas were separated from the 

experimental area. There was a large concrete enclosure called “the cave” in the center of the 

experimental area that was a highly shielded area for neutron-thermalization studies. 

Building 31, HTGR-TCF (SNM-69 1963, p. 332) 

This building complex appears to have been a critical test facility for several programs including the 

HTGR and Thermionic Critical Facility (TCF), as well as being called the Experimental Critical Facility. 

Information about the activities that took place in the Experimental Critical Assembly Cell indicates 

that many different critical assembly tests were conducted in this facility. 

Building 33, Fusion Building 

Divided into a machine building and a setup laboratory annex, the Fusion Buildings were designed for 

large-scale experimental apparatus for the study of controlled thermonuclear reactions. Experiments 

were conducted here with high-temperature plasmas. Activities in this facility preceded the activities 

in Building 34. 

Building 34, Fusion Doublet III Building 

This facility is in Sorrento Valley at the termination of Tower Road, approximately 1,000 ft south of 

Building SV-A (below). It conducts research in magnetic fusion technologies. The operations 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 21 of 143

generate some intense but short-lived radiation sources. Administrative and procedural controls are 

in place to minimize personnel exposure. 

Building 37 or SV-A, Component and Fuel Manufacturing Building (SNM-696 multiple dates, 

p. 11) 

The Component and Fuel Manufacturing Building, SV-A, was approximately 1 mi north of the GA 

Main Campus facilities, off the Torrey Pines Mesa in the Sorrento Valley at an elevation of 40 ft above 

the floor of the valley. The building was in an area of approximately 60 acres. The area was 

physically separate from the Main Campus, and transportation of radioactive materials between the 

two GA sites had to take place over U.S. Department of Transportation-regulated roads. The facility 

was approximately 400 ft long by 120 ft wide. The front (east) portion of the building had two floors 

that included offices, change rooms, laboratories, storerooms, and a cafeteria. A metal machine 

shop, sheet metal shop, and assembly area for mechanical parts also were located in the building. 

The northern portion of the building (about 160 ft between columns 25 and 41) was devoted to fuel 

fabrication activities. This operational area was physically separated from all other areas in the 

building by either a sheet metal wall (to the south) or a masonry wall (to the east). At some time after 

original building construction (and original license submittal), there was an addition to the north end of 

the facility that was “created primarily for storage.” 

The use of SNM in the building was primarily for the fabrication of HTGR fuel materials and elements 

and took place in the 160-ft northern section of the facility. A letter and number designation was used 

to identify each station in the fuel fabrication area of the building, and each station is further defined in 

SNM-696 (multiple dates). The processes were identified in the following manner: 

 Fabrication of pyrolytic carbon-coated uranium oxide or carbide and uranium-thorium oxide or 

carbide particles at the process stations that were designated by the letter A. 

 Fabrication of compacts that consisted of coated particles in combination with other materials. 

These process stations were designated by the letter B. 

 Assembly of compacts and graphite components into finished fuel elements took place at 

process stations that were designated by the letter C. 

 Fabrication of compacts composed of uranium oxide (up to fully enriched 233U, or up to fully 

enriched 235U with maximum enrichment in 236U) and graphite and further into finished fuel 

elements. [No letter designation was given for these operations.] 

 Fabrication of fuel containing a minimum ratio of 3.60 thorium-to-uranium at process stations 

that were designated by the letter E. 

 Fabrication of fuel elements that contained fuel rods that were composed of coated uraniumthorium oxide or carbide particles in combination with other materials took place at process 

stations that were designated by the letter F. 

Radiation protection surveys included 35 air sample locations that were in place in the facility that 

were supported by a house vacuum system. Each air sampling location had a dry gas meter that 

measured total sampled air and used 2-in.-diameter filter paper (not defined). Air samples were 

collected for 8 hr on each shift in all areas where SNM was handled in a form that could become 

airborne. The samples were counted initially after a 4-hr decay and again after 3 days in the Health 

Physics Counting Room and analyzed for both gross alpha and beta activity. Site personnel who 

regularly worked in the fuel fabrication areas were provided with dosimetry that measured beta, 

gamma, thermal, and fast neutrons. In addition, area dosimetry was in place throughout both the 

operating and office areas in the building. All operating personnel were on a routine bioassay 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 22 of 143

monitoring program. Both direct and indirect radiation surveys were normally conducted on a daily 

basis. Protective clothing was company-supplied and required to be worn. 

Building 39 or SV-B, Pilot Plant Building or HTGR Fuel Production, Research, and 

Development Building (SNM-696, p. 83) 

The HTGR Fuel Production, Research, and Development Building, SV-B, was immediately northwest 

of SV-A. Its main purpose was process development and pilot-scale operations in relation to fuel 

production for HTGRs. Process development operations involved bench and pilot-scale operations, 

which typically involved fuel particle kernel manufacture, application of coating onto kernels to form 

coated fuel particles, forming carbonaceous matrix fuel bodies that contained the coated fuel particles, 

and fuel element assembly. The kernel formation included work with both uranium- and thoriumbearing solutions. The typical kernel forming process took place by flowing nitrate solutions through 

an orifice to make spherical droplets. These droplets were gelled by contact with an ammonia 

solution and subsequently dried. Calcining and sintering furnaces were used to form high-density 

oxide microspheres and these particles were then typically carbon-coated in fluidized bed reactors. 

Various metering and blending techniques were developed and tested for optimal fuel body 

performance in an HTGR. The facility conducted process development and refinement of fuel 

element assembly operations in which the fuel bodies were loaded into graphite fuel elements. 

Each handling location in the facility was limited to a maximum of 350 g of 235U. These locations 

consisted of fume hoods, gloveboxes, workbenches, and other types of processing equipment. 

Ventilation (using either fixed or flexible ducting) was provided wherever airborne contamination might 

have occurred. Work with powdered SNM was conducted in fully contained glovebox enclosures. 

Radiation protection surveys in the facility were conducted at least weekly. The action level for wipe 

survey results was compared to a limit of 6 dpm/100 cm2 for controlled areas and, in the event that 

loose contamination was detected exceeding this level contamination control, boundaries were set up. 

Uncontrolled areas in the building were maintained below this level. 

SV-D, Special Nuclear Systems (SNM-69 1963, p. 130) 

This concrete-reinforced structure (floor, walls, and roof are concrete) was approximately 3 mi

northeast of the main GA campus and assembled thermionic and thermoelectric fuel elements. A 

concrete block firewall was in place to physically separate the two operations. The thermionic 

operations in the facility had a site possession limit of 2.5 kg of 235U that was received and handled in 

the form of dense high-fired pellets of UO2. It is not clear from information provided in the site license 

(SNM-696), but encapsulated plutonium (PuN, PuO2, etc.) in a nuclear battery assembly was also 

handled in this section of the building. Radiation protection surveys were conducted throughout the 

facility, including wipe surveys (at least monthly in all areas) and air sampling (of several areas 

including welding operations that plutonium and before continuing any other operations in which 

positive contamination was detected). Thermoelectric generators were assembled in another area of 

the building. 

The SV-D facility possession limit, as of September 15, 1972, was 300 g of plutonium. Laboratories 

[e.g., Mound Laboratory or Harwell of the United Kingdom Atomic Energy Agency] supplied the 

double-walled, encapsulated-plutonium-containing fuel capsules in shipping containers. Capsules 

contained 0.18 g of 238Pu and 20% 239Pu (no amounts were specified for 236/240/241/242Pu, but lesser 

amounts of these isotopes would be expected). Receipt surveys of the entire outer surface were 

conducted and compared to an action level of 6 pCi (~10 dpm). Once the package was received, it 

was moved to the Plutonium Handling Room, placed in a glovebox where it was unpacked, and 

additional surveys with portable alpha scintillators and wipes for alpha activity were conducted. Fuel 

capsules were handled and generators were assembled in a sealed glovebox. When not being 

handled, the fuel capsules were stored in a safe in the Plutonium Handling Room. Assembled 

generators and shipping containers were checked by a Health Physics surveyor before their removal 

from the glovebox. The radioactive material was encapsulated in the generator and might have been 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 23 of 143

transported to other locations in SV-D or shipped to other facilities at the GA site for quality assurance 

examinations. 

General Dynamics/Convair (Kearny Mesa Site, San Diego, California) (SNM-69 1963, p. 170) 

This site is not part of the covered facility. It conducted early electron-beam welding services for 

235U-fueled thermionic generators that appear to have been manufactured in the Experimental 

Building on the main campus site (SNM-69 1963, p. 173). The generator was required to have an 

insulator that was welded into place, and apparently GA did not have the facilities at one time to 

perform this work. The units were inspected (e.g., spectrometric leak testing, radiographs, as well as 

other inspections that are not specifically described) on the GA main campus before shipment to the 

facility. If any defects were noted, the affected assembly was rejected and not shipped. 

The uranium fuel, in solid form, was encapsulated in tungsten containment. A GA health physicist 

was with the generator at all times during the welding activities. Air sampling and a portable alpha 

survey meter were used at all locations where the generator was handled or tested, and the health 

physicist routinely performed alpha wipe surveys (SNM-69 1963, p. 189). The welding at the Kearny 

Mesa site did not take place on the tungsten itself, but rather on a tantalum transition flange brazed to 

the tungsten. This ensured the integrity of the tungsten container would not be breached. The 

welded assembly was immediately returned by a GA representative back to the Main Campus. 

Green Farm Experimental Test Site 

This site is not part of the covered facility. The Green Farm Experimental Test Site was a 500-acre, 

U.S. Air Force explosive-firing facility that GA operated approximately 10 mi east of the main campus 

site near the Air Force Sycamore Canyon Missile test site (GA undated). The facility was originally 

set up for Project ORION. The facility had offices, shops, laboratories, explosive magazines, and 

firing bunkers. There was also a Flash X-Ray Facility at the Green Farm site to test both highexplosive and radiation effects experiments. The instrument operated at a nominal operating voltage 

of 2 MeV, and delivered 6,000-R X-ray doses in 30 nanoseconds at the tube face. A smaller flash 

X-ray machine was installed “in a special shield room” at the Torrey Pines site. This machine was 

used for performing experiments at low dose rates and for laboratory checkout of the instrumentation 

and experiments before irradiation at the LINAC 

2.2 PROCESS DESCRIPTIONS 

GA conducted irradiation and time-of-flight studies with four LINACs, developed and fabricated reactor 

fuels, operated three TRIGA reactors on the site, performed fusion research since the mid-1960s, ran 

experimental criticality test facilities, used an experimental facility that is understood to have 

conducted a diverse set of tests with radioactive materials, and worked with SNM and radioactive 

tracers in its laboratory facilities. Hot cells were used for handling radioactive materials that were 

associated with high radiation fields or very large quantities of radioactive materials that might be 

subject to processes that would disperse the material. 

AEC Contract Number AT(40-1)-3367 (AEC 1965), was for reclamation of highly enriched uranium 

(HEU) from unirradiated Nuclear Engines for Rocket Vehicle Applications (NERVA) scrap fuel. The 

scrap was identified in the contract as being in scrap groups III (uranium compounds such as oxide) 

and IV (combustibles such as paper, tissue wipes, filters, etc.). According to the terms of the contract, 

the scrap materials were to be shipped from the Westinghouse Electric Corporation’s Astrofuel Facility 

in Cheswick, Pennsylvania, to GA for processing into uranium oxides (Type III recovered uranium; 

UO2, UO3, and U3O8). This reclamation was to take place “no later than 150 days after delivery of the 

scrap” to GA. According to the “Special Conditions of the Contract,” the recovered HEU oxides were 

to be shipped to Portsmouth Gaseous Diffusion Plant. The individual who managed the GA Fuel 

Operations Division during the 1960s and through the 1980s (and currently works part time at GA) 

was interviewed and indicated that this HEU processing was one of several similar contracts to 
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process NERVA/Rover process scraps that were conducted through approximately 1972. The scrap 

recovery operations took place in the GA Fuel Fabrication Facility, Building 37, sometimes called 

SV-A (Figure 2-3). No other AEC contracts have been located. 

Figure 2-3. Recovery Operations Area (SNM-696, p. 176). 

The HCF, Building 23, consisted of centrally located hot cells (HLC, LLC, and MC), an adjacent 

operating gallery, support laboratories, change rooms, and administrative office areas. The HCF was 

designed and built in 1958 and 1959, and it was used from 1960 through 1969 to conduct the 

postirradiation examinations of DOE fuels, structural materials, reactor dosimetry materials, and 

instrumentation inspection of irradiated fuel. In late 1959, during the final phase of construction, GA 

decided to expand the original building design to include an additional 1,200 ft2 of laboratory space to 

accommodate the EBOR Fuel Fabrication Facility. Portions of the building have been used to 

conduct other radiological activities, including the development and manufacture of EBOR fuel (in
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 25 of 143

Rooms 108/108A), the engineering-scale tritium extraction system to support the New Production 

Reactor (Rooms 109, 119, and 120), and a plutonium transport experiment that used plutonium oxide 

(Room 116). The building was maintained in a surveillance and maintenance mode from 1991 until it 

was remediated. This building and adjacent area were remediated from 1992 to 1994 and were 

released by the State of California from further radiological control in 1994. 

GA maintains Isotope Committee Approval (ICA) and Work Authorization (WA) records that provide 

information that identified major radionuclides and quantities involved in the activity [many of these 

are currently available in Oak Ridge Associated Universities (ORAU) Team records]. Radiological 

controls to be used during the activity were specified, and the records were signed by the responsible 

investigator (sometimes names of other workers who were involved in the work were included). 

2.2.1 Fuel Fabrication 

Before this work moved to the Fuel Fabrication Facility (SV-A) in 1962, many of the activities were 

carried out in the Experimental Building. Related activities occurred in the Science Laboratory. 

Uranium (up to 93.5% enriched in 235U by mass) and thorium were used in this facility. Radioactive 

materials were coated, blended, compounded, heated, rolled, extruded, machined, compacted, and 

eventually made into fuel elements in the Fuel Fabrication Facility. The station mass limits were 

based on fully enriched uranium and no credit was taken for the neutron absorption effect of thorium. 

In those operations in which either hoods or gloveboxes were used, a minimum airflow of 150 fpm 

was supposed to be maintained to the facilities hood ventilation system. Figure 2-4 shows the layout 

with station designations in the Fuel Fabrication Facility and Attachment A provides a step-by-step 

description of the fuel fabrication process that is described in SNM License 696. Figure 2-5 shows the 

mezzanine of the Fuel Fabrication Facility. 

2.2.2 Uranium Purification

SNM in the form of impure uranium-bearing scrap materials such as oxides or carbides were 

processed to recover and purify the contained uranium. The material was accumulated from several 

sources (both on and off the site), including out-of-specification SNM product, process equipment, 

scrap, waste, and solid residue from liquid evaporation. Some of the materials were products from 

the facility recovery operations. Figures 2-6 and 2-7 show the layout of the purification operations at 

the Fuel Fabrication Facility. The uranium was recovered from unirradiated scrap, but these scraps 

probably included some uranium recovered from recycle operations. Records indicate that this area 

might have been used for thorium recovery as well. A GA letter (Bissonnette 1971) to the AEC on 

airborne radioactivity in excess of permissible limits in the scrap recovery (uranium-thorium) area of 

the Fuel Fabrication Facility on February 3 and 4, 1971, was the earliest documentation found that 

related to thorium scrap recovery. It is unknown whether thorium scrap recovery was occurring before 

1970.

SNM License 696 descriptions of the uranium material extraction and purification steps, and activities 

at specified workstations follow. The general steps are summarized (SNM-696, pp. 191–192): 

1. The impure uranium-bearing material was dissolved in an acid solution. 

2. The uranium was extracted from the acid solution by means of an organic extracting agent. 

3. Impurities that were extracted along with the uranium into the organic phase were removed by 

scrubbing the organic phase with an aqueous salting solution. 

4. The uranium compound was separated from the organic phase by an aqueous extraction. 
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Figure 2-4. Fuel Fabrication Facility layout, 1974 (SNM-696, p. 24

143

1). 
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Figure 2-5. Fuel Fabrication Facility Mezzanine, 1974 (SNM-696, p. 240). 
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Figure 2-6. Uranium Purification Area (SNM-696, p. 205). 

5. The uranium compound was separated from the aqueous solution by precipitation as a 

uranium salt. 

6. Purified U3O8 was obtained from the precipitate by heating in a controlled temperature furnace. 

A more detailed description of the process is provided in Attachment B. 

2.2.3 LINAC Facility

The LINAC complex in Figure 2-8 consists of two LINACs, one 17 MeV and one 45 MeV, which could 

be coupled to obtain greater acceleration. Plans were made to increase the coupled output to 94 

MeV when the 45-MeV klystrons were upgraded (GA undated). The accelerator beam was capable of 

being targeted to any of the three targeting cells (A, B, and C-D) as shown in the figure. The targeting 

could be changed during operations that required strict personnel entry controls and interlocks. 

The LINAC facility has been used for a variety of research activities over the years. The major 

program at the LINAC in 1962 (GA 1962) was Integral Neutron Thermalization; this program, which 

was sponsored by AEC, had as its goal the measurement of integral phenomena of neutron 

thermalization and moderated systems. The electron beam was aimed at a suitable target, which 
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Figure 2-7. Uranium Purification Area (SNM-696, p. 206). 

produced a burst of neutrons that passed through the test moderator. Moderated neutron energy 

spectra were then measured using time-of-flight techniques. 

Differential Neutron Thermalization. In this program, which was also sponsored by AEC, 

investigators attempted to measure such neutron parameters as transport cross-sections and a 

variety of materials. The data included energy loss per collision for number of moderators as well as 

inelastic scattering cross-sections as a function of several variables for different moderated materials. 

Multiplying Assembly. This program used a semicritical, semihomogenous uranium water system to 

produce an enhanced burst of neutrons when exposed by a neutron beam. The goal was to 

approximate the environment of an operating reactor. 

Neutron Capture. This program was to determine the neutron capture cross-sections for a large 

number of materials. 

In 1964, the above studies were still being performed with the addition of the following (GA 1964): 

 Fission. This was a study to investigate the intensity of the delayed radiations from the fission 

of 238U, 235U, 239Pu, and 232Th. 
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Figure 2-8. LINAC Complex 

 Positron and Photonuclear Transport. This created the world's most intense positron beam 

by bombarding a tungsten target that was located after the first section of the accelerator. The 

positron beam was used for such studies as the comparison of electron- and positron-induced 

nuclear reactions. 

 Using positron annihilation in flight, it was possible to generate monoenergetic gamma rays 

that were continuously variable between 3 and 30 MeV. 

 High-Energy Neutron Transport. This program studied the deep penetration characteristics 

of high-energy neutrons. This was an important consideration in the nuclear heating of liquid 

hydrogen rocket propellants and the shielding design for space nuclear systems. 

To ensure personnel safety, the LINAC used a combination of administrative controls with physical 

interlocks. Every electron beam radiation cell (a cell was any room into which the electron beam 

could be deflected) was equipped with a minimum of five safety devices and safety measures that 

were designed to protect personnel when in a cell. One of the safety devices was considered a 

primary safety device and controlled the high voltage on the klystrons. Four safety devices were 

considered secondary because they were in the personnel safety interlock chain. The five safety 

devices were as follows (Trimble et al. 1967, pp. 12–13): 

1. Cell Door Interlock. This interlock prevented the operation of the LINAC while the cell door 

was open. The cell door was required to be open during occupancy. 

2. Cell Safety Plugs. These were required to be in the receptacle by the door for the system to 

be energized. This interlock prevented targeting of the LINAC beam into the cell while a safety 

plug was removed. The safety plugs were at the entrance to each cell and had to be worn on 

the external clothing during cell occupancy. 
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3. Cell Safety Switch. This was a primary safety device that prevented operation of the LINAC 

whenever the switch handles were in the down position. 

4. Radiation Area Monitor (RAM). This was a gamma detection system that would shut down 

the LINAC when a preset level was reached in the cell. Each cell contained a local readout of 

the radiation in the cell, so the radiation level in the cell could be determined immediately. A 

remote reading was provided in the LINAC console room. When the preset level was reached 

the LINAC was shut down and an alarm sounded. The cell was immediately evacuated when 

the alarm sounded. 

5. Mode Switch. The mode switch prevented the entry of an electron beam into an active cell by 

placing certain personnel safety locks (e.g., a beam stopper) into the personnel safety 

interlock chain. The mode switch was in the LINAC console room and was controlled by the 

duty LINAC operator. 

The two safety measures were as follows: 

1. A Radiation Warning Light was activated when the LINAC was switched on in preparation of 

the electron beam to enter the active cell, and it remained on during LINAC operation. 

2. All the shielding doors had switches that allowed them to be opened from the inside. 

There was an additional interlock for any target that required cooling such that the LINAC would shut 

down should there be a loss of cooling (Trimble et al. 1967, p. 14). 

In the early days of operation this led to some potentially unsafe conditions. In one case an individual 

entered one of the cells without removing the necessary plug; this led to his almost being exposed to 

potentially fatal radiation (Bold and Trimble 1966). This incident was one of the prime reasons a set 

of radiological safeguard regulations was issued for the use of the LINAC in 1967 (Trimble et al. 1967) 

with a recommendation of penalties for noncompliance (Bold and Trimble 1966). These regulations 

contained rules for non-LINAC personnel (experimenter), the Duty LINAC Operator, the experimenter, 

and the health physics surveyor. The following is a summary of these rules: 

 Machining of Radioactive Material. Machining on any radioactive material in the LINAC 

machine shop (or in any other area of the LINAC facility) was subject to prior approval of the 

chairman of the LINAC safeguard committee. Any material that registered over twice 

background on an alpha, beta, or gamma detector was considered radioactive. 

 Electron Beam Irradiation Cell Entry Procedure. A cell was considered to be in one of two 

conditions during an 8-hour work shift: 

– Inactive – The accelerator was either not attempting to operate during the shift or the cell 

was opened for the rules that follow. 

– Active – The cell was prepared to receive the electron beam during the shift and the 

accelerator was only temporarily shut down. 

Entry to an inactive cell was permitted if the LINAC was not scheduled to operate during the shift, 

provided the health physics department had surveyed it since the last accelerator operation. The door 

was required to be open throughout occupancy. If the LINAC was operating, the operator had to 

declare the cell open with entry allowed subject to the following (Trimble et al. 1967, pp. 8–9): 

1. The request had to be made to the Duty LINAC Operator. 
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2. If entry was deemed feasible, the Duty Health Physics Surveyor and Experimenter in charge 

had to be notified. 

3. Confirmation of the securing of all electron beam stoppers, removable shielding pieces, and 

deflecting magnets had to be made by the Duty LINAC Operator. 

4. The LINAC mode switch setting had to be confirmed to show that it was not set to the cell 

being opened by the Duty LINAC Operator. 

5. The LINAC operator opened the door and a health physics employee performed a survey. If 

the beam was not steadily on target in an unoccupied cell, and the subject cell needed to be 

occupied, the HP surveyor had to remain on duty in the occupied cell until the desired 

targeting was achieved. When 100% power was achieved, the Duty LINAC Operator notified 

the surveyor, who did a final survey of the cell, and the Duty LINAC Operator would then 

determine if the cell could remain open. 

6. The cell door was to be left open during occupancy. The RAM had to be checked with a 

source, and the LINAC scram had to be proven to be operable with a check of all interlocks by 

the surveyor. 

7. Cell safety plugs were required to be worn on the external clothing of all cell occupants and 

returned to the plug board on leaving the cell. 

All cells were closed before the accelerator was moved to a new target. 

An active cell was never declared open. Entry into such a cell was made only by the Experimenter in 

charge or the Duty LINAC Operator and those under their immediate control. The following procedure 

had to be followed: 

1. The accelerator was turned off by removing high voltage from the klystrons, and the operating 

key was removed from the console. 

2. At least one member of the party that was entering the cell had to carry a gamma survey 

instrument and act as a surveyor for the party. 

3. All persons who were entering the cell had to wear a safety plug on the external clothing and 

return it on leaving. An exception was made for the Duty LINAC Operator who carried the 

LINAC console key. 

4. If the constant air monitor for the cell exceeded its trip level or if fissionable targets were in 

use, protective clothing and shoe covers had to be worn. 

5. A health physics surveyor had to be contacted immediately if significant smearable 

contamination was suspected due to fire, leakage of coolant, etc. 

6. The cell door had to be left open at all times during occupancy. 

7. Persons who were leaving an active cell had to check their hands, feet, and clothing for 

contamination as well as any irradiated items that were removed from the cell. 

2.2.4 Critical Assembly Facilities

A fenced area between the LINAC Complex (to the southwest) and TRIGA/Experimental Area 

Building #1 (to the northeast) was used to conduct many critical and subcritical assembly experiments 
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at the GA site. A January 24, 1961, document indicated three buildings within the fenced area that 

appear to have originally been set up to support both the MGCR and HTGR programs (GDC and H&N 

1961, p. 95). In addition to these two buildings, a fuel assembly building was also located in the area. 

(The facility was later known as the Experimental Critical Facility.) The LINAC facility was also used 

to conduct these criticality experiments. (At some point, the northernmost portion of the LINAC facility 

housed the HTGR critical assembly.) Over the years, several additional reactor fuel development 

programs have occupied the critical experiments buildings: 

 Nuclear Rocket or Thermal Tungsten Rocket 

 Modified HTGR 

 Advanced HTGR 

 Thermionic 

A paper attached to Isotope Committee Amendment #1101 indicated that the purpose of critical 

assembly experiments was to test the criticality safety of fuel designs to establish safe handling 

procedures (Richey, Lloyd, and Clayton 1965). A similar study (Russell 1967, p. 3) stated: 

A portion of this program [thermionic] consists of constructing subcritical assemblies of 

the critical facility fuel elements and measuring the neutron spectrum from these 

assemblies using the General Atomic Electron Linear Accelerator and time-of-flight 

techniques. These spectra will be used to aid in the understanding of the critical 

experiment performed. 

These fuel assemblies were either irradiated with accelerator-produced neutrons or with neutron 

triggers (e.g., Po-Be or Pu-Be sources). 

The design of the fuel and assembly was such that if an undesired criticality were to occur the fuel 

would either rapidly expand (Russell 1967) or the subcritical fuel beds would separate (Pound 1974, 

Attachment), which induced the neutron excursion to terminate. In the first case, a phenolic resin that 

fails at 140°C was used to bind fuel compacts, and conical springs would separate the fuel to 

subcritical conditions. In the latter, beds of fuel were brought together and attached 

electromagnetically. If a criticality occurred, the magnet would fail and one of the beds would slide 

away by gravity. 

A March 29, 1974, memorandum requested that the reactor building and associated fuel assembly 

buildings that made up the Experimental Critical Facility (also known as the TCF) be assigned for 

storage of Fort St. Vrain reload fuel elements (Pound 1974). 

2.2.5 Reactor Facilities

GA operated several reactors in the TRIGA Reactor Building (Building 21). The first reactor that was 

identified through data capture documents was the Critical Isotope Reactor, General Atomic (CIRGA). 

Although it is clear that fuel assemblies were constructed for the reactor in January 1958 (Ray 1958a), 

very little information was found to describe the reactor itself, and it is not clear if this reactor ever 

operated. The critical assembly operated under License CX-3 (Ray 1958a). Reactor Pit 1 was “used 

for subcritical experiments and other tests” (GA 1969, p. 18), and it is possible that this area was used 

with CIRGA. 

Three separate TRIGA reactors – Mark I, Mark F, and Mark III – were operated in Building 21, each 

under its own license – R-38, -67, and -100 respectively. The original building housed Mark I and was 

constructed in 1957. The building was later expanded east (1960) and then north (1966) to house the 

Mark F and III reactors, respectively. These reactors were all below-ground pool reactors (GA 1969, 

p. 13) and were designed to provide university training, provide a source of neutrons and/or gamma 
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radiation for research, aid in the investigation of the effects of radiation on materials, and produce 

radioisotopes. 

The Mark I was graphite reflected and operated routinely at steady-state thermal levels up to 250 kW 

and could also be operated in pulsed mode. The Mark F reactor (also known as the Advanced TRIGA 

Prototype and the FLAIR) also operated routinely at steady-state thermal power levels up to 1,500 kW 

and could be operated in pulsed mode, but it was water reflected. The Mark III reactor operated 

routinely at steady-state thermal levels up to 1,500 kW and was water reflected. 

The TRIGA core contained reactor fuel, dummy fuel assemblies, and one or more neutron source 

holders. As a thermal reactor, the TRIGA reactor depended on slow or thermal neutrons to 

perpetuate chain reactions in its fuel. Because the neutrons born in the fissioning of 235U atoms were 

too energetic to produce significant numbers of further fissions, a moderator was employed to slow 

down thermalized fast neutrons. In the TRIGA reactor, neutron thermalization was achieved by using 

zirconium hydride, water, and graphite as moderating materials. To reduce leakage of neutrons from 

the core, reflectors were installed. Figure 2-9 shows the TRIGA Reactor Facility. 

Figure 2-9. Plan view of TRIGA Reactor Facility (GA 1969). 

2.2.6 Science Laboratory

The Science Laboratory consists of three arc-shaped buildings that surround the laboratory center; 

they contain 150 individual laboratories and associated offices (Figure 2-10) where work in chemistry, 

experimental physics, electronics, metallurgical research, reactor physics, and thermoelectricity was 
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performed. These are two-story buildings in a ring around the central library and cafeteria buildings. 

Laboratory Buildings A, B, and C have 46,000, 49,000, and 41,200 ft2, respectively. 

Figure 2-10. Site map showing location of Science Laboratory (SNM 1963). 

These buildings are arranged and named in a counter-clockwise direction with Building A, then 

Building C, then Building B, beginning from the southernmost segment. The specific size of these 

buildings and the floor plans for Buildings A and C are indicated in Figures 2-11 and 2-12, 

respectively. Some room numbers might have changed over time. 

2.3 SOURCE TERM 

There are three main sources that describe the amounts and types of radioactive material that were 

handled at GA: federal and State of California licenses for the possession and use of radioactive 

materials; descriptions and reviews of proposed experiments or jobs that involved handling of 

radioactive material in the form of ICAs, WAs, and Radiological Work Permits (RWPs), which contain 

information on radionuclides, quantities, and recommended safety precautions for the described 

activity and inventory and material handling (accountability) records. 

The use of SNM was governed by AEC regulations and licenses, SNM-69 (Docket No. 70-72), and 

SNM-696 (Docket No. 70-734). Byproduct materials were governed under AEC License 
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Figure 2-11. Floor plan of Science Building A (SNM 1963). 

Figure 2-12. Floor plan of Science Building C (SNM 1963). 

No. 4-1611-11 (J-60), which appeared to have been transferred to State of California jurisdiction in 

later years. A request for amendment to the byproduct material license in December 1960 or later 

requested additional authorization to possess and use 1 MCi of fission products and to increase the 

amount of byproduct materials 3 to 83 to 100,000 Ci each and up to 500,000 Ci total (GDC and H&N 

1960). Another license, SNM-55, was identified in License Amendment #59 (SNM-69 Amendments 

41-62), but has not been reviewed. Attachment C provides a summary of licensing limits that are 

described in GA records. It appears that in the earliest years of operations, the site was licensed by 

the AEC to possess byproduct materials and SNM, and at some point the request to increase in the 

byproduct material limit and add fission products appears to have been accepted based on the 

California license summary from about 1973. The AEC did not license accelerator-produced material 

or naturally occurring material (except for source material, i.e., uranium and thorium). Information on 

early radioactivity limits for accelerator-produced or naturally occurring radioactive material and 
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source material has not been found, but these materials are included in the California license 

summary information. 

The GA radiological source term included uranium, thorium, plutonium, fission and activation 

products, strontium, and tritium, as well as sealed photon- and neutron-producing sources and 

radiation-producing equipment. Radioactive materials were received from off the site or created on 

the site in the reactor or with the LINAC. 

Uranium. Uranium in the form of metals, oxides, and carbides was used for GA fuel fabrication, 

reactor fuel, research studies in the hot cells, laboratories, and reactor areas. The typical amounts of 

uranium in use in any one area ranged from milligrams to 100s of kilograms. Occasional work with 

other uranium chemical forms, such as UF6 or uranyl nitrate, was also conducted. Uranium isotopic 

forms included depleted, normal (i.e., natural enrichment), and enriched (up to 93.5%), as well as 

232U, 233U and 236U. Uranium from recycling operations would have included relatively small activities 

of nonuranium isotopes, such as 99Tc, 237Np, 230Th, and 239Pu. 

As discussed in Section 2.1.1 (Building 37 or SV-A) 233U was used for fabrication of some of the 

HTGR fuel compacts that were incorporated into finished HTGR fuel elements. Uranium-233 was 

also used to conduct neutron yield studies at the LINAC (Bold 1967a, p. 114) and used for various 

research and development activities at various GA locations (see Attachment D). Site documentation 

(Bold 1965b, p. 242) indicates that a special glovebox designed for handling up to 100 g of 233U for a 

special experiment in one of the Science Building laboratories was inspected in January 1965. 

Uranium-233 is produced by the neutron irradiation of 232Th, which captures a neutron and becomes 

233Th, which decays to 233Pa, which decays to 233U. The 233U at GA was most likely from the Hanford 

Site (Parker 1965, pp. 3–7). 

In addition, 233U always contains the impurity 232U due to the neutron irradiation of 232Th. 

Correspondence related to the 233U program suggests that 232U impurity levels in the 233U at GA would 

have been 8 ppm (Parker 1965, pp. 3–7). The 232U impurity in the 233U results in elevated gamma and 

beta dose rates due to the ingrowth of 232U progeny (228Th and its progeny). In addition to the 

presence of 232U at GA as an impurity in 233U fuel material, site documentation indicates that 232U was 

being used for diffusion studies at the site. 

In October 1966, 232U was the first radionuclide that was handled in the HCF Plutonium Laboratory 

glovebox (Gurren 1966a). Documentation from about March or April 1967 indicates that 232U in 

solution was deposited onto disks and then the liquid was evaporated in gloveboxes at the HCF 

Plutonium Laboratory (Gurren 1967a and 1967b). According to Gurren (1967b), the discs were 

approximately 1 x 105

 cpm alpha and were used for diffusion studies. The reports also indicate that 

there were no contamination problems during the 232U work; however, representative sampling of 

debris from the HCF HLC in 1994 indicated that the concentration of 232U was higher than any other 

uranium isotope (Sills 1994). It appears likely that the 1994 debris sampling included materials that 

had been used in making the 232U discs from previous years. On the other hand, 232U would have 

also been present in the HCF HLC from examination of irradiated HTGR thorium-bearing fuel. 

(NOTE: The Heath report (Heath 1971) cover page says it might contain patentable material). 

Thorium. Thorium use and chemical forms were similar to uranium. The total mass of thorium that 

was used on the site was probably less than that of uranium, but the thorium activity in use in an area 

at a given time could have been greater or less than uranium activity. 

Plutonium. GA conducted the SNAP-15A program for AEC, which utilized a small thermoelectric 

generator that was powered by 238Pu. According to site documentation, over 20 SNAP 15A 

generators had been successfully tested by the end of September 1964 (GA 1964). GA also 
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documented the presence of plutonium on the site in October 1964 when the SNAP-15A Plutonium 

Handling Room was completed at the SV-B Building (Bold 1964a), but plutonium was probably 

handled elsewhere on the site earlier. The SNAP Plutonium Handling Room was later relocated to 

the Experimental Building concrete vault in October 1965 (Bold 1965a). 

Site documentation indicates that encapsulated PuN and or PuO2 were used in nuclear 

(thermoelectric) battery assemblies that were assembled in the SV-D Building (SNM-69 1963). 

Laboratories (e.g., Mound Laboratory or Harwell in the United Kingdom) supplied the double-walled, 

encapsulated-plutonium-containing fuel capsules in shipping containers. Receipt surveys of the entire 

outer surface were conducted and compared to an action level of 6 pCi (~10 dpm). Site 

documentation indicates that fuel capsules contained a nominal 0.18 g of 238Pu as nitride with 20% 

239Pu (no amounts were specified for 236/240/241/242Pu, but lesser amounts of these isotopes would be 

expected) (SNM-69 1963; GA 1971a). 

Plutonium use was also noted in the LINAC and laboratory areas, the reactor areas, and the 

Plutonium Handing Room at the HCF. On October 6, 1964, 82 239Pu foils that were sealed with a 

plastic coating were inspected on receipt in an absolute filtered glovebox at the LINAC (Bold 1964a). 

The foils were presumably going to be used in an experiment that involved the measurement of 

delayed gamma-ray intensities from the photofission of 239Pu and 235U foils that was temporarily 

discontinued the following month (Bold 1964b). After a receipt inspection in October 1968, plutoniumaluminum alloy fuel elements (jacketed rods) for a Euratom plutonium-fueled experiment were 

received at the LINAC for placement into stainless-steel pressure jackets and pressure testing (Bold 

1968a). The Euratom plutonium fueled experiment was completed on October 31, 1968, and the rods 

were placed into storage. Wipes that were taken on the fuel rods, pressure tank, and glovebox where 

the experiment was carried out “indicated no significant contamination” (Bold 1968b). In addition, 

authorization to use an 8.5-L plutonium nitrate solution that contained at least 4.6% 240Pu in a shielded 

container in a glovebox at the LINAC facility was granted in 1968 (SNM-69 1963). A December 1968 

report indicated that dry runs with a simulated plutonium nitrate solution at the LINAC were being 

performed (Bold 1969). 

Receipt surveys of plutonium fuel elements for use in the HTGR Critical Facility were documented in 

several reports beginning in September 1966 (Bedsaul 1966). Site documentation indicates that 

receipt survey precautions were undertaken to ensure that plutonium fuel elements (Bold 1966a) and 

thermoelectric sources (Bold 1964a) were not leaking. 

Table E-1 in Attachment E provides information on the types of work and radiological monitoring that 

were performed at the Plutonium Laboratory in the HCF. The work appears to have involved mostly 

unencapsulated fuel-grade plutonium in gloveboxes. The mention of beta as well as alpha activity 

measurements in the HCF Plutonium Laboratory (Table E-1, November 1968) seems to imply that the 

beta activity fraction was as high as or higher than the alpha activity fraction, which might have been 

due to the presence of 241Pu, which might have been significant (about 85%) in terms of activity in a 

239/240Pu source term, but its maximum energy beta emissions of 21 keV (similar to 3H beta energy of 

18.6 keV) would be hard to detect unless a windowless counter was used. Perhaps the photon 

emissions from the plutonium isotopes and 241Am, as well as the bremsstrahlung from the 241Pu beta 

emissions, were detected and attributed to beta emissions. There is no mention of 241Am in 

association with the plutonium source term, and for fresh source terms the 241Am activity would be 

insignificant, but by 5 years the 241Am ingrowth would account for approximately 8% of the total 

radioactivity in an unirradiated reactor fuel-grade source term. On the other hand, the beta 

contamination that was noted in the November 1968 survey in Table E-1 might have been due to 

contamination that was tracked into the Plutonium Laboratory from elsewhere within the HCF. 

The isotopic composition of the various types of plutonium that would have been encountered at GA is 

presented in Section 3.1.3. 
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Tritium. Little information has been found on the quantities and chemical forms of tritium that were 

used at GA, but site documentation indicates that a Texas Nuclear Corporation Neutron Generator 

console at the LINAC was being used with a “new 3 to 5 curie tritium target” in January 1963 (Bold 

1963a). In February 1965, it was noted that the Activation Analysis group would move into the new 

laboratory building (Experimental Area Building No. 1) and that the bioassay schedule was updated to 

include all personnel who worked in the area (Bold 1965c). The October 1965 monthly report for 

TRIGA and Activation Analysis Facilities (Cockle 1965a) noted that 25 used tritium targets that totaled 

111 Ci were transferred from GA for storage and ultimate disposal. The fusion group might have 

been using tritium targets on the order of 1 Ci. Tritides (particulate), and tritiated water vapor and 

tritiated gas (the latter compounds would be released during off-gassing of the tritium targets) were 

probably the main chemical forms on the site. 

Strontium. GA conducted research and development on thermoelectric generators that were 

powered by radioisotopes for space power applications for AEC. By the end of September 1964, a 

250-W 90Sr-powered thermoelectric generator had been investigated and test generators in the 4- to 

10-W range were being fabricated and tested (GA 1964). Strontium bioassays also indicate its 

presence on the site. Radioactive strontium could have been associated with the irradiated fuels or 

the 90Y production facility (about which little is known) as well as other programs. A 1971 HCF report 

indicates that a 90Sr source had been loaded into a generator housing and transferred to the SV-D 

Building (Bedsaul 1971). Specific information on chemical form or forms of 90Sr at GA is unavailable, 

but strontium that is associated with the irradiated fuels is likely to be in the form of an oxide and not 

titanate; however, if strontium was being used for the production of nuclear batteries (thermoelectric 

generators), it might have been in the form of SrTiO3 (CBWI 2006; OTA 1994; Grissom 1994). 

Fission Products. Irradiated fuel would have a variety of fission and activation products. GA

documentation mentions 134Cs, 

137Cs, and 182Ta as well as 60Co and 154Eu. It was noted that a 

molybdenum production facility was located on the site, although it is not clear if this referred 

specifically to radioactive molybdenum. Additional fission and activation products that are associated 

with intakes at the TRIGA reactors on the site can be obtained from the current revision of ORAUTOTIB-0054, Technical Information Bulletin: Fission and Activation Product Assignment for Internal 

Dose-Related Gross Beta and Gross Gamma Analyses (ORAUT 2007a). 

Activation Products. The TRIGA reactors and the LINAC were used to activate a variety of 

materials including soils, steel, cladding materials, uranium, and plutonium. A more detailed list of 

possible activation products at the LINAC is provided in Attachment F. 

Other Radioactive Materials. Sealed sources, including 226Ra, 

137Cs, Pu-Be, and Am-Be, were 

available on the site for irradiation or calibration work. There is mention of releases of noble gases 

and iodines from the reactors and use of 252Cf in a glovebox. GA was licensed to possess 50,000 Ci 

of 170/171Tm, but no documentation was found to describe its use or form. A possible use of 170/171Tm is 

in isotope-powered thermoelectric converters (Kay 1994). No other information has been found on 

this potentially very large activity source with low-energy beta emissions. There was also a mention 

of using unsealed sources of 226Ra. 

Radiation-Producing Devices. Reactors and a variety of particle accelerators and X-ray producing 

machines were used on the site. Some descriptions are in the Reactor, Critical Assembly, LINAC, 

and Laboratory sections of this document. External radiation from neutrons as well as photons and 

electrons could be generated on the site from the reactors and the LINAC, and from sources such as 

Pu-Be or Am-Be that were listed in GA documentation. Table 2-1 lists radiation-producing devices in 

the Science Laboratory. 
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Table 2-1. Radiation-producing equipment in the Science Laboratory as of April 1971. 

X-ray machines and/or location Manufacturer Voltage (kVp)

L-556 Phillips Electronics 5

L-556 Picker X ray Corporation 50

L-556 Picker X ray Corporation 40

Pulsed X-ray machines location Science B Field Emission Corporation Model 265 700

X-ray diffraction location

L-147 Phillips Electronics 60

L-147 Phillips Electronics 60

L-409 Phillips Electronics 60

Electron microscope location L-543 Hitachi 100

Electron microprobe location L-154 Applied Research Laboratory 50

Source Term in the Laboratory 

The determination of the amount and variety of radionuclides that were the source term for exposures 

in the Laboratory is a complex process due to changes in radionuclides over time as well as 

fluctuations in the amount of the SNM (235U, 233U, and plutonium) that appears to have been 

omnipresent in the Laboratory. The number of laboratories with the variety of activities that are 

associated with work in chemistry, experimental physics, electronics, metallurgical research, reactor 

physics, and thermoelectricity imply the potential for a large variety of radionuclides as well as variant 

quantities of these radionuclides in the Laboratory at any given time. Beyond these considerations, 

there is the potential to view individual laboratories as separate entities in the Laboratory, each with 

an individual source term. The alternative – viewing all the individual laboratories as one entity – has 

validity based on inventory changes alone but is most strongly supported by the conclusions of a1970 

evaluation of the Laboratory (Bold 1970a). There are also sources of information on specific 

radionuclides in the Laboratory, such as the example of annual throughput data (Attachment D), 

which is only a partial list of radionuclides that would have been in the Laboratory and other GA 

facilities (GA 1972). 

Because there is no compilation of all the radionuclides and their activities in the Laboratory at any 

point in time, several information sources were examined for data that contributed to defining the 

Laboratory source term. Because these information sources were not created to define source terms, 

but rather for other purposes, only portions are relevant for defining source term. The primary 

sources of this information are: 

 SNM throughput data for the Laboratory (Attachment D), 

 License information (Attachment C), 

 ICA data, and 

 Health physics reports. 

Isotope Committee Approval forms 

There was a requirement at GA that a request to use radionuclides be formally reviewed by a 

standing group that was referred to as the Isotope Committee. Approval from this committee was 

necessary to proceed with work. The approval forms generally included a reference to where the 

work was to be performed, the radionuclides to be involved, and a brief description of the physical 

form of the radionuclide. Some ICA summaries are included in Table 2-2 to indicate this level of 

information. The data were taken from about 6 months of 1962 and as such represent some of the 

activities that took place in these laboratories during this time. The variety and quantities of 

radionuclides in the Science Laboratories at other times might have been similar or not. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 41 of 143

Table 2-2. Summary information for selected ICA forms.

ICA Laboratory room Radionuclide(s) Date

326 L-343 Natural uranium 3E-2 Ci solution-nitrate

Cesium-137 3E-2 Ci solution-nitrate

Strontium-90 3E-4 Ci solution-nitrate

Natural thorium, 3E-3 Ci,solution-nitrate

4/10/1962

328 L-647

Metallurgy X-ray Laboratory

L-657

Graphite sample 3/22/1962

327 L-307

L-147

Zn-65

5E-3 Ci metal

4/12/1962

352 L-635

L-637,39

U-235, <1 mCi, pyrolytic coated

Tl-232, <1 mCi, (U-Th) C2 particles in graphite compacts

4/12/1962

333 L-540

L-637, -639

Unknown isotope, <100 mCi, irradiated graphite 4/9/1962

334 L-141 Any nuclides resulting from neutron irradiation of steel 

including Fe-55, Fe-59, Cr-51, Ni-63, Ni-65, Mn-56, Co-58, 

Co-60, Mo-93, Mo-99

The nuclide of initial interest is:

Fe-59,0.001 Ci, solid steel plate

I-131, 0.015 Ci, solid iodine crystals

4/23/1962

349 L-137 Fission products, 0.1 Ci, as produced in uranium carbide 5/22/1962

400 L-218

High temperature lab to 

3500oC

BeO irradiated and graphite, solid 

Ci to be determined in hot cell 

9/10/1962

402 L-425

L-427

L-429

U-238, wire 9/17/1962

406 L-554

L-556

L-558

L-439

L-441

U-238, UO2, UC2 up to 1,500 g

U-235, UO2, UC2 up to 300 g

Th-232, ThO2, ThC2 up to 1,500 g

9/25/1962

407 Unstated laboratory U-238, U-235, Th-232, 10 g 9/21/1962

412 L-343 Z = 3 to 93, 1 mCi 9/21/1962

413 L-151 Ir-192, 10 μCi 9/25/1962

411 L-554

L-556

L-558

L-439

L-441

U-235, U-233, Th-232 3/22/1962

320 L-236

L-143

Z = 32 to 83, solid or liquid, 0.1 Ci 

Z = 90 to 92, solid or liquid, 0.1 Ci

Pneumatic system between TRIGA and L-236

Micarta (a proprietary polycarbonate reinforced cotton 

material), polyethylene encapsulated)

3/26/1962

Health Physics Reports 

The level of information provided in the health physics reports varies considerably based on when and 

by whom they were written. Some reported items included discussion, radionuclides, and quantities; 

others were quite brief. Some of this information is summarized in Table 2-3.

Health physics reports also included occasional discussions of monitoring or exposure estimating 

methods. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 42 of 143

Table 2-3. Radionuclides in specific locations in the Laboratory. 

Location Source term information

L-540 150 g of normal U

L-441 Spill, 50,000 cpm α

L-403 MFPs

L-307 Cs-134, 10 mCi

L-441 Uranium fire, 127 g of U-235

L-236 Sr-89 plus Sr-90, 2.5 R/hr β+γ at 

16 in., contact 160 R/hr β+γ

L-218 Ta-182

L-407 Ce-144/Pr-144, 0.1 μCi

Science Building B service core K-85 release, 5 Ci 

A memorandum from May 22, 1963, described a fire in L-540 that involved uranium chips and 

turnings, which produced an estimated maximum air concentration of 4.4 × 10-7 μCi/ml (Bold 1963b). 

The memorandum went on to state that “although there were no air samplers in operation at the time,” 

the volume of the room was 114 m3 and therefore based on the estimated amount of uranium that was 

consumed in the fire the maximum air concentration estimated was 4.4 × 10-7 μCi/ml. The report went 

on to state that as the maximum permissible concentration (MPC) for natural uranium was 6 × 10-11

μCi/ml and the time-adjusted MPC based on the exposure time was 9.6 × 10-9 μCi/ml. 

This memorandum raises several questions, such as why there were no operating air samplers in 

adjacent areas of the Laboratory, because if there was a sampler in L-540 at the time it was either not 

functioning or not operating (L-540 is characterized in Bold 1970a as one of two laboratories that were 

designed to be most appropriate for handling the most significant radionuclides in the Science 

Laboratory). 

2.4 SAFETY 

The GA program broke ground in 1956. For December 15, 1960, through December 31, 1964, 

Table 2-4 lists the number of film badges that were distributed (biweekly before January 6, 1962; once 

a month thereafter), the number of pocket dosimeters that were distributed (two per employee), the 

number of dosimeters that were provided to visitors, the number of bioassays that were reported for a 

given period, the number of wipes that were counted (alpha and beta counts might have been 

counted separately), the number of air samples that were counted (sometimes first and second counts 

or alpha and beta counts are tabulated separately), and the number of people who were considered 

trained to handle radioactive material. 

Safety precautions were included in the ICAs and WAs. A review of the first 100 ICAs (there are at 

least several hundred more of these documents that have not yet been obtained) shows that the 

authorized user’s name is on the applications, but the names of associates were not always included. 

A review of the first 100 ICAs and claim file information resulted in identification of a single name that 

appeared in both information sets, although it should be noted that the current ICA capture is not 

complete. 

Formal health physics procedures were found for the early to mid-1970s (after the operational period); 

before then, policy documents appear to have governed site radiological activities. A discussion with 

Paul Maschka, a radiological control technician, who began work at GA in the mid-1960s and who still 

works there part time, indicated that early site radiological protection personnel were required to cycle 

through training at all areas where radiological activities occurred, before working unsupervised. It 

was estimated that this training took 6 months to complete. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 43 of 143

Table 2-4. Radiological monitoring from December 15, 1960, through December 31, 1964. 

Period

Film 

badgesa

Dosimeters 

reset and 

read

Visitor 

dosimeters

Bioassays 

processed 

and 

evaluated

Qualified to 

handle 

radioactive 

material

Wipe and 

air 

samples 

counted Wipesb

Air 

samplesc

12/1–31/1964 791 1,770 878 188 - 9,193 - -

11/1–30/1964 755 1,795 730 232 - 8,620 - -

10/1–31/1964 785 2,273 861 185 - - 3,088 6,057

9/1–30/1964 692 1,613 674 285 - 9,699 - -

8/1–31/1964 686 1,916 622 374 - 9,704 - -

No July report -

d

- - - - - - -

6/1–6/30/1964 880 2,320 868 199 - - 4,500 6,536

4/1–31/1964 810 1,705 750 254 - - 5,695 3,380

3/1–31/1964 692 1,831 484 222 - 6,134 - -

2/1–29/1964 644 1,990 617 262 - 6,341 - -

1/1–31/1964 718 2,259 504 274 - 7,574 - -

12/1–31–1963 873 1,825 636 210 - 6,358 - -

11/1–30/1963 867 2,049 731 257 - 6,736 - -

10/1–31/1963 879 1,901 824 165 - 8,258 - -

9/1–30/1963 819 1,992 593 (e) - - 2,222 4,463

8/1–31/1963 808 2,251 722 (e) - - 3,146 4,709

7/1–31/1963 746 2,035 600 (e) - - 2,692 3,400

6/1–30/1963 719 2,096 502 (e) - - 2,274 2,708

5/1–31/1963 783 1,796 636 (e) - 6,318 - -

4/1–30/1963 658 1,890 480 (e) - - 2,112 2,582

3/1–31/1963 629 2,063 563 (e) - - 2,084 2,322

2/1–28/1963 611 1,798 364 (f) 215 - 2,426 2,106

1/1–31/1963 563 2,187 519 (f) - - 2,760 2,040

12/1–31/1962 528 1,669 453 (e) - - 1,636 1,066

11/1–30/1962 527 1,619 562 (e) 203 - 3,084 888

10/1–31/1962 528 1,659 492 (e) 180 - 3,436 830

9/1–30/1962 472 1,322 477 (e) - - 3,474 662

8/1–31–1962 374 2,461 687 (e) - - 2,250 858

7/1–31/1962 336 1,684 501 (e) - - 2,720 5,080

6/1–30/1962 331 1,631 542 (e) - - 2,454 618

5/1–31/1962 330 1,322 484 (e) - - 3,000 700

3/16–4/30/1962 598 2,242 710 (e) 116 - 3,151 570

2/1–3/15/1962 308 1,340 372 (e) - - 2,335 370

1/15–2/15/1962 380 2,109 621 (e) - 2,536 - -

12/16/1961–

1/15/1962

474g1,757 680 (e) - 1,268 - -

8/16–9/15/1961 864g1,836 948 (e) - - - -

7/16–8/15/1961 558g2,334 984 (e) - - - -

6/16–7/15/1961 519g2,880 1,232 (e) - - - 17

5/16–6/15/1961 934g2,422 707 (e) - - - -

4/16–5/15/1961 528g1,247 684 (f) - - - 36

3/16–4/15/1961 541g1,280 640 (e) - - - -

2/16–3/15/1961 465g2,742 622 (e) - - - -

1/16–2/15/1961 556g1,657 - (e) - - - -

12/15/1960–

1/15/1961

546g1,300 - (e) - - - -

a. Film badges were biweekly before January 6, 1962 and monthly thereafter. 

b. Wipes sometimes include alpha and beta counts separately. 

c. Air samples are sometimes added in twice (or possibly more) as an initial count and a delayed count, or as an alpha count and as a 

beta count. 

d. No data found. 

e. Although the Health Physics Reports for these time frames do not indicate that bioassays were performed or the number of bioassay 

 analyses performed, there is documentation to indicate that some bioassays were performed during some of these times (GA 1956–

1963, GA 1963–1967, GA 1963–1970). 

f. The Health Physics Reports for these periods made references to the analysis of bioassay samples, although specific numbers were 

not called out. 

g. Monitoring was biweekly during this period, and monthly during later periods. 
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Engineering controls to reduce radiation exposure appear to have been present throughout the site. 

The HCF was designed with thick concrete and leaded glass shielding to minimize external dose to 

individuals who worked in the area. An evaluation of the laboratory in June 1970 compared the 

radiotoxicity of radionuclide uses and limits to the laboratories that contained them. The basis for this 

evaluation was an article published entitled “United States of America Standards Institute (USASI) 

Design Guide for Radioisotope Laboratory” (Bold 1970a). This article set guidelines for 

radiochemistry laboratories based on the four classifications of relative radiotoxicity – very high, high, 

moderate, and slight. Recommended maximum activities for each laboratory were given. The 

laboratories were classified as A, high-level laboratory; B, radioisotope laboratory; and C, good 

chemical laboratory. A description of what the four classes of isotopes entails is included in 

Attachment G. At GA, most were classified as Class 2 and Laboratory types B and C. More 

importantly, these references are necessary to be able to understand the discussion of the original 

description of the laboratories in the Science Building that follows (Bold 1970a): 

When the facilities of the Gulf General Atomic were originally constructed, two 

laboratories located in the Science Buildings, L-307 and L-540 were designed as 

radioisotope laboratories. They were marginal in design and probably fell somewhere 

between the USASI classification of a good chemical laboratory and a radioisotope 

laboratory. As the need for additional facilities to do radiochemistry work arose various 

chemical laboratories were modified by the addition of high efficiency filters on the 

chemistry hoods, shielding, and controlled access to allow for this work. Thus, some of 

these originally designed chemical laboratories were developed into what could be 

classed as relatively good radioisotope laboratories. 

Ventilation equipment with particulate filtering appears to have been used in many areas throughout 

the site to minimize airborne contaminants. However, reports of work activities that did not use 

appropriate ventilation are common, especially in the early years. Open doors would occasionally 

cause contamination to blow into a lobby or similarly undesirable place. Air sampler placement was 

not always ideal in relation to measuring what exposures (Bold 1963b). On the other hand, it was not 

unusual to use a portable air sampler or localized ventilation to measure or control some of the 

potentially higher hazard exposures. 

The use of personal protective equipment (gloves, coveralls, respirators, air hoods, etc.) and personal 

survey requirements were required based on the perceived hazard, and hot and cold change rooms 

were available in some areas. Much work occurred in the early years to keep the cold side cold. 

Available site documentation (Quimby 1966) indicates that supplied air hoods were required in the 

Fuel Fabrication Facility thorium process area by January 1966. The use of supplied air hoods was 

also noted at the HCF in 1968 (Cockle 1968a), but the use of supplied air hoods probably dates back 

before 1966. In June of 1971, GA received approval from AEC and State of California to use air-line 

respirators (hoods) for personnel in controlled areas where airborne concentrations averaged over 

40 hours in any 1 week exceeded the limits that were specified in Federal and State regulations (Bold 

1971a). 

Little information is available on directly measured contamination levels at GA areas, but numerous 

results of wipe tests are available for the 1960s. Area contamination appeared to be extensive in 

some areas, such as the Fuel Fabrication Facility, and more localized in other areas. Most 

contamination measurements were reported in terms of removable activity and adjusted to picocuries 

or picocuries per unit area. Site documentation indicates that the removable contamination limits for 

clean areas were 6 pCi alpha and 30 pCi beta based on wiping an area of approximately 413 cm2

using a 2-in. filter paper (SNM-69 1963, p. 36). These levels were twice the sensitivity of the counting 

equipment. For controlled areas, contamination limits were established that depended on the specific 

operations, but alpha contamination limits were not to exceed 500 cpm as measured with a portable 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 45 of 143

alpha counter that was equipped with a 60-cm2 detector on a large-area masslinn wipe of 

approximately 100 ft2. Site documentation indicates that contamination levels were expressed in 

terms of 100-cm2 areas by the 1972–1973 time frame (GEES 1972, p.65). 

GA utilized a dedicated decontamination crew throughout the operational period (Bold 1965d,e) and 

during much of the residual period (Bold 1977) to minimize contamination levels in controlled areas. It 

appears that contamination levels were high enough at the Fuel Fabrication Facility during the 

operational period to require decontamination of the floors twice every 24 hours during times of 

increased workload (Bold 1965d) and continuously during the residual period (Bold 1977). 

Workers were to be monitored or check themselves for contamination when leaving contaminated 

areas, but in the early years there were a number of reports of missed opportunities and 

instrumentation that was not working properly. GA designed and fabricated a number of its own 

instruments, and although its innovations were admirable, and probably very acceptable, there is a 

sense that calibrations of radiation instrumentation in the early years sometimes lagged. Alpha hand 

and foot counters were available in the earliest days, but they were frequently partially out of service 

(some detectors did not work) and detection thresholds are unavailable. 

Low-energy beta measurements did not appear to be a concern in the earliest years, but these 

radiations gained interest by the mid-1960s. An internal proportional counter (IPC) reportedly 

measured tritium at levels approximately 20 times higher than the low beta counter. During October 

1969, the highest tritium value from the IPC was reported as 32,000 pCi and a minimum value of 

<100 pCi. Because of tritium’s very low beta energy (0.0186 MeV maximum), quantitative 

measurement of tritium on wipes is very dubious, especially on samples in which significant selfabsorption or self-shielding can take place or contamination from other radionuclides might be 

present. 

Unreliable performance of the IPC reduced the tritium wipe and air sample counting throughput in 

November 1965 (Cockle 1965b), and it was noted that the tritium monitor had not detected significant 

activity to date. In January 1966, a cleanup of the bunker area took place and several areas were to 

be individually cleaned by Activation Analysis Facility personnel until clean. A preliminary procedure 

was written for interlock operations, target changes (to be performed in a “glovebox type” enclosure), 

and maintenance at the bunker. Tritium tritide air concentrations were measured as <1 × 10-10

μCi/cm3 during the January 1966 time frame. The June 1966 health physics report (Gurren 1966b) 

suggested a complete cleanup of the neutron generator bunker room before the next target change, 

investigation of a practical method to calibrate the tritium monitor, and adherence to the bioassay 

schedule. In January 1967, Eberline tritium film badges were distributed to all personnel in the facility, 

and additional badges were placed at various locations within the facility (Kesting 1967a) in an effort 

to determine whether it was feasible to use the film badges in lieu of bioassay samples. Results 

reported in February were all less than the 1,000 pCi detectable limit of the film (Kesting 1967b, 

p. 52). How these results might be used for dose reconstruction is unclear, but the next month it was 

noted that quarterly samples for tritium had been provided by all personnel on the March 1967 

schedule (Kesting 1967c). In June 1967, it was noted that the semiannual bioassay had been 

completed. 

Individual monitoring programs were part of the GA program, although bioassay before October 1963 

occurred in response to nose wipes and incidents and was not routine. Routine uranium urinalysis 

and some mixed fission product (MFP) and beta urinalyses were also added on then. Bioassay for 

tritium and strontium isotopes came a bit later. A review of the NIOSH-Office of Compensation 

Analysis and Support (OCAS) Claims Tracking System data indicates that external dosimetry was 

provided to workers who were directly involved in radiation work. Workers who were less likely to 

have elevated radiation exposures, such as a telephone operator, showed only sporadic individual 

external dosimetry. Radiation workers were provided with a film badge and two pocket dosimeters. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 46 of 143

The film badge could measure fast neutrons, and the dosimeter measured slow neutrons. It is not 

clear if the dosimeter results are included in the individual records. Some extremity badges were 

available, but it was common to substitute a less desirable wrist dosimeter for the unavailable ring 

dosimeters. According to the explanatory notes in the early reports of external dosimetry (e.g., p. 2 of 

both Landauer 1959 and Landauer 1961) a beta dose was reported only for betas of energy greater 

than 1.5 MeV. One report mentioned that the ratio of this lower energy beta to gamma dose might be 

on the order of 50. One surveyor’s anecdote on surveying while a worker adjusted an X-ray beam 

indicated that the worker had directed the beam to his forehead, and that perhaps head shields 

should be made available, which points out the difficulty of assessing nonuniform radiation doses, 

which were not unusual in some GA areas. 

Air sampling devices (including giraffe, portable, and continuous air monitors) were used on the site. 

Early on, complaints of a lack of these samplers and the equipment to count them were common. 

The number seemed to increase over time. Many air monitoring results (for individual, area, and 

environmental samples) were observed at GA as handwritten entries that began appearing in

technician logbooks back in the 1960s. Some limited environmental monitoring data were collected in 

the early years, but were not consolidated. For tritium monitoring at GA, Cockle (1965a) noted the 

following: 

Air Sampling: Air samples have been taken on a daily basis since October 18, 1965. 

Sampling has been specifically for suspected air-borne titanium tritide in the generator 

room and the console room. The present sampling is not for gaseous tritium or tritiated 

water vapor. Plans have been discussed to use the tritium monitor when opening the 

drift tube to sample for tritium gas. 

Area contamination measurements of floors and work areas are available in technician logbooks 

(which have not all been captured) and some of the more notable results appear in health physics 

reports. A table of instruments that were on hand and the desired instruments and how they would be 

used is in Attachment H. 

2.5 INCIDENTS 

The table in Attachment I of information on radiological incidents and accidents for 1960 to 1969 is 

based on information from GA documents; it is not a complete list of radiological incidents and 

accidents, but rather provides a general sense of these events at GA. It should not be construed from 

this summary that documentation of more significant incidents will not be found. 

2.6 PHYSICAL EXAMINATIONS – X-RAYS 

According to 1961 reports, “Special tests for employees working around BeO. Special radiation tests, 

X-rays and Spirograms,” ranged in number from 93 to 169 per month (Bethard 1961a,b,c,d). No other 

documentation on occupationally required medical X-ray examinations has been found. Current 

health physics personnel were unaware of a requirement for occupationally required X-ray 

examinations of BeO workers in the early years of operations; however, it was thought that workers in 

the respiratory protection program had chest X-ray examinations. Two claimants indicated that chest 

X-ray examinations were given either every 3 months or annually. A former GA nurse recalled that 

chest X-rays were required for beryllium workers and stated the X-ray examinations were conducted 

at La Jolla Radiology. To date no information on equipment and techniques has been found. Medical 

records that indicate if an X-ray examination was made might be available at La Jolla Radiology, but 

copies or originals of the X-ray films are believed to no longer exist. Many of the inactive medical 

records at GA have been destroyed in the past several years, but it is possible that medical files exist, 

either on or off the site, for some of the still-active workers. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 47 of 143

In the early years at GA, complete blood counts were used to assess potential biological detriment 

from external and internal radiation exposure. A checklist was used (Form 1-1) to document medical 

history, previous radiological exposure, assignment of a film badge or pocket dosimeter, attendance 

at radiological safety training, receipt of a controlled copy of the Radiological Safety Guide, etc. 

2.7 SUMMARY OPERATIONAL PERIOD ASSUMPTIONS, WORKDAYS, WORK HOURS, 

WORK CATEGORIES 

When bioassay data, film badge results, and other monitoring results are used to estimate internal 

and external exposures during the AEC operations, an estimate of the number of workdays or work 

hours per year is generally not important. However, whenever airborne concentrations are used for 

dose reconstruction the number of hours in a standard workweek needs to be known. The use of a 

56 hour MPC in GA site documentation (Bold 1963e, p. 62) and (Bold 1963f, p. 71) indicates that 

some employees worked 56 hours per week. The inhalation intake rate estimates for unmonitored 

workers presented in Tables 5-1 through 5-7, and in Tables 6-2 and 6-3, are based on a 10 hour 

workday for 250 workdays per year. If claimant files contain exposure time information, the intakes 

should be adjusted accordingly. 

No attempt has been made to sort workers into exposure categories. Individual dosimetry appears to 

be available for most workers. One unusual job category at GA was the custodian, who had the 

responsibility for surveying clean trash and decontaminating areas. In addition, it was noted that a 

number of different workers might have changed filters on hoods and exhaust systems, including the 

metallurgist and Administrative Services staff. 

2.8 RESIDUAL CONTAMINATION AND REMEDIATION PERIODS 

Documentation that was reviewed by NIOSH (2006c) indicates a potential for significant residual 

contamination outside the period of weapons-related production at GA. The residual contamination 

period was from January 1, 1970, through December 31, 1995. A DOE remediation period is listed by 

the DOE Office of Worker Advocacy as 1996 through 1999, and this document assumes it ran from 

January 1, 1996, through December 31, 1999. The residual and remediation periods are discussed 

further in Sections 5.0 and 6.0 respectively. 

3.0 ESTIMATING INTERNAL EXPOSURE

This section addresses internal exposure during the operational period from 1960 through 1969. 

Air monitoring was achieved with both fixed-head and portable air samplers, but a summary report 

from December 1960 stated that insufficient uranium air sampling was occurring in uranium process 

areas (Ray 1960a). Iodine-specific sampling in the facilities did not appear to be in place at startup, 

as indicated by a December 1960 health physics report notation that 11 mCi of 131I had been released 

through the stack of the HCF, but “the calibration of our iodine measurement is subject to question” 

(Ray 1960b), and by a statement by Ray (1960c) that iodine was not suitably detected by the GA April 

1960 methods of evaluating radioactive particulate concentrations (collection of sample on a filter and 

counting). The February 15 to March 16, 1961, progress report stated, “A new filter paper holder that 

incorporates a charcoal filter as well as a paper filter has been designed for use on the ’giraffe' air 

samplers. The charcoal filter is used to sample radioactive iodine. With a minor modification of the 

filter paper holders on the Continuous Air Monitors, the charcoal filters can be used in them” (Bold 

1961a). Fixed-head air samplers were often located near hoods and gloveboxes. Air samples 

included considerable activity from thoron (220Rn) progeny, and so required a couple of days’ wait 

before long-lived air concentration measurements were available. No explicit concentration 

measurements of thoron or its progeny have been found for 1960 through 1969. A later set of 

measurements from 1975 is not considered representative of 1960 through 1969. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 48 of 143

The initial focus of the bioassay program appeared to be uranium. Results were commonly reported 

in units of micro-microcuries, which is equivalent to picocuries. As noted in section 2.4, bioassay 

before October 1963 occurred only in response to nose wipes and incidents and was not routine. 

Records of urinalysis for fission and activation products are available as early as 1963. The early 

(1963) urinalysis procedures for uranium have in general been described. In December 1963, the HP 

Laboratory Report (Andrews 1964a) stated: 

The urinalysis procedure for uranium is being changed. The old procedure required 

that the entire volume of urine be boiled dry with an excess of HNO3. This resulted in 

excessive odor and noxious fumes. The new procedure follows the MFP scheme and 

precipitates the uranium. Most of the urine is then poured down the drain and the 

precipitate is redissolved and ashed with a minimum of nitric acid. The procedure is 

essentially unchanged from here on. Extraction of the uranium with ether, back 

extraction into H2O and drying in a planchet complete the sample preparation. The use 

of ether remains a safety hazard. 

The December 1963 HP Laboratory Report (Andrews 1964a) also reported that the Eberline MAC 

with a ZnS(Ag) phosphor would be used in lieu of an internal proportional counter to analyze urine 

samples. It was additionally noted in that report that the MAC ZnS (Ag) phosphor in contact with the 

sample would have essentially the same background and efficiency as the internal proportional 

counter, and so the MAC was available for counting the urine samples, which would reportedly speed 

up counting by about 500%. 

No records of thorium or plutonium bioassays have been found for 1960 through 1969. 

In vivo bioassay measurements were first performed on site in 1966 by Helgeson Nuclear Services 

(HNS) in its mobile counting van. Based on a review of available Helgeson reports, the van came to 

the site every 4 months (HNS 1967, HNS 1969a, HNS 1969b, HNS 1969c, GA 1970) . According to 

the site documentation (GA 1970) individuals were selected for counting based on their potential for 

exposure to beta-gamma emitters and 235U as well as on the results of their urinalyses results. Both 

total body counts and chest counts were performed. A review of some of the 1969 Helgeson reports 

indicates that 60Co, 58Co, 137Cs, 182Ta, 95Zr/Nb, 131I and 85Sr were occasionally detected (HNS 1969a, 

HNS 1969b, HNS 1969c). The focus of the early chest count was 235U; a program for detection of 

thorium was added in the 1970s (HNS 1975), which is after the operational period. An in-house 

counting system, based on a Health Physics Journal article design, including a 4- by 4-in. sodium 

iodide crystal and an office chair lined with lead was used when counting was desirable between 

Helgeson’s scheduled visits (GA 1966, GDC 1966-70, Tammany 1973). 

Both in vitro and in vivo bioassay data are available for GA employees. For uranium bioassay results, 

it is favorable to the claimant to assume 93.5% enrichment unless there is documentation that 

indicates a different enrichment. For bioassay results in the operational period that were reported as 

gross alpha, it is favorable to the claimant to assume that the gross alpha result is attributable to the 

alpha-emitting radionuclide (234U or 233U) that results in the greatest organ dose, unless there is 

documentation that indicates that the gross alpha result is representative of only one radionuclide (or 

radioelement). For bioassay results that were reported as gross beta, it is favorable to the claimant to 

assign radionuclide-specific intakes of mixed fission and activation products using the methods in 

ORAUT-OTIB-0054 (ORAUT 2007a), using 90Sr as the indicator radionuclide that document 

describes, as discussed in Section 3.3.6.2 of this document. 

In vitro data, specifically urinalyses for the 1960s and later exist at the site in three formats: laboratory 

records, transcribed results on individual cards, and an electronic database that was developed later 

to better assimilate individual results. It appears that the bioassay analyses were performed on the 

site. Some nasal swab results are available, apparently in response to incidents. Wound monitoring 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 49 of 143

to identify photon emitters was observed in at least one incident report. Some fecal samples were 

analyzed. 

The general codes for bioassay data are listed in Table 3-1.

Table 3-1. Bioassay codes [3]. 

PU, PA, and 

PX

Uranium urinalysis (results written as U-235 in urine can be assumed to be total uranium. 

These samples could either have been analyzed by fluorimetry or gross alpha counting. Any 

alpha emitter that made it through the chemical processing of the urine would be detected by 

the gross alpha counting methods, but the actual chemical recovery for nonuranium alpha 

emitters is unknown, so only uranium is assumed).

a

UX Whatever radionuclide is stated in the heading of the database column, these are samples 

that were sent to U.S. Testing.

PT Tritium.

PF Mixed fission and activation products.

PC Cesium (assumed to Cs-137 unless Cs-134 or another isotope is specifically stated).

PS Sr-90.

PI I-131.

a. The individual GA Internal Radiation Monitoring reports (in vitro section) typically include the summary heading: 

“Analysis Type: Gross Uranium Urinalysis (U-235).” 

3.1 IN VITRO DATA 

Most of the information in this section applies to urinalysis, but the documentation indicates that some 

fecal samples might have been analyzed at GA or by an offsite laboratory. No fecal sample results 

have been located in the files. 

Based on a review of urine sample calculation sheets (GDC 1965a; GDC 1965b), most individual 

urinalyses appear to be based on the collection of spot samples with results recorded as picocuries 

per 24 hours. GA normalized these results to a 24-hour sample on the basis of collection time; that is, 

they multiplied the results of these spot samples by a “yield factor” that was the ratio of 24 hours to 

the urine sample collection time to determine the 24-hour sample result. 

There are a number of bioassay history cards; these are available in two electronic image files. One 

set, Bioassay History Cards 00000 to 49999 (GA 1963–1967), was reviewed to better understand the 

format. The file numbers appear to be assigned to specific individuals (i.e., everyone was assigned a 

unique file number). This data set is organized in batches based on individual personnel (file number) 

at GA. The whole dataset contains results for about 1,000 people from November 25, 1963, to June 

18, 1970. There are several card formats as shown in examples below: 

Example 1 

Date μμc/24 hrs Analysis Reference

11-25-63 <1.1 α PUXXX

Gross α

Gross α PUXXX

U-235 PUXXX

MFP PFXXX

<13.0 β PFXXX

T

3

or H3PTXXX

Example 2 

Date pCi/24 hr % MPL Analysis Sample No.

<0.93 <2.3 α PUXXX

0.80 <1.1 α PUXXX
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Example 3 

Date pCi/24 hr 20 % MPL Reference

0.89 0.82 2.1 PUXXX

0.1 0.8 <2.1 PUXXX

0.4 0.6 <1.4 PUXXX

Name, sometimes Social Security Number (SS# or SSN), Department, Room Number, File # (which 

corresponds to each individual person), Date of sample collection, etc., were included on reports. The 

less-than symbol (<) was used starting in 1963 with the earliest samples for alpha results, but not 

many beta results have a < symbol. The heading that appears to be 20 means 2-sigma and refers to 

the error that was associated with the result. It is likely that the error was based primarily on the 

counting statistics. 

The code IH in the reference column could mean that either the sample was taken for industrial 

hygiene purposes rather than health physics purposes or, before about 1963, the sample was being 

analyzed by the GA Industrial Hygiene group. 

The bioassay data card that contains the data entry from the calculation above (GDC 1965a, p. 22, 

sample PU321) was identified in claim files and is shown below: 

Date μμc/24 hrs Analysis Reference

2-26-65 0.6 α PU321

Most GA dosimetry files also contain a database printout with the heading “INTERNAL RADIATION 

MONITORING REPORT” that contains the worker’s name, Social Security Number, birth date and 

sex. The format of the results is shown below with the results from the bioassay card and claim file 

above. 

Analysis Type: GROSS URANIUM URINALYSIS (U-235) 

DATE RESULT 2 STD DEV UNITS 

02/26/65 PU321 = 0.600 0.00 DPM/24HRS DPM/L = 0.425 

Note that the result entry begins with “PU321;” the “PU” entry, as noted in Table 3-1, signifies a total 

uranium result, so this result is assumed to be a gross alpha result. The favorable to claimant 

assessment for uranium, as described in 3.3.1, would be used for dose reconstruction. 

Based on a review of claimant files, there are a number of instances where the bioassay data card 

result was expressed in pCi/24 hr but was entered into the database with units of DPM/24HRS as in 

the example above. In addition, less-than values entered on bioassay data cards appear to have 

been entered as a 0.000 result in the database. The dose reconstructor should carefully review all of 

the internal dosimetry data for a claimant to ensure that the data supports the reported units (i.e., 

results that are reported in “DPM/24HRS should be 2.22 times greater than other results reported 

pCi/24HRS). If the claimant dosimetry data review indicates the same numerical values with differing 

units, the most primary data source should be used. For the purposes of this direction, the order of 

primacy of internal dosimetry data records is (1) urine sample calculation sheets and offsite laboratory 

reports, (2) bioassay data history cards, and (3) computer-generated summary reports. 

3.2 IN VIVO DATA 

The Helgeson Nuclear Services mobile counter was brought to the site about every 4 months 

beginning sometime in 1966. The perceived potential for internal exposure was the early basis for 

including a worker in the in vivo bioassay program. Many individual measurement spectra are 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 51 of 143

included with results that identify the measured radionuclide and quantity. Background and calibration 

measurements were also observed in the records from HNS (1966a, 1969a, 1969b, and 1976). 

In addition to the in vivo counting by Helgeson, site documentation indicates that GA conducted its 

own in-house internal dosimetry measurements beginning in 1966. These occurred in Room L-122, 

where the worker sat in an office chair lined with 0.25-in.-thick lead and a 4- by 4-in. NaI (Tl) detector 

was in a lap geometry, and appeared primarily to be in response to incidents (GA 1966, GDC 1966-

70, Tammany 1973). 

Site documentation indicates that workers with suspected intakes were sometimes sent off the site for 

evaluation. At least two workers in the 1960s with 137Cs intakes were sent to the University of 

California at Los Angeles. Later documentation from 1975 (Bold 1975) and (Jordan and Kesting 

1975) indicates that sometimes personnel who were involved in contamination incidents were sent to 

the Sacramento Municipal Utility District facility for whole-body counting by Helgeson. 

Whole body counts (sometimes noted in GA dosimetry records as “total body count”) are counts 

performed of a broader area of the body than that for the lung (chest) counts. The whole body counts 

are performed to detect higher energy gamma emitters (typically fission products), while the chest 

counts are performed to detect the lower energy x-rays (as is done to detect 235U) in the lung. 

The GA dosimetry record entries for in vivo count results typically appear as separate pages on the 

Internal Radiation Monitoring Report are headed either as “In-Vivo Total Body Count Gross Gamma 

Scan,” or “In-Vivo Lung Count (235U, natural Thorium, etc.). Both sets of results include the analytical 

result and the two sigma result for the count. Both sets of summary results typically include the 

individual count data in addition to the summary results (as appears on the Internal Radiation 

Monitoring Report). The individual count data sheets additionally typically contain the name, height, 

weight, date of count, time of count, count identifier, value of % MLPB for positive results, and 

individual results for the analyzed energy regions-of-interest for the particular count. The Helgeson 

raw gamma spectra are also typically included in the dosimetry record. The lung count results are 

presented in mass units (micrograms for 235U, milligrams for natural thorium), and the total body 

results are presented in activity units (nanocuries) [4]. 

By August 1970, whole-body counting was performed three times a year by Helgeson (GA 1970). 

Whole-body counting was performed on all individuals who might have become “internally 

contaminated with beta-gamma emitters.” The frequency of whole-body counting was determined by 

the job or exposure hazard index of particular personnel with the maximum frequency of three times 

per year. 

3.3 RADIOELEMENT-SPECIFIC INFORMATION 

3.3.1 Uranium 

3.3.1.1 General 

As noted in Section 1.0, it has been determined that it is not feasible to reconstruct internal radiation 

doses for GA employees from exposures to uranium before October 1963 (Leavitt 2007a). The 

uranium at GA included depleted, normal, and enriched uranium. In addition, 232U, 

233U, and 236U 

were used on the site. As discussed in Section 2.3, uranium as depleted, normal, or enriched 

uranium, or as 233U were used in various locations and at various times on the site; accordingly, if 

multiple work locations must be inferred for dose reconstruction, then the dose is reconstructed 

assuming the uranium is either highly enriched, or 233U/232U, whichever is favorable to the claimant. 

Assuming that the 233U fuel material at GA contained 232U impurity at 8 ppm as discussed in Section 

2.3, the 233U to 232U activity ratio would have been 55.7 to 1 [5]. The forms might have included foils, 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 52 of 143

capsules, solutions, metal, or graphite compacts. Recycled uranium would also have been 

encountered. In general, uranium was in the form of metals or oxides at GA, although scrap recovery 

operations could have resulted in other chemical forms. Table 3-2 lists half-lives and specific 

activities for uranium isotopes. 

Table 3-2. Uranium isotopic half-lives and specific activities [6].

Radionuclide Half-life (yr) Specific activity (pCi/μg)

U-232 72 22,000,000

U-233 160,000 9,800

U-234 240,000 6,300

U-235 700,000,000 2.2

U-236 23,000,000 65

U-238 45,000,000,000 0.34

Table 3-3 provides some factors that might be useful for interpreting uranium results. In general, 

when no information is provided on the enrichment or isotopic composition of a uranium result, it is 

favorable to claimants to assume that the material is either highly enriched uranium plus the recycled 

uranium contaminants, or 233U/232U with an activity ratio of 233U to 232U of 55.7, whichever is more 

favorable to the claimant. Where uranium urinalysis results are based on gross alpha counting, the 

uranium enrichment is not needed to account for the total uranium activity in a urine sample. Uranium 

mass measurements in urine should be interpreted as natural uranium. HEU mass measurements 

(e.g., pCi per μg) reported for in vivo chest counts are valid. 

Table 3-3. Uranium mixtures specific activity and isotopic fractions [7].

Highly enriched (93.5%)

Activity 

fraction

Mass 

fraction

Activity ratio 

to U-235

Mass ratio 

to U-235

U-234 0.96811 0.010606 32.64 0.0113

U-235 0.02966 0.934633 1 1

U-236 0.00197 0.002075 0.0664 0.00222

U-238 0.00026 0.052687 0.00877 0.05637

Useful factors

U (93.5%) pCi/μg 68.109

Uranium activity ratio to U-235 mass

U (93.5%) (pCi): U-235 (μg) 72.9

Natural uranium

Activity 

fraction

Mass 

fraction

Activity ratio 

to U-235

Mass ratio 

to U-235

U-234 0.4886 5.37E-05 21.4 0.00745

U-235 0.0228 7.20E-03 1 1

U-238 0.4886 9.93E-01 21.4 138

Useful factors

U-natural pCi/μg 0.683

Uranium activity ratio to U-235 mass

U (pCi): U-235 (μg) 94.8

Depleted

Activity 

fraction

Mass 

fraction

Activity ratio 

to U-235

Mass ratio 

to U-235

U-234 0.1546 1.00E-05 14.45 0.00502

U-235 0.0107 1.99E-03 1 1

U-236 0.0005 3.11E-06 0.0467 0.00156

U-238 0.8342 9.98E-01 78.0 501

Useful factors

DU pCi/μg 0.4021

Uranium activity ratio to U-235 mass

DU (pCi): U-235 (μg) 202
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 53 of 143

It does not appear that large quantities of pure 232U were used routinely and continuously by GA 

employees. As discussed in Section 2.3, during the operational period 232U in solution was being 

deposited onto disks and then evaporated in gloveboxes at the HCF Plutonium Laboratory (Gurren 

1967a,b). According to Gurren (1967b), the discs were approximately 1 x 105 cpm alpha and were 

used for diffusion studies. The latter reports also indicated that there were no contamination problems 

during the 232U glovebox work so, except where a 232U incident is identified for a claim in the 

operational period, it is reasonable to make the assumptions in the previous paragraph for dose 

reconstruction. 

Human and animal studies have indicated that oxides of uranium can be very insoluble in the lung 

(ICRP 1995), which indicates absorption type S (0.1% and 99.9% with clearance half-times on the 

order of 10 minutes and 7,000 days, respectively). Other in vitro dissolution studies of compounds 

that are found at uranium facilities have shown that oxides of uranium exhibit moderate solubility 

(Eidson 1994; Heffernan et al. 2001), which suggests absorption type M (10% and 90% with 

clearance half-times on the order of 10 minutes and 140 days, respectively). In vitro dissolution tests 

on oxides produced from uranium metal during depleted uranium armor penetrator tests have 

indicated multicomponent dissolution rates, with 25% of uranium dissolving with a half-time of less 

than or equal to 0.14 day and 75% dissolving with a half-time of 180 days. Because there was no 

specific information on the solubility of aerosols produced, this analysis assumed that both types M 

and S were available. The selection of absorption type should depend on the organ of interest. Dose 

reconstructions should assume International Commission on Radiological Protection (ICRP) 

Publication 66 default parameters for particle deposition (ICRP 1994). 

The scrap processing might have resulted in the formation of more soluble uranium compounds. GA

documents include occasional mentions of UF6 and uranyl nitrate, which are absorption type F, but 

the quantities of these chemicals were relatively small, and so, unless specific information indicates 

that type F exposure occurred, type M or S should be assumed. 

All of the uranium that was that was handled at GA should be assumed to have been composed of 

recycled uranium. An estimate of contaminants that might contribute the most to internal doses, 

based on a review of recycled uranium contaminants at the Hanford and Fernald sites, is listed in 

Table 3-4. The relative internal dose contributions from 99Tc, 

228Th, and 232Th were low enough to be 

ignored. In addition, the activity fractions assume that the uranium specific activity is based on 

depleted uranium, which increases the proportion of the contaminants by activity. The contaminant 

levels for depleted uranium overestimate the contaminants in uranium of normal enrichment by about 

40%. Plutonium is assumed to behave as absorption type M or type S material and neptunium as 

absorption type M. 

Table 3-4. Estimated recycled uranium activity 

fractions for internal dose reconstruction. 

Uranium Pu-239 Np-237

1 0.00246 0.00182

3.3.1.2 Uranium In-Vitro Analyses and Data 

Some uranium urinalysis results for 1958 were found, but a 1960 memorandum (Andrews 1960) 

reported the development of a uranium urinalysis procedure to count samples from 18 people who 

were involved in a December 3, 1959 incident. A reported detection threshold of 1 cph for the 0.5-cph 

background alpha counter and a 30% uranium recovery reportedly corresponded to a detection 

threshold of 30 μg/L for the samples. Two of the 18 samples exceeded the counter background, but 

none exceed 1 cph. The “weights” of the samples (50 to 100 mg/cm2) exceeded the values on their 

graphs of sample self-absorption correction factors. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 54 of 143

The Industrial Hygiene group activated the bioassay laboratory in early October 1960 and assumed 

responsibility for beryllium analyses from the Health Physics group. Analytical procedures for 

enriched uranium, normal uranium, and thorium were reportedly outlined that year. The 1960 Medical 

Department (Bethard 1960) also noted, "A laboratory technician ... was hired November 7, 1960, and 

trained in tentative uranium procedures," and 13 urine samples were analyzed for uranium (as of 

December 19, 1960). 

Fires in the Hot Suite of the Experimental Building on March 18 and October 31, 1960, resulted from 

spontaneous combustion of uranium zirconium hydride (Skeehan 1960); 24-hour urine samples were 

collected from involved personnel. The March 1960 health physics report states that after the March 

18, 1960, fire in the filter plenum of the Enriched Uranium Suite, 48-hour urine samples were collected 

and sent to U.S. Nuclear Corporation for analysis, and each was found to contain <1 μg/L of uranium 

(Ray 1960d), which would have been considered a very low sensitivity for that time by fluorimetric 

methods. However, if the uranium was highly enriched, the sensitivity in terms of activity would have 

been 68 pCi/L. In the October 1960 incident, 24-hour samples were collected, and might have been 

counted on the site. The sensitivity of the October urinalysis was given as 10 dpm (4.5 pCi) per 24-

hour sample and all results were less than this amount (Bold 1960a). 

The June 22, 1961, health physics report stated, “Receipt of a Jarrel-Ash fluorimeter has enabled us 

to establish calibration curves for fluorimetric determination of uranium” (Bethard 1961b). The 

September 18, 1961, report (Bethard 1961d) stated, “One urine sample analyzed for uranium 

contained less than our detectable amount by fluorimetric method (10 μg/L). Radioassay of the 

sample is in process” (Bethard 1961c). 

There might have been a few samples analyzed thereafter, but in September 1963 it was reported 

that the bioassay program for counting alpha and beta activity in urine had been set up, equipment 

had been ordered and “one set of samples had been run with apparent success” (Andrews 1963). 

The October 1963 health physics report included the first reporting of the number of bioassay samples 

analyzed for the period. 

In December 1963, the Health Physics Laboratory Report (Andrews 1964a) stated: 

The urinalysis procedure for uranium is being changed. The old procedure required 

that the entire volume of urine be boiled dry with an excess of HNO3. This resulted in 

excessive odor and noxious fumes. The new procedure follows the MFP scheme and 

precipitates the uranium. Most of the urine is then poured down the drain and the 

precipitate is redissolved and ashed with a minimum of nitric acid. The procedure is 

essentially unchanged from here on. Extraction of the uranium with ether, back 

extraction into H2O and drying in a planchet complete the sample preparation. The use 

of ether remains a safety hazard. 

Andrews (1964a) also reported that the Eberline MAC with a ZnS(Ag) phosphor in contact with the 

sample would have essentially the same background and efficiency as the internal proportional 

counter, and so the MAC was available for counting the urine samples, which would reportedly speed 

up counting by about 500%. It was noted that the timers on the MAC should be replaced so they 

could be set with more precision. 

The nominal minimum detectable activity for uranium urinalyses after September 1963 is assumed to 

be 1 dpm/24 hr for gross alpha counting, and 10 μg/L for fluorimetry analyses [8], (Bethard 1961d). 

According to an August 1970 GA internal radiation monitoring program description, urine sample 

specimens from individuals working alpha emitters, tritium and strontium were routinely analyzed 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 55 of 143

while personnel working with other radionuclides were monitored using whole body counting 

techniques (GA 1970): 

Bioassay for alpha emitters in urine is done using a gross alpha technique involving coprecipitation of alkaline earth phosphates followed by ether extraction. This procedure is 

simple and effective for use where enriched uranium is the principle isotope of interest. 

Plutonium and other alpha emitters will be carried along in this procedure. 

Gross alpha urine sampling frequency varied from weekly to annually depending on an individual’s job 

or “exposure hazard index” and samples were to be the first voiding after normal bed rest. 

3.3.1.3 Uranium In-Vivo Analyses and Data 

Early 235U lung (chest) counting results (1966 to about 1967) were considered "experimental" by HNS 

(1966a) or “not thoroughly refined” (HNS 1966b) and either were associated with some large 

uncertainty or assumed to be invalid. Workers were chosen for lung counts based on their potential 

for 235U exposure and the results of their uranium urinalyses. The early GA 235U measurements are 

indicators that a worker had the potential for 235U exposure. Helgeson (HNS 1976, p. 8) reported 235U 

chest count minimum sensitivities of 30 to 60 μg based on 20-minute counts. For some results, 2-

sigma counting errors (HNS 1969b) might be in the claimant data, and these could be used to 

estimate positive results (frequently defined as greater than 3-sigma). .As an alternative, 60 μg is a 

reasonable choice for the nominal lung (chest) detection sensitivity, unless there are reported specific 

data to use instead of the nominal value [9]. 

Uranium-235 chest counts were available for some workers beginning in 1966, but are not considered 

valid for dose reconstruction purposes until 1968 based on an entry on a whole-body monitoring data 

card (GDC 1966–1970, p. 69), which states that results are valid using a “new curve fitting technique.” 

In addition, a review of approximately 1,000 whole-body monitoring data cards (GDC 1966–1970) 

indicates that before January 1968, 235U chest counts were “not valid” (e.g. GDC 1966–1970, pp. 

564–567). The reason for the “not valid” determination in the pre-1968 counts is explained by HNS as 

constant bias in the 235U data on p. 45 of HNS 1967. By mid January 1968, HNS had solved the bias 

issue in counting for 235U (HNS 1968). If the phrase “not valid” or “not corrected for chest wall 

thickness” does not appear on the record of a positive in-vivo count performed after 1967, the 

reported result can be used in the dose reconstruction [10]. By August 1970, 235U at 25% of the lung 

burden could be detected using in-vivo counting (GA 1970). 

3.3.2 Thorium

Thorium in the form of oxides, carbides, and nitrates was documented at GA. One of the prime 

thorium uses was fabrication of thorium-uranium fuel. This document assumes that absorption types 

M and S were both on the site. Thorium appears to have been on the site before and throughout the 

operational period. No in vitro bioassay results specifically for thorium have been found in the GA

records; former employees indicate that thorium in vitro bioassays were not performed. In vivo

bioassays on 12 workers for thorium were performed in March 1966, but results were not included in 

the report because the measurements were experimental and were undergoing careful review before 

final results were provided (HNS 1966a). Thorium in vivo counting was not mentioned again until May 

1969, when HNS (1969a) observed no activity in one count. In October 1975, the thorium chest 

counts appeared to become more routine. 

There does not appear to be a defensible way to ratio thorium intakes to uranium intakes, because 

some tasks included pure thorium handling. Thorium source terms that were associated with internal 

exposures would have most commonly been associated with the Fuel Fabrication Facility, the scrap 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 56 of 143

handling facility, the laboratory buildings, the Experimental Building High Bay and HTGR critical 

assembly, and the waste handling areas. 

Air samples were measured in alpha counters with no differentiation between radionuclides. Gross 

alpha air concentrations in the Fuel Fabrication Facility thorium work areas ranged from 1.8 × 10-12 to 

9.6 × 10-9 μCi/ml alpha plus beta during parts of July, August, and October 1966. 

Air concentration controls were based on enriched uranium concentration limits, because GA

concluded that in most areas, although the thorium air concentration limits were lower, thorium was 

only a small component of the radioactive source term. This probably was true in most areas; 

however, there appear to be some exceptions. 

No bioassays specific for thorium exposures have been found for the GA covered operational period. 

In later years, chest counts were used to identify thorium intakes, but care would have to be taken to 

address the actual period of exposure. These later chest counts, when available, can be used to 

estimate intakes of thorium based on a worker’s individual history. 

Consideration was given to using later chest counts to estimate coworker exposures, but availability of 

records and their representativeness especially for earlier exposures was deemed questionable. 

As noted in Section 1.0, it has been determined that it is not feasible to reconstruct internal radiation 

doses for GA employees from exposures to thorium during the period of January 1, 1960, through 

December 31, 1969. 

3.3.3 Plutonium

There is insufficient information to characterize Pu in the breathing zone air for workers during the 

operational period. This section is provided to characterize the operational period Pu source term, 

since that information is needed to assist in developing intake estimates during the Residual and 

Remediation periods (Sections 5 and 6). 

Much of the plutonium that was used at GA was in the form of sealed sources, but unencapsulated 

plutonium was also used. Records indicate that reactor-grade plutonium and heat sources were 

handled as well as 239Pu foils (assumed to have the composition of weapons-grade plutonium), and 

plutonium that was associated with irradiated fuel examination at the HCF. In addition, 241Am would 

have been associated with some of these source terms. 

Attachment E contains a summary of the only available workplace monitoring data at GA during 

plutonium-specific work during the operational period. 

Chemical forms of plutonium at GA included metals, oxides, and nitrides. Table 3-5 contains the 

isotopic composition of plutonium by activity from the HCF based on the analysis of representative 

debris collected from the HLC in 1994 (Sills 1994). Table 3-6 contains the theoretical isotopic 

composition of unirradiated plutonium grades that were handled at GA. 

3.3.4 Strontium

It appears that strontium analyses involved chemically removing the strontium from the urine sample, 

counting the new sample, and counting again when the 90Y has started to grow back in to determine 

the activity of 90Sr. Strontium could have been associated with the irradiated fuels, or the 90Y 

production facility (about which little is known), as well as other programs. A 1971 HCF report 

indicates that a 90Sr source had been loaded into a generator housing and transferred to the SV-D 

Building (Bedsaul 1971). Specific information on chemical form or forms of 90Sr at GA is unavailable, 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 57 of 143

Table 3-5. HLC alpha sampling results.a

Radionuclide

Sample alpha 

concentration (μCi/g) Relative alpha fraction

U-232 4.50E-01 4.94E-02

U-233 3.80E-01 4.17E-02

U-234 7.90E-02 8.66E-03

U-235 8.00E-04 8.77E-05

U-236 1.20E-03 1.32E-04

U-238 5.00E-04 5.48E-05

Th-228 3.10E-01 3.40E-02

Th-229 2.40E-02 2.63E-03

Th-232 2.00E-03 2.19E-04

Pu-238 5.70E+00 6.25E-01

Pu-239/240 5.00E-01 5.48E-02

Pu-241 5.50E-04b6.03E-05

Pu-242 NDc

Am-241 7.10E-01 7.79E-02

Am-243 1.60E-02 1.75E-03

Cm-242 2.00E-03 2.19E-04

Cm-243 4.00E-03 4.39E-04

Cm-244 9.20E-01 1.01E-01

Cm-245 NDc

Cm-246 NDc

Cm-247 1.00E-03 1.10E-04

Cm-248 5.00E-04 5.48E-05

Cm-250 1.40E-02 1.54E-03

Cf-248 8.00E-04 8.77E-05

Cf-250 3.00E-04 3.29E-05

Cf-252 1.00E-03 1.10E-04

Total 9.12E+00 1.00E+00

a. Adapted from Sills (1994).

b. Based on 0.0022% alpha decay (which would have been detected by 

gross alpha air sample analysis) from 25 μCi/g of Pu-241 determined by 

beta analysis (Sills 1994). 

c. ND – not detected 

Table 3-6. Isotopic composition by weight and activity percent of unirradiated 

plutonium grades.a

Isotope

Heat source/thermoelectric 

generator

Unirradiated reactor 

grade fuel

Weapons grade

(Pu-239 foils)

Weight % Activity % Weight % Activity % Weight % Activity %

Pu-238 90.0 99.8 1.5 2.12 0.05 1.73

Pu-239 9.1 0.0365 58.1 0.297 93.6 11.8

Pu-240 0.6 0.0088 24.1 0.451 6 2.76

Pu-241 0.03 0.2 11.4 97.1 0.4 83.7

Pu-242 0.01 0.000003 4.9 0.016 0.05 0.0004

a. Adapted from DOE (2006). 

although strontium that was associated with the irradiated fuels is likely to be in the form of an oxide 

and not titanate; but, if strontium was being used for the production of nuclear batteries 

(thermoelectric generators), it might have been in the form of SrTiO3 (CBWI 2006; OTA 1994; 

Grissom 1994). If a thermoelectric generator cannot be ruled out as the source of strontium being 

assessed, the larger organ dose between the following options should be assigned: (1) all 90Sr (as 

material type S), or (2) strontium as part of an ORAUT-OTIB-0054 (ORAUT 2007a) evaluation, which 

is reactor-based and in which strontium is assessed as material type F. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 58 of 143

3.3.5 Tritium

The possibility of counting urine samples for tritium and maybe fission products “by using a plastic 

phosphor wet with urine” was mentioned in 1964 (Andrews 1964a). The March 1965 health physics 

report (Bold 1965c) noted that the bioassay schedule was updated to include all personnel working in 

the Activation Analysis area (it is not clear if bioassay was for tritium or perhaps other 

radionuclides)(Cockle 1965a). In October 1965, personnel using the Bunker Facility were sampled 

routinely for tritium, and the samples were analyzed by an outside service (Cockle 1965a). In 

November 1965, all personnel who were associated with the bunker were on bioassay for tritium 

(Cockle 1965b). 

As noted in Section 1.0, it has been determined that it is not feasible to reconstruct internal radiation 

doses from exposure to tritium before October 1965. 

By August 1970 urinalysis for tritium was performed using liquid scintillation (GA 1970). Routine urine 

sampling at 1-month intervals for tritium was performed for individuals who regularly worked with 

0.1 Ci or more of tritium. Individuals working with 10 Ci or more were to be sampled the following day. 

3.3.6 Mixed Fission and Activation Products

3.3.6.1 General 

Urine samples that contained mixed fission and activation products were likely placed in a container 

and gamma-counted, or evaporated to dryness and beta-counted. The radionuclides were typically 

not identified, and the results were reported in terms of gross activity or gross beta and adjusted to a 

24-hour time, if a 24-hour sample was not provided or analyzed. 

At some point, maybe in 1963, the urine samples were chemically processed and then beta-counted, 

but it is not known if the gamma-counted results were chemically processed or if results that were 

designated as MFPs or beta represent the same or different analysis types. 

Because of the variety of activities ongoing at GA, there is much uncertainty about which 

radionuclides to attribute these gross results. Long-lived beta/gamma emitters were typically not 

identified in the early GA source term, probably because the short-lived activity tended to dominate. 

Following a 134Cs hand contamination incident on April 23, 1964, in Sorrento Valley, urine was 

collected immediately and from every voiding for 48 hours after the incident. Graphs of the excretion 

rate as a function of time showed that most of the 134Cs showed up in the urine the first day. 

Subsequently, single specimens were collected once a week. Additional samples were collected 

1 day after the incident and showed little activity due to 134Cs, even though they were counted for 

1 hour on a 256-channel analyzer. “The rapid excretion of Cs-134 is probably due to the versene 

soak which complexed the cesium and got it quickly into the bloodstream” (Andrews 1964b). The 

intake was modeled as an ingestion intake. 

For this incident, urine specimens were wet-ashed with nitric acid and hydrogen peroxide; then the 

salts were transferred to a planchet and counted in contact with the 2- by 2-in. NaI(Tl) crystal for 

50 minutes. Background corrections were made, but no correction was made for 40K in the sample. 

Count rates for all but the first few samples were on the order of 10 cpm. All counts were compared 

to a 134Cs standard that was prepared by the Analytical Chemistry group. Decay corrections were 

made to correct the sample count to the time at excretion. The lowest urine activity for April 26, 1964, 

at 2100 hours to April 27 at 0800 hours was 54 pCi/24 hr and results ranged from 54 to 

1,000 pCi/24 hr. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 59 of 143

From reviews of bioassay cards and urine calculation sheets during the operational period, a nominal 

range for the minimum sensitivity of a gross beta result (without normalization to time or urine volume) 

would be 5 to 10 pCi, and the nominal minimum sensitivity for a mixed fission product result would be 

on the order of 70 pCi [11] (GDC 1965a, GDC 1965b, GA 1963-1967, GA 1963-1970). 

3.3.6.2 Applicability of ORAUT-OTIB-0054 to Experimental High-Temperature Gas-Cooled 

and Reduced-Enrichment Research and Test Reactor Fuels 

Five basic categories of nuclear reactors were considered in the development of ORAUT-OTIB-0054, 

Fission and Activation Product Assignment for Internal Dose-Related Gross Beta and Gross Gamma 

Analyses (ORAUT 2007a) to represent sources of mixed fission and light activation products that 

were encountered by workers at DOE and contractor sites. Table 5-1 of ORAUT-OTIB-0054, 

reproduced below, lists the specific reactors that were considered and their corresponding category 

(ORAUT 2007a). 

The fuels and reactors covered a range of enrichments from natural uranium to highly enriched 

uranium, fission spectra from highly thermalized to fast, and a variety of cladding and 

Reactor description Category

Hanford N-Reactor Plutonium production reactors

Hanford Single-Pass Reactors Plutonium production reactors

Fast Flux Test Facility Experimental sodium-cooled reactors

Advanced Test Reactor Advanced test reactors

TRIGA Reactor (aluminum-clad fuel) Research reactors

TRIGA Reactor (stainless-steel-clad fuel) Research reactors

Pressurized waster reactor Generic reactor

structural materials. Both 235U and 239Pu fuels were considered. The two TRIGA reactors at GA were 

among the reactors ORAUT-OTIB-0054 considered. 

Some of the operations at the HCF and Experimental Critical Facility involved irradiated, experimental 

HTGR fuels for a thorium fuel cycle and irradiated fuel for the Reduced-Enrichment Research and 

Test Reactor (RERTR) program (Section 5.2.9). Because the experimental HTGR and RERTR fuels 

are not specifically mentioned in ORAUT-OTIB-0054, this evaluation of the applicability of ORAUTOTIB-0054 to GA dose reconstructions is provided. The experimental RERTR fuel was a prototype 

based on that for TRIGA reactors and was essentially the same as the TRIGA reactor fuel that was 

considered in ORAUT-OTIB-0054 (ORAUT 2007a), as evidenced by information in Tables 2.3 and 2.4 

of GA (2002) and in Section 4.1 of GA (1994). The experimental RERTR fuel is therefore not 

considered further in this evaluation. 

The experimental thorium-bearing HTGR fuels are described in Tables 2.1 and 2.2 of GA (2002) and 

in Section 4.1 of GA (1994). A more detailed description of these fuels is provided in Section 7.3 of 

the second edition of Nuclear Chemical Engineering (Benedict, Pigford, and Levi 1981). The HTGR 

design employs two types of fuel, both in the physical form of microspheres. The microspheres are 

embedded in a graphite matrix to form fuel rods. One type of microsphere consists of enriched 

uranium carbide; the other type is thorium dioxide. 

It appears from Section 4.1 of GA (1994) that other, similar uranium or thorium compounds were 

being evaluated, but the overall fuel cycle is unchanged: initially the reactor is powered solely by 235U, 

with the fissile material transitioning from 235U to 233U as the 235U is consumed and 233U is bred from 

the thorium. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 60 of 143

It is not known if the experimental fuel samples that were handled at GA were irradiated for long 

enough for there to be significant breeding and fission of 233U, but this has no impact on the 

applicability ORAUT-OTIB-0054. The thermal fission yield curves for 235U and 233U do not differ to a 

significant degree, and differ less than those for 235U and 239Pu (for example, see Benedict, Pigford, 

and Levi 1981, Table 2.9 or Figure 2.12). The latter consideration is already addressed in ORAUTOTIB-0054, so the fact the irradiated fuel materials that were handled at GA might have included 

fission products from the fission of 233U does not proscribe the use of ORAUT-OTIB-0054 for 

estimating intakes for workers who were exposed to these materials. The mix of fission products in 

the irradiated HTGR fuel materials would not have differed to any appreciable extent from those that 

were considered in ORAUT-OTIB-0054. 

The fact that irradiated, experimental HTGR fuel materials were handled in specific facilities at GA

does not limit the applicability of ORAUT-OTIB-0054 for the evaluation of gross beta or gross gamma 

analyses. This is true regardless of the extent to which any contamination that is associated with the 

experimental HTGR fuels contributed to the overall mixed fission and activation product contamination 

in the facilities of concern. ORAUT-OTIB-0054 may be applied for these facilities in the same manner 

as it would be applied for other facilities where the mixed fission and light activation product source 

term stemmed from irradiated nuclear fuels. 

3.3.6.3 Internal Exposure Assessment of Mixed Fission and Activation Products via 

ORAUT-OTIB-0054 Guidance 

To determine chronic intakes from nonspecific radionuclide analyses, each urinalysis result should be 

multiplied by the fraction of activity from one of the appropriate indicator radionuclides in Table 7-1 or 

Table 7-2 of ORAUT-OTIB-0054 (ORAUT 2007a), which provides the urine activity ratio of the 

indicator radionuclide for use in intake analyses. Table 7-1 fractions should be used for beta-counted 

urine samples that have not been chemically processed. Table 7-2 should be used for beta-counted 

samples that were chemically processed to remove certain radionuclides. 

In the absence of any information about the urine sample processing at GA during the operational 

period, it is favorable to the claimant to assume that the analysis is strontium; this results in the largest 

strontium values on which other radionuclides are based. The fractions for this type of analysis are 

shown in Table 3-7 below. 

Table 3-7. Urine activity fraction for bioassay codes PF, MFP and β or gross beta.a 

Radionuclide Average fraction of Sr-90 activity in Sr urine sample

Time since last fuel irradiation 10 d 40 d 180 d 1y

Sr-90 0.045 0.061 0.23 0.75

a. Adapted from ORAUT 2007a, Table 7-1. 

b. The presence of radioiodides and naturally occurring beta emitters in urine will lead to an 

overestimate of intake. 

The decay time or time since the last fuel irradiation in Table 3-7 above refers to the mix of 

contamination to which workers would have been exposed rather than the age of the fuel itself. Table 

3-8 below provides additional guidance on the choice of the decay time for various activities at GA. 

Table 3-8. Fission and activation product decay times for GA facilities and/or activities.a

Activities GA Facility Decay time

Reactor operations, 

fuel examination

Bldg. 21 TRIGA reactors, Bldg. 23 HCF hot cells, Bldg. 2 Science 

labs, Bldg. 27 EA 1 labs, Bldg. 27-1 # 1 bunker, Bldg. 30 Linac 

complex, Bldg. 31 HTGR-TCF critical experimental facilities

10d

Fuel dissolution Bldg. 23 HCF hot cells 180d

Waste management Bldg. 25 Waste yard/Incinerators 1 y

a. Adapted from ORAUT 2007a, Table 5-12. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 61 of 143

3.3.7 Thoron

Specific monitoring and reporting of thoron (220Rn) were not done at GA until after the operational 

period. The earliest identified thoron measurements are October and November 1975 measurements 

of thoron concentrations that range from 41.4 to 126 dpm/L in the Thorium Pilot Plant Fume 

Scrubber 2 effluents during operations and airflows of 30 to 40 lpm (McDonald 1975). The 

background concentration was reported as <0.05 dpm/L when the Pilot Plant was not operating. 

Given the fairly routine references to contamination on air samples from thoron progeny in the fuel 

fabrication area and the lack of information about what activities took place at the Thorium Pilot Plant, 

it seems unreasonable to assume that these later Pilot Plant measurements are useful for estimating 

thoron concentrations during the operational period in areas where large quantities of thorium were 

stored and handled. However, the following calculations are provided to give a sense of the 

magnitude of dose for 2,000 hours of exposure to thoron at the reported levels. 

To determine exposure due to a specified thoron concentration (assumed to be 220Rn), the 

concentration C in picocuries per liter was converted to potential alpha energy concentration (PAEC) 

in units of working level (WL) using: 

PAEC = C × F ÷ (7.47 pCi/L per WL) (3-1) 

where the equilibrium factor F was taken as 0.8, a value recommended by NIOSH (2006b). Using the 

reported 1975 range of thoron concentrations above that were measured for the Pilot Plant, the 

exposure would range from 2.0 to 6.1 WL. Potential alpha energy exposure (PAEE) in units of 

working level month (WLM) was obtained from PAEC and the number of months of exposure M using: 

PAEE = PAEC × M (3-2) 

where a month is assumed to be 170 work hours, which results in 24 to 73 WLM/yr using the Pilot 

Plant 1975 values (background would be less than 0.29 WLMs). Using the factors of 7.61, 1,750, and 

2.90 rem/WLM for lung, ET1, and ET2 (NIOSH 2006b) results in dose ranges of 183 to 555, 42,000 to 

128,000, and 69.5 to 212 rem/yr, respectively. Background would be <0.22, <51, and <0.084 rem/yr, 

respectively, for thoron exposure to each organ. 

3.4 OTHER RADIONUCLIDES 

Although licensing documentation (see Attachment C) indicates that GA was licensed to possess 

radionuclides such as 170/171Tm, it is not clear whether GA personnel were ever exposed to 170/171Tm. 

Site documentation from the operational period at GA also indicates that 228Ra and 224Ra were offgassed during thorium heats in the thorium conversion furnace at the FFF (Bold 1967c, Bold 1967d, 

and Ray 1958b) which were not accounted for in the bioassay program, and it is not clear what the 

exposure potential was. 

For beta or gamma emitting radionuclides other than those discussed elsewhere in Section 3, 

bioassay results (particularly in vivo) specific to those radionuclides can be used for dose 

reconstruction. Alternatively, the favorable to claimant estimates in ORAUT-OTIB-0054 (ORAUT

2007a) can be used to assess organ dose due to beta emitters, as described in 3.3.6.3. 

NIOSH has determined, and the Secretary of the U.S. Department of Health and Human Services 

(DHHS) has concurred, that it is not feasible to reconstruct internal radiation doses from exposures to 

thorium for General Atomics (GA) employees during the period from January 1, 1960, through 

December 31, 1969 (Leavitt 2007). Consistent with that SEC determination relative to thorium 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 62 of 143

exposures during the operational period, no methodology is proposed for reconstructing doses due to 

any intakes of alpha emitting radionuclides other than those discussed elsewhere in Section 3. 

3.5 INTERNAL DOSIMETRY SUMMARY 

Because of the SEC determination (Leavitt 2007) that it is infeasible to adequately reconstruct internal 

dose during the AWE operational period covered by the SEC (January 1, 1960 through December 31, 

1969), dose estimates for this period are considered partial dose estimates. 

For workers who had uranium bioassay after September 1963, estimates of intakes can be made. 

The uranium urinalysis detection threshold is assumed to be 1 dpm/24 hr for gross alpha counting, 

and 10 μg/L for fluorimetry analyses [8], (Bethard 1961d). 

Uranium-235 chest counts were available for some workers beginning in 1966, but are not considered 

valid for dose reconstruction purposes until 1968 based on an entry on a whole-body monitoring data 

card (GDC 1966–1970, p. 69), which states that results are valid using a “new curve fitting technique.” 

If the phrase “not valid” or “not corrected for chest wall thickness” does not appear on the record of a 

positive in-vivo count performed after 1967, the reported result can be used in the dose reconstruction 

[10]. 

For bioassay results in the operational period that were reported as gross alpha, it is favorable to the 

claimant to assume that the gross alpha result is attributable to the alpha-emitting radionuclide (234U 

or 233U) that results in the greatest organ dose, unless there is documentation that indicates that the 

gross alpha result is representative of only one radionuclide (or radioelement). For bioassay results 

that were reported as gross beta, it is favorable to the claimant to assign radionuclide-specific intakes 

of mixed fission and activation products using the methods in ORAUT-OTIB-0054 (ORAUT 2007a), 

using 90Sr as the indicator radionuclide that document describes, as discussed in Section 3.3.6.3 of 

this document. 

For an individual with bioassay data and on a case-by-case basis using individual bioassay data, it 

might be possible to estimate intakes and associated doses for some periods and some 

radionuclides. Although NIOSH is unable to adequately estimate total internal exposures for 

members of the [SEC] class, internal dose due to intake of uranium can be reconstructed for 

exposures starting in October 1963, and tritium internal doses can be estimated after September 

1965. Because of uncertainties in the exposure periods, the source terms, and the analysis methods 

as documented in the SEC determination (Leavitt 2007), these estimates are to be considered partial 

dose estimates for the SEC period January 1, 1960, through December 31, 1969. 

4.0 ESTIMATING EXTERNAL EXPOSURE

External whole-body dosimetry results are available for monitored GA workers from January 1960 

through 1999. Beginning sometime before 1960, film badge dosimetry services were provided by 

R. S. Landauer, Jr. and Company; in July 1961, Radiation Detection Company replaced Landauer. 

Site documentation indicated that dosimetry services from Landauer resumed in 1965. Personnel 

who worked in areas where potential neutron exposure could occur were issued film badges sensitive 

to beta, gamma, and neutron radiation. Otherwise, personnel were issued film badges sensitive to 

only beta and gamma radiation. Beginning in 1981, film was replaced by thermoluminescent devices 

(TLDs) for monitoring external exposure. 

Some details on the Radiation Detection Company dosimeters are available from site documentation 

(i.e., reported nonpenetrating and penetrating doses) which implies that they were comparable to 

Landauer dosimeters. The nearby location of the Radiation Detection Company in Palo Alto, 

California, and faster turnaround time for results were the primary factors in the choice to use 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 63 of 143

Radiation Detection Company. Film badge detection thresholds for Landauer badges were reported 

in January 1961 as 10 mrem from gamma, 40 mrem from beta and 20 mrem from fast neutrons. 

Other documentation indicates that film badge detection thresholds in the early days were assumed to 

be as high as 50 mrem for gamma and 60 to 80 mrem for beta. A detection threshold of 40 mrem for 

photons has been recommended for this project (ORAUT 2006b). This document assumes a 

detection threshold of 40 mrem for gamma, beta, and neutrons. 

In addition to the film badge dosimeters, personnel used direct or indirect reading dosimeters as early 

as 1959, such as personal or pocket ionization chambers (PICs). Early on, the procedure required 

that two PICs be worn in addition to a film badge, later that was cut back to one, and later it was rare 

to use direct-reading dosimeters (only around low-energy X-rays). Wrist and/or ring dosimeters were 

also used in some operations, for instance during removal of irradiated samples from the reactor. 

Results from the dosimeters were logged by employees on a routine basis, but at this time it is not 

clear if these onsite readings are included in the individual dosimetry records. Through the early 

1960s, doses were reported only for betas of energy greater than 1.5 MeV, according to the 

explanatory notes in the early reports of external dosimetry (e.g., p. 2 of both Landauer 1959 and 

Landauer 1961). 

Many of the various direct- and indirect-reading pocket dosimeters from Landsverk Electrometer 

Company were used, including Landsverk Models L-49, L-50, L-65, and L-69. The Model L-49 was a 

direct-reading dosimeter responsive to thermal neutrons and gamma radiation up to 200 mrem. The 

Model L-50 was a direct-reading dosimeter responsive to gamma radiation up to 200 mrem. There 

was also a Model L-50 ring dosimeter that employed an indirect-reading chamber designed to 

measure hand exposures up to 2 R that might have been used at the GA site. The Model L-65 

indirect-reading pocket dosimeter was responsive to gamma radiation up to 200 mR (ORAU 2006). 

According to site documentation, Model L-69 dosimeter was used to measure total gamma plus 

thermal neutron dose (Ray 1959). 

Site documentation indicates that film badges, fast and thermal neutron detection equipment, and 

direct- and indirect-reading pocket dosimeters were irradiated to evaluate the radiation each device 

recorded in the 1959 to 1961 period. Beginning in early 1960, indium foils were incorporated into 

personnel security badges in case of a criticality event (Ray 1960e). By July 1961, all personnel film 

holders were equipped with a built-in threshold detector for neutron dose evaluations in the event of a 

criticality (Bold 1961b). As a consequence, the previous practice of routinely issuing Landsverk L-65 

and L-69 PICs in addition to a film badge was discontinued by August 1961 because thermal neutron 

exposures could be determined from the film badge by use of a cadmium filter over the gammasensitive film. Nevertheless, GA continued the practice of issuing Model L-49 self-reading dosimeters 

to visitors and selected workers throughout the 1960s. 

Site documentation from 1960 indicates that wrist badges were issued due to a lack of available finger 

rings (Ray 1960f). Correspondence from April 1966 indicates that finger ring exposures were 

expected to be at least 5 times greater than self-reading wrist dosimeter indications (Gurren 1966c). 

A California Division of Industrial Safety inspection of the X-ray equipment and LINAC included 

mention that X-ray and diffraction spectrograph operators were to wear finger tab badges for 3 weeks 

each year, presumably to provide some assurance that radiation exposures were not routinely 

exceeded (Bethard 1961e). 

The variety of external radiation sources that were present at the GA site resulted in a wide range of 

energies from penetrating and nonpenetrating radiations. Little information is available to quantify the 

precise distributions of photon and neutron energy ranges. For the purpose of expediting dose 

reconstruction, the assumption favorable to claimants is made that workers were exposed to photon 

energies from 30 to 250 keV and neutron energies from 0.1 to 2 MeV. Nonpenetrating dose is 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 64 of 143

assumed to consist primarily of electrons or of photons less than 30 keV. For consistent presentation, 

exposure or dose is reported as: 

 Penetrating, assumed to be associated with photons of energies 30 keV or greater, 

 Nonpenetrating, assumed to be associated with photons of energies less than 30 keV or with 

electrons, and 

 Neutron, assumed to be associated with neutrons of energies of 0.1 to 2 MeV. 

Because a considerable amount of work at GA occurred in gloveboxes and hoods, correction factors 

from OCAS-TIB-0010 (NIOSH 2005) should be used to determine the dose to organs in the lower 

abdomen for which the dosimeter measured dose may not be reasonably accurate unless it is clear 

that an individual’s exposure would not have occurred in this manner. 

If it is known that a worker only received external dose from working with plutonium, the guidance on 

correction factors provided in Section 4.2.3 of NIOSH 2007b should be used for dose reconstruction. 

Badge exchange frequency at GA in 1960 and 1961 was bimonthly (Bold 1962). The badge 

exchange frequency from 1962 through 1969 was monthly (Bold 1968c). From 1996 through 1999, 

the badge exchange was quarterly. The exchange frequency was probably monthly between 1969 

and 1996. 

A summary of GA personnel dosimetry by type, provider, period of use, exchange frequency, and 

minimum detection level (MDL) is provided in Table 4-1.

Table 4-1. Dosimeter type, period of use, exchange frequency, MDL, and potential annual missed 

dose. 

Dosimeter type/ 

provider Period of use

Exchange

frequency MDL (rem)a,b

Max annual missed 

dose (rem)c

Film badge/ 

Landauer

Pre-1960–July 1961 Bi-monthly (n=24) 0.010 gamma,

0.040 rem beta,

0.020 rem neutron

0.960

Film badge/ 

Radiation Detection 

Company

July 1961–1965 Monthly (n=12) Up to 0.050 rem 

gamma,

Up to 0.080 rem 

beta

0.480

Film badge/ 

Landauer

1965–1981 Monthly through 

1969, likely monthly 

through 1981 (n=12)

NA

d

0.480

Thermoluminescent 

dosimeter/NA

1981–present Likely monthly

through 1996,

Quarterly from 1996

NA 0.480

a. Reported MDLs as summarized in Section 4.0. 

b. MDLs assumed for gamma, beta, and neutron: 0.040 rem (ORAUT 2006b).

c. (Annual exchange quantity) × (0.040 rem). 

d. NA = not available. 

Very few individual monitoring data on either Landauer or Radiation Detection Company original 

reports have been found. External monitoring data for workers at GA were maintained on an IBM 

computer system as of January 7, 1961, and these records along with a few AEC Form 5 records in 

claimant files can be used to reconstruct external dose to workers at GA beginning January 7, 1961. 

Based on a review of claimant files the only record of external monitoring data before January 7, 

1961, are AEC Form 4 records with only the external, whole-body, penetrating dose that was received 

at GA from the beginning of initial monitoring to January 6, 1961. An approach that is favorable to 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 65 of 143

claimants is to assign the entire dose from before January 7, 1960, to 1960. This dose can be 

assigned as either gamma or neutron, whichever is more favorable to the claimant. In addition, 

available records do not always allow determination of whether an individual GA worker was 

monitored for gamma and beta exposure or gamma, beta, and neutron exposure. Blanks or zeroes in 

the dosimetry database do not necessarily relate to whether a person wore a neutron badge; as a 

consequence, missed neutron dose should be assigned when only summary records are available. 

Claimant files do not appear to contain any extremity monitoring data for January 1, 1960, through 

January 6, 1961, and film badge reports from Landauer from January 1961 through June 1961 

indicate that the finger ring doses were reported by Landauer in “arbitrary units.” During the use of 

two-element film badges, the term arbitrary units was used to report an open-window dose. This term 

was an acknowledgement that the open-window reading could not be accurately interpreted without 

knowledge of the workplace radiation type and energy (ORAUT 2006a). It is possible that the 

arbitrary units were entered as extremity doses in the database before July 1961; if so, these results 

should not be directly used to reconstruct dose. Radiation Detection Company badges were used 

beginning in July 1961, and reported extremity results were in millirem. 

Form AEC-190 summaries of GA-reported external penetrating doses for 1960 through 1964 and

1967 are listed in Table 4-2 and provide a general sense of the program size. The reported doses do 

not include nonpenetrating and extremity doses. These summaries, which were prepared for the 

Table 4-2. External penetrating dose summaries for 1960 through 1964. 

Year

Number of workers in each monitoring/dose category

Not monitored Monitored 0–1 rem 1–2 rem 2–3 rem 3–4 rem 4–5 rem 5–6 rem 6–7 rem

1960a1,389 209 194 4 6 0 4 0 1

1961b1,388 299 290 7 1 0 1 0 0

1962c1,500 341 336 4 1 0 0 0 0

1963d1,656 585 561 16 4 4 0 0 0

1964e1,559 610 580 22 6 2 0 0 0

1967

f

– 576 562 8 5 1 0 0 0

a. Hughey (1961).

b. Hughey (1962).

c. Jackson (1963).

d. Jackson (1964). 

e. Bird (1965).

f. Bold (1968d, p. 66). 

AEC Headquarter Health and Safety Division, are expected to include significant beta, photon and/or 

neutron occupational dose from GA accelerator operations because of the intermingling of radiation 

sources at GA. 

4.1 DOSE RECORD FORMAT 

This section discusses the external dose record format in claimant files for monitored GA personnel 

between January 7, 1961, and December 31, 1969 (the records before this are in summary form). 

Each page of external monitoring results contains a file number, employee name, Social Security 

Number, and birth date at the time of data entry. The birth date was used to calculate allowable 

lifetime dose as shown in the PERMISS column in accordance with the regulations in effect at the 

time. The interpretation of the remaining columns is based on correspondence from the current GA

Radiation Safety Officer. Table 4-3 lists the codes for the DOS columns. Each page contains only 

one DOS code; as a consequence, there are three different pages for each monitoring period that 

document (1) whole-body, (2) skin, and (3) extremity doses. For doses after the third quarter of 1961, 

each page contains the quarterly monitoring results for whole-body or skin or extremity doses. Even 

though the doses in the last quarter of 1961 could be presented as monthly, site documentation 

indicates that the badge exchange frequency was weekly in 1960 and 1961. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 66 of 143

Table 4-3. DOS column code.a [12] 

1 Whole body

2 Skin

3 Extremity

a. There is only one DOS designation per page. 

The DATE column contains either the month (01 through 12) or a 2-week monitoring period in 1961 

followed by the last digit of the year. From January 7, 1961, to January 5, 1962, the first two-digit 

numeral in the DATE column is interpreted according to Table 4-4, which presents the 2-week 

monitoring periods for 1961 according to site documentation (GA ca. 1961). A review of claimant files 

indicates that by the fourth quarter of 1961 monitoring data was reported quarterly rather than weekly, 

as listed in Table 4-4. After January 5, 1962, the first two-digit numeral in the DATE column 

designates the month with 3 months per quarter on each page. After a space, the last digit in the 

DATE column is the last digit of the year (i.e., 1961 through 1969 are designated by the numerals 1 

through 9). 

Table 4-4. Exposure periods for monitoring devices for 1961a (GA ca. 1961). 

1st quarter 3rd quarter

Period no. From Through Period no. From Through

01 1/07/61 1/20/61 14 7/08/61 7/21/61

02 1/21/61 2/3/61 15 7/22/61 8/04/61

03 2/04/61 2/17/61 16 8/05/61 8/18/61

04 2/18/61 3/3/61 17 8/19/61 9/1/61

05 3/4/61 3/17/61 18 9/2/61 9/15/61

06 3/18/61 3/31/61 19 9/16/61 9/29/61

07 4/1/61 4/14/61 20 9/30/61 10/13/61

2nd quarter 4th quarter

Period no. From Through Period no. From Through

08 or 1 4/15/61 4/28/61 21 10/14/61 10/27/61

09 or 2 4/29/61 5/12/61 22 10/28/61 11/10/61

10 or 3 5/13/61 5/26/61 23 11/11/61 11/24/61

11 or 4 5/26/61 6/9/61 24 11/25/61 12/8/61

12 or 5 6/10/61 6/23/61 25 12/9/61 12/22/61

13 or 6 6/24/61 7/07/61 26 12/23/61 1/05/62

a. 2-week periods from Saturday 0001 to Friday 2400. 

Table 4-5 lists the MON column codes. It is uncertain what the numeric codes 3, 4, 8, and 9 mean. 

Table 4-5. MON column codes [12].

0 TLD 

1 Film 

2 Ring

3 Unknown

4 Unknown

5 Unknown

6 Calculated dose

7 Not monitored (in the 1980s, a 7 was also used to indicate a dose 

received before monitoring at GA)

8 Unknown

9 Unsure, but there is some indication that 9 was used to indicate mixed 

information such as recorded dose plus an estimated dose result 

The GAMMA, BETA, and NEUTRON columns indicate dose in rem. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 67 of 143

4.2 FILM BADGE CORRECTION FACTORS 

Because very little neutron spectral information is available, this analysis assumed that reported 

neutron doses were calculated using the quality factors (QFs) from National Committee on 

Radiological Protection and Measurements (NCRP) Report 38 (NCRP 1971). The Report 38 QFs 

were averaged over the ICRP Publication 60 (ICRP 1991) energy groups. A dose multiplier of 2.0 

(actually 1.91 in the technical information bulletin ORAUT-OTIB-0055, Technical Basis for Conversion 

from NCRP Report 38 Neutron Quality Factors to ICRP Publication 60 Radiation Weighting Factors 

for Respective IREP Input Neutron Energy Ranges (ORAUT 2006d)) is used for conversion of the 

NCRP Report 38 dose equivalent to the ICRP Publication 60 equivalent dose for the 0.1- to 2-MeV 

energy range (ORAUT 2006a). 

According to GA site documentation, the nuclear track emulsion, type A (NTA) Landauer badges that 

were used at GA in the 1960s had a neutron energy threshold of 1 MeV. Other documentation 

indicates that NTA film was a poor detector of neutron energies below 500 to 800 keV (ORAUT 

2006a). Table 4-6 contains neutron dose correction factors for NTA film that was developed for dose 

reconstruction at Argonne National Laboratory-East (ANL-E) to account for <1-MeV neutrons. The 

facilities and activities at ANL-E appear to encompass the activities and facilities at GA. For NTA film: 

an unmonitored neutron dose correction factor of 1.5 is recommended for GA due to the presence of 

the LINAC, uranium-fueled critical assemblies, and the TRIGA reactors; if work handling plutonium is 

indicated (e.g., certain hot cell work), an unmonitored neutron dose correction factor of 4 is 

recommended. As a consequence, the recorded neutron doses from film for personnel at GA should 

be multiplied by a factor of 3 or 8 (as indicated by work location) to account for changes in QFs and 

NTA film energy threshold. 

Table 4-6. Neutron dose correction factors.a 

Facility

NTA film neutron 

dose correction 

factorsa

ICRP 38 to ICRP 

60 adjustment 

factor

Total neutron 

dose correction 

factor

Accelerators 1.5 2 3

Reactors and critical assemblies 1.5 2 3

Pu handling facilities 4 2 8

a. Adapted from ORAUT (2006a). 

GA site documentation also indicates that the Landauer badges used in the early 1960s at GA had a 

beta energy reporting threshold on the order of 1.5 MeV (Landauer 1959, Landauer 1961), which 

would have resulted in an underreporting of shallow dose equivalent to lower energy beta radiation. 

An estimated fraction (0.49) of reported dose has been calculated for similar ANL-E two-element 

badge designs that were used from 1945 to 1959 (ORAUT 2006a). As such, for exposures to MFPs, 

which contain mixed photon and beta radiation, an upper bound to the recorded nonpenetrating dose 

from film badge data can be estimated by multiplying the beta dose by 2.04 based on Table 6-19 of 

ORAUT (2006a), which indicates fraction of reported dose as 0.49. However, if a GA employee 

worked in the early 1960s with a source term that included primarily beta-emitting radionuclides with a 

maximum beta energy of less than 1.5 MeV (e.g., 170/171Tm), the personal beta dose would not have 

been accurately reported. 

4.3 OCCUPATIONALLY REQUIRED MEDICAL X-RAYS 

According to 1961 correspondence, “Special tests for Employees Working around BeO. Special 

Radiation Tests, X-rays and Spirograms,” were conducted for employees who worked around BeO 

(Bethard 1961a,b,c,d). No other definitive information was found that indicated the type or frequency 

of occupationally required medical X-ray examinations, although an assumption of an annual chest 

X-ray examination appears to be a reasonable minimum, and some workers (probably BeO workers) 

indicated that chest X-ray examinations occurred quarterly. The type of X-ray should be based on 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 68 of 143

current ORAU Team guidance. Organ doses can be obtained from the current revision of ORAUTOTIB-0006, Technical Information Bulletin: Dose Reconstruction from Occupationally Related 

Diagnostic X-Ray Procedures (ORAUT 2005). There is no documentation that indicates that 

photofluorography was performed on workers at GA. 

4.4 MISCELLANEOUS INFORMATION ON EXTERNAL EXPOSURE 

This section includes external dose information that could be of value for specific dose 

reconstructions. This analysis did not consider such information generically because of its limited 

applicability or because of limited details. 

Site documentation from May 1968 on an extremity overexposure at the LINAC indicates that even 

though the modulator high voltage and RF were turned off following each tuning and during sample 

changes, there was still an emission of <1 mA of direct current at 56-kV accelerating potential 

because of a malfunctioning electron gun (Cockle 1968b). A subsequent dose reconstruction using 

badges and finger rings indicated that the X-ray energy was approximately 20 keV and that the actual 

exposure was 7.700 rem, in comparison to the 7.650 rem gamma plus 35.870 rad beta Landauer 

originally determined. Procedures were changed at the LINAC to prevent reoccurrence. 

Employees such as those working with or in the areas of the accelerators and the reactor target 

irradiation areas might have received nonuniform radiation exposures, and no factors are available to 

adjust these exposures. It was also noted that plugs were occasionally removed from the HCF walls, 

which created beams of radiation. 

4.5 EXTERNAL DOSIMETRY SUMMARY 

Because of the SEC determination (Leavitt 2007) that it is infeasible to adequately reconstruct internal 

dose during the AWE operational period covered by the SEC (January 1, 1960 through December 

31), dose estimates for this period are considered partial dose estimates. External whole-body doses 

from penetrating photons can be determined from individual dosimetry records. Neutron doses can 

be estimated by adjusting the reported neutron dose by the appropriate factors in Table 4-6. For 

exposures to MFPs, the recorded nonpenetrating dose from film can be conservatively estimated by 

multiplying the beta dose by 2.04 based on guidance that indicates the fraction of reported dose as 

0.49 (ORAUT 2006a, Table 6-19). The assumption of initial, annual, and termination chest X-rays 

appears warranted. 

Correction factors for nonuniform radiation doses that might have occurred in the accelerator areas, 

the reactor target handling areas, or the HCF are not available. 

Extremity dose records are not available until about 1961; initially results were reported in arbitrary 

units. 

Sufficiently accurate nonpenetrating dose estimates cannot be made for workers who might have 

encountered significant quantities for extended periods of beta emitters with energies between 0.03 

and 1.5 MeV, which were not mixed fission or activation products. 

5.0 ESTIMATION OF EXPOSURE FROM RESIDUAL RADIOACTIVITY

5.1 INTRODUCTION 

GA continued to operate after the AWE operational period ended on December 31, 1969. NIOSH 

(2006c) noted that a significant potential for residual contamination remained after 1969. NIOSH has 

determined the residual radiation period to be from January 1, 1970 through December 31, 1995, and 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 69 of 143

the DOE Remediation Period to be from January 1, 1996 through December 31, 1999 (NIOSH 

2006c). During operations, airborne contamination can be caused directly by operations with 

radioactive materials as well as by the resuspension of surface contamination. This airborne 

contamination can be removed by settling on floors or other surfaces, or by ventilation (both 

engineered and natural building ventilation). The surface contamination levels can be increased 

directly by radioactive material processing and handling or by the settling of airborne contamination. 

This level is decreased by various incidental means such as general housekeeping or tracking of 

contamination out of the area on equipment or personnel. 

5.2 ESTIMATION OF INTERNAL EXPOSURE FROM RESIDUAL RADIOACTIVITY 

5.2.1 Introduction

The residual source term at GA was identical to the operational period source term that is discussed 

in Sections 2.3 and 3.1.1 through 3.1.7. It is difficult to estimate exposures to residual activity from 

workplace measurements, because non-covered operations continued during much of the residual 

period; however, it is reasonable to estimate exposures to residual contamination from long-lived 

radionuclides after 1969 by assuming that the residual contamination and or airborne radioactivity 

levels beginning in 1970 were equivalent to those that GA measured in December of 1969 or earlier 

depending on workplace measurement availability. Workplace measurements of surface 

contamination and or general area airborne contamination (non breathing zone air measurements) 

have been used to bound residual period intakes, because there are insufficient data of worker 

breathing zones for thorium (SEC determination) and other radionuclides. 

Two different methods have been used to estimate bounding intakes for the residual period. The first 

method utilizes surface contamination data if it was available. For controlled areas, the workplace 

was assumed to be uniformly contaminated at the maximum operational period contamination level 

within 1 year from the beginning of the residual period. The maximum level was then assumed to 

decrease exponentially to a level that was measured some time later in the residual period as shown 

below [13]: 

Ct = Co × e

-λt (5-1)

where: 

Ct = contamination level at time t measured during residual period 

t = time in years between contamination measurements 

Co = contamination level measured at time zero during the operational period 

e = base of the natural logarithms 

λ = exponential constant 

After solving for the exponential constant λ, the daily intake rate was calculated for each year of the 

residual period (1970 through 1995) by multiplying the contamination level by a resuspension factor of 

1 × 10-6/m (Abu-Eid et al. 2002) and applying a breathing rate of 1.2 m3/hr for 10 hours a day for 250 

workdays per year. This method was used to estimate thorium and uranium intakes (Table 5-1) 

during the residual period at the GA Fuel Fabrication Facility. For tritium intakes, the highest level of 

3H contamination at the Activation Analysis Facility in 1968 was 6,000 pCi, but no residual period 

contamination level was available; as a consequence, the tritium radioactive decay constant was 

assigned for the value of the exponential constant (Table 5-1). 

For uncontrolled areas, the GA removable contamination limits for clean areas (6 pCi alpha and 

30 pCi beta over 400 cm2) (Section 2.4) were used to derive intakes assuming a resuspension factor 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 70 of 143

of 1 × 10-6/m (Abu-Eid et al. 2002; ORAUT 2008) and applying a breathing rate of 1.2 m3/hr for 10 

hours a day for 250 workdays per year (Table 5-2). 

Note that the controlled area inhalation intake rates (bounding) in Table 5-1 are for areas other than 

the GA HCF during the residual period. For 1993, 1994, and 1995, the controlled area intake rates for 

thorium and uranium were lower than the uncontrolled area intake rates due to the exponential 

removal of radioactivity. As a consequence, the uncontrolled area intake rates have been be used as 

a lower limit of intake in 1993, 1994, and 1995. 

The second method used to estimate bounding intakes for the residual period utilized general area 

airborne monitoring data from 1968 (gross alpha and gross beta) in the GA HCF along with debris 

sample analytical data from the High Level Cell obtained in 1994. Based on airborne measurements 

of plutonium made later in the residual period, the residual period plutonium and other HCF 

radionuclide intakes are assumed to have remained constant throughout the residual period. The 

methodology for calculating intakes during the residual period is discussed in detail in the sections 

below. 

5.2.2 Thorium

The Fuel Fabrication Facility (SV-A) at GA used both thorium and uranium in the HTGR fuel 

fabrication process. Based on the frequency of decontamination crew activities at the Fuel 

Fabrication Facility (Section 2.4), it appears that the highest (bounding) intakes for uranium and 

thorium at GA would have occurred at this facility. The highest gross alpha, large area, masslinn wipe 

result was reported near the beginning of 1970 (in the Main Particle Area) as 70,000 cpm/ft2 as 

measured with a PAC-1S (Bold 1970b). A detection efficiency of 18% (Eberline 1977; Bold 1968c) 

results in the calculation of 419 dpm/cm2. A survey of the SV-A control area in March 1976 indicates 

a maximum contamination level of 30 dpm/cm2 (Bold 1976a). Assuming that all of the contamination 

was due to thorium, an exponential constant of 0.422 per year was determined. The daily inhalation 

intake rates for each year during the residual period in controlled areas are shown in Table 5-1. The 

intake rates in the thorium column of Table 5-1 would be applicable to the sum of the 232Th and 228Th 

activities. Dose reconstructing methodologies are presented for apportioning 232Th/228Th activities 

[14]. 

Thorium intakes during the residual period in uncontrolled (clean) areas of the facility are based on 

constant resuspension of the clean area alpha contamination limit for removable contamination (6 pCi 

alpha over an area of approximately 400 cm2) (Section 2.4). The resultant intake would be 1.23 × 10-3

pCi/d from 1970 through 1995 as shown in Table 5-2. This intake rate would be applicable to the sum 

of the 232Th and 228Th activities. Available worker data and information should be used to determine 

the likelihood of whether or not a worker would have entered controlled areas. 

If claimant information specifically indicates that work was performed involving HTGR fuel fabrication 

Building 37 or 39, either the thorium or uranium intakes in Table 5-1 (whichever is favorable to the 

claimant) should be used to reconstruct dose. While evaluating uranium intakes the dose 

reconstructors should assign uranium intakes from Table 5-1 by assuming that the material is either 

highly enriched uranium or 233U/232U with an activity ratio of 233U to 232U of 55.7, whichever is favorable 

to the claimant. If claimant information specifically indicates that work was associated with the GA 

HCF, the thorium and associated intakes in Tables 5-3 and 5-4 should be assigned. If claimant 

information specifically indicates that work involving thorium was performed at multiple or unspecified 

locations, the thorium intakes in Table 5-5 should be assigned. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 71 of 143

5.2.3 Uranium

During the fabrication of HTGR fuel, 93.5%-enriched uranium was mixed with thorium at the Fuel 

Fabrication Facility (SNM-696, pp. 28–66). As previously discussed, 233U was used for fabrication of 

some of the HTGR fuel compacts that were incorporated into finished HTGR fuel elements. Uranium233 was also used to conduct neutron yield studies at the LINAC and used for various research and 

development activities at various GA locations. Assuming that all of the contamination that is 

discussed in Section 5.2.2 (70,000 cpm/ft2) was due to 93.5% enriched uranium or 233U/232U with an 

activity ratio of 233U to 232U of 55.7 (whichever uranium source term is favorable to the claimant) and 

not thorium, the daily uranium inhalation intake rates for each year during the residual period in 

controlled areas is shown in Table 5-1. 

Uranium intakes during the residual period in uncontrolled (clean) areas of the GA facility are based 

on constant resuspension of the clean area alpha contamination limit for removable contamination 

(6 pCi alpha over an area of approximately 400 cm2) (Section 2.4). The resultant intakes from 1970 

through 1995 are shown in Table 5-2. 

If claimant information specifically indicates that work involving HTGR fuel fabrication was performed, 

the uranium intakes in Table 5-1 should be used to reconstruct dose. The dose reconstructors should 

assign uranium intakes from Table 5-1 by assuming that the material is either highly enriched uranium 

or 233U/232U with an activity ratio of 233U to 232U of 55.7, whichever is favorable to the claimant. If 

claimant information specifically indicates that only work related to Triga reactor fuel was performed, 

only intakes associated 93.5% enriched uranium in Table 5-1 should be assigned. If claimant 

information specifically indicates that work was associated with the GA HCF, uranium and associated 

intakes in Tables 5-3 and 5-4 should be assigned. If claimant information specifically indicates that 

work involving uranium was performed at multiple or unspecified locations, the uranium intakes in 

Table 5-5 should be assigned. 

5.2.4 Plutonium

Attachment E contains a summary of the only workplace monitoring data at GA during plutonium work 

during the operational period. However there are monthly gross alpha air sample summary results 

from fixed air sampling stations in HCF areas including the Operating Gallery and Service Gallery 

from 1965 through 1968. The results for maximum and average concentrations appear to be fairly 

consistent throughout 1968. The maximum gross alpha air concentration from November 1968 was 

7.4 × 10-13 μCi/ml in the HCF Service Gallery (Cockle 1968a). To estimate the fraction of gross alpha 

airborne radioactivity attributable to plutonium isotopes, the radionuclide mix that was determined 

from sampling representative debris from the HLC in 1994 (Tables 5-3 and 5-4) is assumed to be 

representative of the radionuclide mix at the beginning of the residual period. 

The bounding daily intake rate was calculated for the residual period (1970 through 1995) by 

multiplying the gross alpha air concentration of 7.4 × 10-13 μCi/ml by the fraction of activity attributable 

to 238Pu, 239/240Pu, 241Am, 244Cm, and other alpha emitters (Table 5-3) and applying a breathing rate of 

1.2 m3/hr for 10 hours a day for 250 workdays per year. The resultant daily intake rates for 238Pu and 

239/240Pu are 3.81 and 0.33 pCi/d, respectively (Table 5-1). These intakes are considered to be 

constant because the analytical results for 238Pu and 239/240Pu for air samples from March of 1976 in 

the HCF Service Gallery during a planned power outage indicated only slightly lower 238Pu and 

239/240Pu air concentrations as a result of activity that escaped the cells (Bold 1976b). Using the 

fraction of activity attributable to 238Pu and 239/240Pu from Table 5-3 and the gross alpha air 

concentration of 7.4 × 10-13 μCi/ml from November 1968 results in 238Pu and 239/240Pu air 

concentrations of 4.63 × 10-13 μCi/ml and 4.06 × 10-14 μCi/ml, respectively. Air sampling followed by 

alpha spectroscopy in 1976 indicated 238Pu and 239/240Pu air concentrations of 1.25 × 10-13 μCi/ml and 

1.81 × 10-14 μCi/ml during a planned power outage during which time the ventilation system in the 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 72 of 143

HCF was not operating. Given the close agreement between radionuclide-specific air concentrations 

derived from the 1976 alpha spectroscopy data and those derived from the maximum air 

concentration that were observed in 1968, the methodology presented above does not underestimate 

intakes for the residual period. 

Intakes of 238Pu, 239/240Pu, and 241Am during the residual period in uncontrolled (clean) areas of the GA 

facility are based on constant resuspension of the clean area alpha contamination limit for removable 

contamination (6 pCi alpha over an area of approximately 400 cm2) (Section 2.4). The resultant 

intakes would be 1.23 × 10-3 pCi/d for the sum of 238Pu, 239/240Pu, 241Pu, and 241Am at the same 

percentages in Table 3-5 from 1970 through 1995 (Table 5-2). 

If claimant information specifically indicates that work was associated with the GA HCF, plutonium 

and associated intakes in Tables 5-3 and 5-4 should be assigned. If claimant information indicates 

that work involving plutonium was performed at multiple or unspecified locations, the plutonium 

intakes in Table 5-6 should be assigned. 

It is possible that upon intake some forms of plutonium at GA could be retained in the lung longer than 

would be predicted by the normal material type S model (commonly referred to as “Type Super S”). 

Organ doses based on intake of plutonium should be evaluated in accordance with the guidance in 

ORAUT Team Technical Information Bulletin Estimating Doses for Plutonium Strongly Retained in the 

Lung (ORAUT 2007b). 

Summary Tables for Sections 5.2.1, 5.2.2, 5.2.3 and 5.2.6: 

Table 5-1. Controlled area inhalation intake rates (pCi/d)a,d during the 

residual period. 

Year

Th-232 + 

Th-228b

93.5% 

enriched 

uranium U-233/U-232 H-3

1970 1.55E+01 1.55E+01 1.52E+01 2.73E-01 4.93E+00

1971 1.02E+01 1.02E+01 1.00E+01 1.80E-01 4.66E+00

1972 6.67E+00 6.67E+00 6.55E+00 1.18E-01 4.40E+00

1973 4.37E+00 4.37E+00 4.29E+00 7.71E-02 4.16E+00

1974 2.87E+00 2.87E+00 2.82E+00 5.06E-02 3.93E+00

1975 1.88E+00 1.88E+00 1.85E+00 3.32E-02 3.72E+00

1976 1.23E+00 1.23E+00 1.21E+00 2.17E-02 3.51E+00

1977 8.09E-01 8.09E-01 7.95E-01 1.43E-02 3.32E+00

1978 5.30E-01 5.30E-01 5.21E-01 9.35E-03 3.14E+00

1979 3.48E-01 3.48E-01 3.42E-01 6.14E-03 2.97E+00

1980 2.28E-01 2.28E-01 2.24E-01 4.02E-03 2.80E+00

1981 1.50E-01 1.50E-01 1.47E-01 2.65E-03 2.65E+00

1982 9.81E-02 9.81E-02 9.64E-02 1.73E-03 2.50E+00

1983 6.43E-02 6.43E-02 6.32E-02 1.13E-03 2.37E+00

1984 4.22E-02 4.22E-02 4.15E-02 7.44E-04 2.24E+00

1985 2.77E-02 2.77E-02 2.72E-02 4.89E-04 2.11E+00

1986 1.81E-02 1.81E-02 1.78E-02 3.19E-04 2.00E+00

1987 1.19E-02 1.19E-02 1.17E-02 2.10E-04 1.89E+00

1988 7.81E-03 7.81E-03 7.67E-03 1.38E-04 1.78E+00

1989 5.12E-03 5.12E-03 5.03E-03 9.03E-05 1.69E+00

1990 3.36E-03 3.36E-03 3.30E-03 5.93E-05 1.59E+00

1991 2.20E-03 2.20E-03 2.16E-03 3.88E-05 1.51E+00

1992 1.44E-03 1.44E-03 1.41E-03 2.54E-05 1.42E+00

1993c1.23E-03 1.23E-03 1.21E-03 2.17E-05 1.34E+00

1994 1.23E-03 1.23E-03 1.21E-03 2.17E-05 1.27E+00
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 73 of 143

Year

Th-232 + 

Th-228b

93.5% 

enriched 

uranium U-233/U-232 H-3

1995 1.23E-03 1.23E-03 1.21E-03 2.17E-05 1.20E+00

a. The results are 365-day intake rate estimates, based on a 250-day work year, 10 

hours per work day, in the GA FFF (Th and U) and in the Activation Analysis 

Facility (H-3) . 

b. Th-232/Th-228 activity concentration assumptions for dose reconstruction: 1-to-1 

for best estimate for respiratory cancers; 70 to 30 for best estimate for 

nonrespiratory cancers [14].

c. From 1993 through 1995, the uncontrolled area intakes for Th-232 and uranium 

were higher than the controlled area intakes; as a consequence the uncontrolled 

area intakes have been used in Table 5-1 from 1993 through 1995. 

d. Use the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 

2004), to derive estimates of ingestion intakes from the inhalation intake. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 74 of 143

Table 5-2. Uncontrolled area inhalation intake rates (pCi/d)a during the residual period. 

Year

Th-232 + 

Th-228b

93.5% 

enriched 

Uranium Pu-239/240 U-233/U-232 Pu-238 Pu-241 Am-241 H-3 Sr-90

1970 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1971 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1972 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1973 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1974 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1975 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1976 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1977 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1978 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1979 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1980 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1981 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1982 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1983 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1984 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1985 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1986 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1987 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1988 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1989 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1990 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1991 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1992 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1993 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1994 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

1995 1.23E-03 1.23E-03 1.97E-05 1.21E-03 2.17E-05 2.20E-04 9.63E-04 2.71E-05 2.05E-02 6.16E-03

a. The results are 365-day intake rate estimates, based on a 250-day work year, 10 hours per work day. 

b. Th-232/Th-228 activity concentration assumptions for dose reconstruction: 1-to-1 for best estimate for respiratory cancers; 70 to 30 for best estimate for 

nonrespiratory cancers [14].

c. Use the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004), to derive estimates of ingestion intakes from the inhalation intake. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 75 of 143

5.2.5 Mixed Fission and Activation Products

Mixed fission and activation product intakes have been calculated for two different locations because 

of the unique nature of contaminants that are associated with the HCF and the TRIGA reactors and 

critical assemblies. Mixed fission and activation product intakes that are associated with the HCF are 

based on gross beta air sample summary results from fixed air sampling stations in HCF areas 

including the Operating Gallery and Service Gallery from 1965 through 1968. The results for 

maximum and average concentrations appear to be fairly consistent from 1967 through 1968. The 

highest gross beta air concentration of 1.1 × 10-9 μCi/ml was measured in the same month in 1968 as 

was the highest gross alpha air concentration at the same Service Gallery sample station (Cockle 

1968a). To estimate the fraction of gross beta airborne radioactivity attributable to the airborne 

radionuclides, the radionuclide mix that was determined from sampling representative debris in the 

HLC in 1994 (Table 5-4) is assumed to be representative of the radionuclide mix at the beginning of 

the residual period. It was also assumed that the gross beta measurement was not able to detect the 

3H, 55Fe, 241Pu, 63Ni, and 106Ru present on the air sample filter. Concentrations of the latter 

radionuclides were calculated based on their sample proportion. 

The daily intake rate was calculated for the residual period (1970 through 1995) by multiplying the 

gross beta air concentration of 1.1 × 10-9 μCi/ml by the ratio of the total beta sample activity to the 

detectable beta activity (1.06) and then multiplying by the fraction of activity attributable the beta 

emitting radionuclides present (Table 5-4) and applying a breathing rate of 1.2 m3/hr for 10 hours a 

day for 250 workdays per year. The intake rates in Table 5-4 are considered to be constant. 

Mixed fission and activation product beta-gamma emitter intake rates that are associated with the 

uranium-fueled critical assemblies, the HTGR and TRIGA reactors can be calculated using the 

methods in ORAUT-OTIB-0054 (ORAUT 2007a) with use of 90Sr as the indicator radionuclide (Section 

3.1.6.2), and removable gross beta-gamma contamination results (352 dpm/100 cm2) from the TCF, 

which was shut down in September 1969 (Farney 1973). Applying a resuspension factor of 1 × 10-6/m 

(Abu-Eid et al. 2002; ORAUT 2008) results in an estimate of gross beta-gamma airborne 

concentration of 1.59 × 10-5 pCi/L. 

If claimant information specifically indicates that work was associated with the GA HCF the intakes in 

Tables 5-3 and 5-4 should be assigned. If claimant information indicates that work involving mixed 

fission and activation product at multiple or unspecified locations, the mixed fission and activation 

product intakes in Tables 5-6 and 5-7 should be assigned. 

Intakes of alpha and beta emitters during the residual period in uncontrolled (clean) areas of the GA 

facility are based on constant resuspension of the clean area alpha and beta contamination limits for 

removable contamination (6 pCi alpha or 30 pCi beta over an area of approximately 400 cm2) 

(Section 2.4). The resultant intakes would be 1.23 × 10-3 pCi/d for alpha emitters and 6.16 × 10-3

pCi/d for beta emitters from 1970 through 1995. 

5.2.6 Tritium

As discussed in Sections 2.3 and 3.1.5, tritium exposures at GA were primarily associated with tritium 

target changes at the GA Activation Analysis Facility. The only workplace monitoring data available 

for the Activation Analysis Facility are monthly health physics reports from 1965 through 1968. A 

review of the reports indicates that by 1968, the number and severity of tritium contamination 

incidents had decreased significantly. In 1968, the highest level of 3H contamination at the facility was 

6,000 pCi, which is assumed to be over an area of 100 cm2 (Smith 1968). Although the report 

indicated that the area was decontaminated to <100 pCi, it is assumed that at the beginning of the 

residual period contamination at a level of 6,000 pCi/100 cm2 was uniformly distributed at the facility. 

There are no residual period 3H contamination levels available; as a consequence, the tritium 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 76 of 143

radioactive decay constant (0.05635/year) was assigned for the value of the exponential constant. 

The daily intake rates in Table 5-1 were calculated for each year of the residual period (1970 through 

1995) by multiplying by a resuspension factor of 1 × 10-6/m (Abu-Eid et al. 2002; ORAUT 2008) and 

applying a breathing rate of 1.2 m3/hr for 10 hours a day for 250 workdays per year. 

If claimant information specifically indicates that work was associated with the GA HCF, tritium and 

associated intakes in Tables 5-3 and 5-4 should be assigned. If claimant information indicates that 

work involving tritium was performed at the Activation Analysis Facility the intakes in Table 5-1 should 

be assigned. If claimant information indicates tritium work at multiple or unspecified locations, the 

tritium intakes in Table 5-5 should be assigned. 

Intakes of 3H during the residual period in uncontrolled (clean) areas of the GA facility are based on 

constant resuspension of the detection limit of 100 pCi/100 cm2 for removable 3H. The resultant 

intake would be 2.05 × 10-2 pCi/d from 1970 through 1995 (Table 5-2). 

5.2.7 Strontium

There are no workplace monitoring data available for strontium thermoelectric generator work during 

the operational period at GA. However there are monthly gross beta air sample summary results from 

fixed air sampling stations in HCF areas including the Operating Gallery, Machine Shop, Decon 

Room, and Service Gallery from 1965 through 1968. These results appear to be fairly consistent from 

1967 through 1968. The highest gross beta air concentration in 1968 was 1.1 × 10-9 μCi/ml in the 

HCF Service Gallery (Cockle 1968a). To estimate the fraction of gross beta airborne radioactivity 

attributable to 90Sr, the radionuclide mix determined from sampling representative debris from the 

HLC in 1994 (Table 5-4) is assumed to be representative of the radionuclide mix at the beginning of 

the residual period. The daily intake rate was calculated for the residual period (1970 through 1995) 

by multiplying the gross beta air concentration of 1.1 × 10-9 μCi/ml by the fraction of activity 

attributable to 90Sr and applying a breathing rate of 1.2 m3/hr for 10 hours a day for 250 workdays per 

year. The resultant daily intake rate of 2.55 × 103 pCi/d is shown in Tables 5-4 and 5-7. 

If claimant information specifically indicates that work was associated with the GA HCF, strontium and 

associated intakes in Tables 5-3 and 5-4 should be assigned. If claimant information indicates that 

work involving strontium was performed at multiple or unspecified locations, the strontium intakes in 

Table 5-7 should be assigned. 

Intakes of 90Sr during the residual period in uncontrolled (clean) areas of the GA facility are based on 

constant resuspension of the clean area beta contamination limit for removable contamination (30 pCi 

beta over an area of approximately 400 cm2) (Section 2.4). The resultant intakes would be 6.16 × 10-3

pCi/d from 1970 through 1995 (Table 5-2). 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 77 of 143

Summary Tables for Sections 5.2.4, 5.2.5, and 5.2.7: 

Table 5-3. HLC alpha sampling results and intakes.a

Radionuclide

Sample alpha 

concentration (μCi/g) Relative alpha fraction

Daily total activity 

intake rate (pCi/d)b,e

U-232 4.50E-01 4.94E-02 3.00E-01

U-233 3.80E-01 4.17E-02 2.54E-01

U-234 7.90E-02 8.68E-03 5.28E-02

U-235 8.00E-04 8.79E-05 5.34E-04

U-236 1.20E-03 1.32E-04 8.02E-04

U-238 5.00E-04 5.49E-05 3.34E-04

Th-228 3.10E-01 3.40E-02 2.07E-01

Th-229 2.40E-02 2.64E-03 1.60E-02

Th-232 2.00E-03 2.20E-04 1.34E-03

Pu-238 5.70E+00 6.26E-01 3.81E+00

Pu-239/240 5.00E-01 5.49E-02 3.34E-01

Pu-241c5.50E-04 6.04E-05 1.67E+01

Pu-242 NDd NDd NDd

Am-241 7.10E-01 7.80E-02 4.74E-01

Am-243 1.60E-02 1.76E-03 1.07E-02

Cm-242 2.00E-03 2.20E-04 1.34E-03

Cm-243 4.00E-03 4.39E-04 2.67E-03

Cm-244 9.20E-01 1.01E-01 6.15E-01

Cm-245 NDd NDd NDd

Cm-246 NDd NDd NDd

Cm-247 1.00E-03 1.10E-04 6.68E-04

Cm-248 5.00E-04 5.05E-05 3.34E-04

Cm-250 1.40E-02 1.23E-04 9.35E-03

Cf-248 8.00E-04 8.79E-05 5.34E-04

Cf-250 3.00E-04 3.30E-05 2.00E-04

Cf-252 1.00E-03 1.07E-04 6.68E-04

Total 9.12E+00 1.00E+00

a. Adapted from Sills (1994).

b. The results are 365-day intake rate estimates based on a 250-day work year, 10 hours per work day. 

c. Based on 0.0022% alpha decay (which would have been detected by gross alpha air sample analysis) from 25 μCi/g of 

Pu-241 determined by beta analysis (Sills 1994). 

d. ND – not detected. 

e. Use the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004), to derive estimates of ingestion 

intakes from the inhalation intake. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 78 of 143

Table 5-4. HLC beta sampling results and intakes.a

Radionuclide

Sample concentration 

(μCi/g)

Relative beta 

fraction 

Daily intake rate 

(pCi/d)b,g

H-3

c

2.00E-01 1.40E-04 1.34E+00

Fe-55c4.30E+01 3.00E-02 2.88E+02

Ru-106c6.00E+00 4.19E-03 4.02E+01

Ni-63c3.30E+01 2.31E-02 2.21E+02

Rh-106d6.00E+00 4.19E-03 4.02E+01

Co-60 7.50E+01 5.24E-02 5.03E+02

Sr-90 3.80E+02 2.65E-01 2.55E+03

Y-90e3.80E+02 2.65E-01 2.55E+03

Nb-94 4.00E-01 2.79E-04 2.68E+00

Sb-125 2.00E+00 1.40E-03 1.34E+01

Cs-134 2.10E+02 1.47E-01 1.41E+03

Cs-137 2.70E+02 1.89E-01 1.81E+03

Ce-144 2.00E+01 1.40E-02 1.34E+02

Eu-154 4.00E+00 2.79E-03 2.68E+01

Eu-155 1.40E+00 9.78E-04 9.38E+00

Ta-182 3.00E-01 2.10E-04 2.01E+00

Th-231f8.00E-04 5.59E-07 Not applicable

Total 1.43E+03 1.00E+00

a. Adapted from Sills (1994). 

b. The results are 365-day intake rate estimates based on a 250-day work year, 10 hours per work day. 

c. These radionuclides are assumed to have been undetected by gross beta air sample counting. Intake rates for H-3, 

Fe-55, Ni-63, and Ru-106 were calculated based on their proportional sample concentration. 

d. Assumed to be in secular equilibrium with Ru-106. 

e. Assumed to be in secular equilibrium with Sr-90. 

f. Organ doses due to intakes of Th-231 are insignificant in comparison to other assessed intakes [15].

g. Use the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004), to derive estimates of ingestion 

intakes from the inhalation intake. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 79 of 143

The bounding residual period intakes for GA personnel regardless of location have been tabulated 

below (Tables 5-5 through 5-7) by combining the results of Tables 5-1, 5-3 and 5-4 (the highest value 

for each respective radionuclide in Tables 5-1, 5-3, and 5-4 was carried forward into Tables 5-5, 5-6, 

and 5-7). 

Table 5-5. Sitewide bounding controlled area inhalation intake rates (pCi/d)a during the 

residual period. 

Year Th-232 + Th-228b

93.5% enriched 

uranium U-233/U-232 H-3

1970 1.55E+01 1.55E+01 1.52E+01 3.01E-01 4.93E+00

1971 1.02E+01 1.02E+01 1.00E+01 3.01E-01 4.66E+00

1972 6.67E+00 6.67E+00 6.55E+00 3.01E-01 4.40E+00

1973 4.37E+00 4.37E+00 4.29E+00 3.01E-01 4.16E+00

1974 2.87E+00 2.87E+00 2.82E+00 3.01E-01 3.93E+00

1975 1.88E+00 1.88E+00 1.85E+00 3.01E-01 3.72E+00

1976 1.23E+00 1.23E+00 1.21E+00 3.01E-01 3.51E+00

1977 8.09E-01 8.09E-01 7.95E-01 3.01E-01 3.32E+00

1978 5.30E-01 5.30E-01 5.21E-01 3.01E-01 3.14E+00

1979 3.48E-01 3.48E-01 3.42E-01 3.01E-01 2.97E+00

1980 2.28E-01 2.28E-01 2.54E-01 3.01E-01 2.80E+00

1981 2.08E-01 1.50E-01 2.54E-01 3.01E-01 2.65E+00

1982 2.08E-01 9.81E-02 2.54E-01 3.01E-01 2.50E+00

1983 2.08E-01 6.43E-02 2.54E-01 3.01E-01 2.37E+00

1984 2.08E-01 5.45E-02 2.54E-01 3.01E-01 2.24E+00

1985 2.08E-01 5.45E-02 2.54E-01 3.01E-01 2.11E+00

1986 2.08E-01 5.45E-02 2.54E-01 3.01E-01 2.00E+00

1987 2.08E-01 5.45E-02 2.54E-01 3.01E-01 1.89E+00

1988 2.08E-01 5.45E-02 2.54E-01 3.01E-01 1.78E+00

1989 2.08E-01 5.45E-02 2.54E-01 3.01E-01 1.69E+00

1990 2.08E-01 5.45E-02 2.54E-01 3.01E-01 1.59E+00

1991 2.08E-01 5.45E-02 2.54E-01 3.01E-01 1.51E+00

1992 2.08E-01 5.45E-02 2.54E-01 3.01E-01 1.42E+00

1993 2.08E-01 5.45E-02 2.54E-01 3.01E-01 1.34E+00

1994 2.08E-01 5.45E-02 2.54E-01 3.01E-01 1.34E+00

1995 2.08E-01 5.45E-02 2.54E-01 3.01E-01 1.34E+00

a. The results are 365-day intake rate estimates, based on a 250-day work year, 10 hours per work day. 

b. Th-232/Th-228 activity concentration assumptions for dose reconstruction: 1-to-1 for best estimate 

for respiratory cancers; 70 to 30 for best estimate for nonrespiratory cancers [14].

c. Use the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004), to derive 

estimates of ingestion intakes from the inhalation intake. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 80 of 143

Table 5-6. Sitewide bounding alpha sampling results and intakes.a

Radionuclide

Sample alpha 

concentration (μCi/g) 

Daily total activity 

intake rate (pCi/d)b 

Th-229 2.40E-02 1.60E-02

Pu-238 5.70E+00 3.81E+00

Pu-239/240 5.00E-01 3.34E-01

Pu-241c5.50E-04 1.67E+01

Pu-242 NDd NDd

Am-241 7.10E-01 4.74E-01

Am-243 1.60E-02 1.07E-02

Cm-242 2.00E-03 1.34E-03

Cm-243 4.00E-03 2.67E-03

Cm-244 9.20E-01 6.15E-01

Cm-245 NDd NDd

Cm-246 NDd NDd

Cm-247 1.00E-03 6.68E-04

Cm-248 5.00E-04 3.34E-04

Cm-250 1.40E-02 9.35E-03

Cf-248 8.00E-04 5.34E-04

Cf-250 3.00E-04 2.00E-04

Cf-252 1.00E-03 6.68E-04

a. Adapted from Sills (1994). 

b. The results are 365-day intake rate estimates based on a 250-day work year, 10 hours 

per work day. 

c. Based on 0.0022% alpha decay (which would have been detected by gross alpha air 

sample analysis) from 25 μCi/g of Pu-241 determined by beta analysis (Sills 1994). 

d. ND – not detected. 

e. Use the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004), to 

derive estimates of ingestion intakes from the inhalation intake. 

Table 5-7. Sitewide bounding beta sampling results and intakes.a

Radionuclide

Sample concentration 

(μCi/g)

Relative beta 

fraction 

Daily intake rate 

(pCi/d)b

Fe-55c4.30E+01 3.00E-02 2.88E+02

Ru-106c6.00E+00 4.19E-03 4.02E+01

Ni-63c3.30E+01 2.31E-02 2.21E+02

Rh-106d6.00E+00 4.19E-03 4.02E+01

Co-60 7.50E+01 5.24E-02 5.03E+02

Sr-90 3.80E+02 2.65E-01 2.55E+03

Y-90e3.80E+02 2.65E-01 2.55E+03

Nb-94 4.00E-01 2.79E-04 2.68E+00

Sb-125 2.00E+00 1.40E-03 1.34E+01

Cs-134 2.10E+02 1.47E-01 1.41E+03

Cs-137 2.70E+02 1.89E-01 1.81E+03

Ce-144 2.00E+01 1.40E-02 1.34E+02

Eu-154 4.00E+00 2.79E-03 2.68E+01

Eu-155 1.40E+00 9.78E-04 9.38E+00

Ta-182 3.00E-01 2.10E-04 2.01E+00

Th-231f8.00E-04 5.59E-07 Not applicable

a. Adapted from Sills (1994). 

b. The results are 365-day intake rate estimates based on a 250-day work year, 10 hours per work day. 

c. These radionuclides are assumed to have been undetected by gross beta air sample counting. Intake rates for H-3, 

Fe-55, Ni-63, and Ru-106 were calculated based on their proportional sample concentration. 

d. Assumed to be in secular equilibrium with Ru-106. 

e. Assumed to be in secular equilibrium with Sr-90.

f. Organ doses due to intakes of Th-231 are insignificant in comparison to other assessed intakes [15].

g. Use the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004), to derive estimates of ingestion 

intakes from the inhalation intake. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 81 of 143

5.2.8 Other Radionuclides

For radionuclides other than those discussed in this section, dose reconstructors should evaluate 

intakes and organ doses from bioassay results using current Project guidance. 

5.2.9 High-Temperature Gas-Cooled and Reduced-Enrichment Research and Test Reactor

Irradiated Fuel Materials Information

In November 1995, DOE irradiated fuel materials (IFM) that were stored in the HCF were remotely 

inspected, inventoried, and packaged so that the IFM specimens could be removed to allow for the 

decontamination and decommissioning. The IFM consisted of HTGR and RERTR fuel materials. The 

IFM samples were associated with DOE-sponsored reactor fuel development programs at GA 

between 1967 and 1987. Both of these programs involved the in-pile reactor exposure of fuel test 

capsules or fuel test elements. Radiological characterization of HTGR IFM was performed using the 

GARGOYLE computer code, while RERTR IFM was characterized using the ORIGEN2 code for 

fission and activation products and documented actinide content. The results for individual specimens 

are documented in HTGR/RERTR Fuel Materials Characterization and Packaging Report (GA 2002). 

The initial enrichment of HTGR IFM ranged from 10 to 93.15 wt % of 235U while RERTR IFM 

enrichment ranged from 20 to 45 wt % (GA 2003). The radioactive content of the consolidated HTGR 

and RERTR shipment decayed to September 30, 2003, is presented in Attachment J. 

5.2.10 Ingestion

When intake estimates are based on other than bioassay results (as in Section 5.0), the ingestion 

pathway must be considered in addition to the inhalation pathway. Use the guidance in OCAS-TIB009, Estimation of Ingestion Intakes (NIOSH 2004), to derive estimates of ingestion intakes from the 

inhalation intake estimates that are determined in Sections 5.2.2 through 5.2.8.

5.3 ESTIMATION OF EXTERNAL EXPOSURE FROM RESIDUAL RADIOACTIVITY 

To estimate external exposure to residual activity, external dosimetry results that are available after 

1969 are used to provide an upper bound of the dose, and this is assumed to be a constant 

distribution. It would be reasonable to exclude external doses in this period that are clearly 

attributable to ongoing operations (e.g., X-ray exposures) rather than activities that might involve 

exposure to residual radionuclides. It is unlikely that there would be significant nonuniform or 

extremity exposures from residual contamination. Neutron doses from residual contamination are 

assumed to be insignificant. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 82 of 143

5.4 SUMMARY 

To summarize, the information in Table 5-8 is used to account for dose during the residual exposure 

period. 

Table 5-8. Residual radiation exposure summary. 

Source Exposure information

Period of 

applicability Annual exposure

IREPa

distribution

Inhaled residual 

radioactivity

Select absorption type from 

information in Section 3.0

Beginning 

January 1, 1970

Use intake rates determined 

from guidance in Sections 

5.2.2 to 5.2.7 and data in 

Tables 5.1 to 5.4.b

Constant

Ingested 

residual 

radioactivity

Select absorption fraction 

from information in Section 

3.0

Beginning 

January 1, 1970

Apply methodology in 

NIOSH (2004) to inhalation 

intake estimates.

Constant

Residual 

contamination

Penetrating (photons, 

30–250 keV, AP, acute)

Beginning 

January 1, 1970

Use reported dosimetry 

results.

Constant

Residual 

contamination

Nonpenetrating

(electrons >15 keV, acute)

Beginning 

January 1, 1970

Use reported dosimetry 

results multiplied by 2.04.c

Constant

a. IREP = Interactive RadioEpidemiological Program. 

b. Use controlled area or uncontrolled area intake rate estimates as appropriate for claim documentation. 

c. For exposures to MFPs, the recorded nonpenetrating dose from film can be estimated by multiplying the beta dose by 

2.04 based on Table 6-19 of ORAUT (2006a), which indicates the fraction of dose reported as 0.49. 

6.0 ESTIMATION OF EXPOSURE FROM DOE REMEDIATION

The DOE remediation period is defined as 1996 through 1999 (NIOSH 2006c). Remediation activities 

occurred before DOE defined remediation period and included D&D in Buildings 9, 25, and 31-1 from 

1986 to 1996 (Gonzales 2004a). Initial decommissioning of Building 37 began in 1995 and was 

completed in 1999 (Gonzales 2004a). The decommissioning of the HCF is described below. 

Additional remediation activities at GA that occurred from 1996 to 1999 included the Group 7, 8A, 8B, 

8C, 9, 10, and 11 laboratories in Building 2; Buildings 22, 27,30, 31-1, 35-1, 36, 37, 42, and 45; 

Room 049B in Building 9, and certain portions of Building 39 (Gonzales 2004a). The L-307 Tank 

Area was remediated from 1996 to 1999 (Gonzales 2004a). 

In December 1994, GA notified the U.S. Nuclear Regulatory Commission (NRC) formally of the intent 

to cease “principal activities” at the HCF (Building 23) (GA 1998). 

The HCF included the EBOR Fuel Fabrication Facility. Floor space totaled approximately 7,400 ft2

(690 m2

) and consisted of office spaces, three hot cells, and an Operating Gallery; the surrounding 

fenced service area was 46,740 ft2(4,340 m2). The service area included several concrete pads for 

material transfer in and out of the HCF. The remaining area of the yard consists of asphalt and soil. 

There is a small ancillary building of 400 ft2 (37 m2) and two potable liquid waste storage tanks of 

200-gal capacity. Some areas beyond the yard that had been influenced by HCF operations were 

included in remediation activities. 

The General Atomics Hot Cell Facility Decommissioning Plan (GA 1998) describes the work that had 

been done with the HCF:

The primary purpose of the HCF was to facilitate remote inspection of irradiated highly 

radioactive High Temperature Gas-cooled Reactor (HTGR) fuel. The inspection of 

other types of irradiated fuel and materials was secondary to this function. The HighLevel Cell (HLC) was utilized as a remotely operated “hot” machine shop, to cut open

and section irradiated fuel samples according to the inspecting engineer’s direction. 

The Metallography Cell was used for analytical inspection and research activities, 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 83 of 143

including the preparation of samples, which entailed grinding, polishing, and mounting. 

The functions of the Metallography Cell were supported by tasks performed in two 

other areas, the Warm Metallography (Room 107) and the Tritium Extraction Lab 

(formerly the Physical Test Lab) (Room 109), during certain periods of operation. The 

Low-Level Cell (LLC) was used as the interface between the other “hot” cells and other

areas of the building, and housed all shipping cask off-loading and packaging 

operations...Other programs were, at times, conducted within the HCF. What is now 

the Controlled Machine Shop and Weld Area (Room 108/108A) was previously used to 

develop and manufacture the Experimental Beryllium Oxide Reactor (EBOR) fuel. The 

Engineering Scale Tritium Extraction System (ESTES) utilizing Rooms 109, 120, and 

119 demonstrated the ability to extract tritium from New Production Reactor (NPR) 

target compacts. The X-ray Lab (Room 116) was formerly used to perform plutonium 

transport experiments using plutonium oxide. 

The addressed facility and site areas were: 

 Restricted Area Outside Yard, 

 Restricted Area Outside Yard Ventilation System Restricted Area, 

 Restricted Area Outside Yard Waste Tanks Restricted Area, 

 Operating Gallery (Room 105), 

 Dark Room and Change Room (Rooms 105A and 106), 

 Warm Metallography (Room 107), 

 Controlled Machine Shop and Weld Area (Rooms 108 and 108A), 

 Tritium Extraction Laboratory (Room 109), 

 Service Gallery (Room 111), 

 Metallography Cell (Room 112), 

 LLC and HLC (Rooms 113 and 115), 

 Pump (Room 114), 

 X-Ray Laboratory (Room 116), 

 Decontamination Room (Room 118), 

 Tritium Effluent, Tritium Sample Preparation (Rooms 119 and 120), 

 Boiler Room (Room 121), and 

 Manipulator Repair (Room 122). 

From April 1993 through October 1995, plans and procedures to support the decontamination and 

dismantlement of the HCF were completed. In addition, the HCF was characterized and legacy waste 

was packaged for disposal. The DOE-owned legacy IFMs were relocated for onsite storage. 

According to the HCF Closeout Report (Davis undated), “much of the remote-handled, Category 3 

Low-Level Waste stored in the cells was shipped in July 1995.” The GA decommissioning plan 

received interim approval by the NRC in May 1996 (GA 2004). 

Based on a review of the HCF closeout report (Davis undated), most of the HCF was shipped as 

radioactive waste with only a very small fraction of material being released for unrestricted release 

after dismantlement. According to the closeout report dismantlement activities at the HCF appear to 

have been complete by October 1998 and from December 1998 through December 1999, waste from 

with the dismantlement was shipped for disposal or relocated to another building. Other site 

documentation about the termination of whole-body counting at GA (Gonzales 1998) supports the 

chronology in the closeout report. According to Gonzales (1998), a whole-body counter was no 

longer required because the HCF was expected to be dismantled and packaged and air sample 

concentrations were not expected to exceed 10% of the derived air concentration (DAC). Additional 

information on HCF D&D is provided in Attachment K. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 84 of 143

6.1 SOURCE TERM 

The HCF characterization report (Aycock 1995) and D&D plan (GA 1998) relied on gamma 

spectroscopy results and gross beta-gamma measurements although the gamma spectroscopy 

results for some sample media in the HCF characterization report do indicate the presence of 241Am at 

some locations. As a consequence, the source term for the remediation period is based on the 

radionuclide mix that was determined from the results of sampling representative debris from the HLC 

in 1994 (Tables 5-3 and 5-4). GA used the results of the debris sampling and analysis to develop a 

composite DAC specifically for work in the HCF even though the evaluation of smears and air 

samples from the facility indicated that the 90Sr content in the debris from the HLC was higher than the 

90Sr present in the general facility (Sills 1994). According to the characterization report, 21 samples 

from the HCF were analyzed for 137Cs and 90Sr content and the mean 137Cs-to90Sr ratio was 3.51; 

however, none of the samples were from inside the HLC, LLC, or MC. 

The survey and sampling of underlying and surrounding soil areas after HCF dismantlement identified 

elevated activity that consisted primarily of 137Cs, 60Co, and 134 Cs. The 90Sr concentrations were 

estimated using a 137Cs-to90Sr ratio of 2.1:1 (GA 2000). The ratio of 137Cs to 90Sr in HCF debris in 

Table 5-4 is 0.71:1. The results for approximately 1,000 soil samples that were analyzed by gamma 

spectroscopy during post remediation surveys contained <0.3 pCi/g of 241Am and typically <0.3 pCi/g 

of 235U. No detectable activity was identified in 10 final survey soil samples that were selected for 

isotopic plutonium and tritium analysis (GA 2000). 

6.2 SAFETY 

Access to the HCF was limited to a specific group of individuals during the remediation period (1996 

to 1999). Authorization for work was controlled by the use of WAs that identified the proposed work 

scope and activities (GA 1998). Work was to be performed in strict accordance with the methods and 

precautions provided in approved WAs. The GA Radiation Safety Officer compiled a list all personnel 

who had worked at the HCF from 1993 through the remediation period along with the WA that they 

worked under and a description of the activity (Gonzales 2004b). The GA Radiation Safety Officer 

also compiled a list all personnel who had worked at the Nuclear Waste Processing Facility from 1993 

through the remediation period (Gonzales 2004c). 

RWPs were used at the HCF as a means to define specific radiation safety requirements for specific 

activities. When work area conditions were such that the use of engineering controls alone were not 

enough to control airborne radioactivity, an analysis of alternatives was performed before a decision 

was made to place workers in respiratory protective equipment. In any situation in which workers 

were allowed access to an airborne radioactivity area or allowed to perform work that had a high 

degree of likelihood to generate airborne radioactivity in excess of 0.1 DAC, the decision to allow 

access was to be accompanied by the performance of representative airborne radioactivity 

measurements to assess worker intakes (GA 1998). 

Workers at the HCF or frequent visitors were provided with a TLD when working in the HCF restricted 

area or in accordance with a prospective external exposure evaluation. Prospective external 

exposure evaluations were performed annually at a minimum or whenever changes in worker 

exposures warranted (GA 1998). Internal monitoring was conducted in accordance with a prospective 

internal exposure evaluation. Prospective external exposure evaluations were performed annually at 

a minimum or whenever significant changes in planned work evolutions warranted. Facility workers 

and frequent visitors were required to participate in the whole-body counting program (GA 1998). 

Site documentation includes in vivo and in vitro monitoring results beginning in 1989 through June 

1999 (GA 1999). Personnel who were authorized to work in the HCF and Nuclear Waste Processing 

Facility from 1993 through the remediation period (1996 to 1999) are included in the in vivo and in 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 85 of 143

some of the in vitro data. It also appears that personnel under WAs related to the HCF remediation 

were participating in the GA whole-body counting program (GA 1999). 

6.3 ESTIMATION OF INTERNAL EXPOSURE FROM REMEDIATION 

Bounding intakes during the remediation period (1996 to 1999) have been calculated assuming that 

the maximum gross alpha and gross beta air concentrations during remediation activities would be 

bounded by the latter half of the operational period (1965 to 1969) gross alpha and gross beta air 

concentrations. 

Based on available site documentation for the remediation period, the presence of alpha emitters in 

the HCF cannot be ruled out; as a consequence, remediation period intakes have been calculated 

based on the radionuclide mix that was determined from the results of sampling representative debris 

from the HLC in 1994 (Tables 5-3 and 5-4) and multiplying the residual period alpha emitter intakes 

(Table 5-3) by a factor of 3.11 and residual period beta emitter intakes (Table 5-4) by a factor of 1.36. 

The factor of 3.11 is the ratio of the average, maximum gross alpha air concentration of 2.3 × 10-12

μCi/ml (Table 6-1) to the gross alpha air concentration of 7.4 × 10-13 μCi/ml from November 1968 that 

is used for residual period intakes. The factor of 1.36 is the ratio of the average of the monthly 

maximum gross beta air concentrations (1.5 × 10-9 μCi/ml) measured in the Service Gallery (Table 6-

1) to the maximum monthly gross beta air concentration of 1.1 × 10-9 μCi/ml from November 1968 that 

is used for residual period intakes. The resultant intakes are presented in Tables 6-2 and 6-3. 

Table 6-1. HCF air sampling results from 1965 through 1968.a. 

Month and year

Operating gallery Service gallery Operating gallery Service gallery

Monthly average concentration (μCi/ml) Maximum concentration (μCi/ml)

alpha beta alpha beta alpha beta alpha beta

Nov-68 1.20E-14 6.30E-14 1.70E-13 1.15E-10 2.90E-14 7.90E-14 7.40E-13 1.10E-09

Oct-68 6.60E-14 2.10E-13 1.30E-13 2.60E-11 2.00E-14 3.50E-12 1.50E-13 2.20E-10

Sep-68 1.20E-14 6.00E-14 1.10E-13 3.30E-12 1.50E-14 8.00E-14 3.60E-13 2.25E-10

Aug-68 1.50E-14 7.00E-14 1.30E-13 6.50E-11 3.20E-14 1.00E-13 4.00E-13 2.20E-10

Jul-68 1.10E-14 7.00E-14 1.80E-13 2.20E-10 1.40E-14 1.20E-13 1.30E-12 8.60E-10

Jun-68 1.80E-14 8.00E-14 1.90E-13 1.30E-10 8.20E-14 1.20E-13 4.80E-13 5.20E-10

May-68 1.10E-14 1.00E-13 1.10E-13 3.20E-11 2.60E-14 1.70E-13 4.30E-13 2.90E-10

Apr-68 1.80E-14 2.20E-13 1.20E-13 8.60E-12 6.00E-14 1.20E-12 3.80E-13 2.40E-10

Mar-68 1.10E-14 1.00E-13 1.50E-13 1.10E-11 3.70E-14 1.50E-13 6.50E-13 6.80E-11

Feb-68 2.00E-14 1.00E-13 1.90E-13 2.00E-11 2.70E-13 2.50E-13 1.80E-12 4.60E-10

Jan-68 1.50E-14 1.50E-13 2.00E-13 1.60E-11 7.30E-14 5.00E-13 6.60E-13 2.10E-10

Dec-67 1.30E-14 6.00E-14 1.60E-13 2.84E-12 2.40E-14 9.00E-14 1.90E-13 1.30E-11

Nov-67 2.10E-14 1.00E-13 1.70E-13 5.50E-12 3.40E-14 7.10E-13 4.90E-13 4.80E-11

Oct-67 1.50E-14 1.50E-13 1.10E-13 6.70E-12 8.90E-14 1.40E-12 5.70E-13 3.60E-11

Sep-67 1.20E-14 1.80E-13 1.40E-13 1.50E-11 1.90E-14 1.70E-12 1.90E-13 9.80E-11

Aug 19–Sep 1, 67 1.80E-14 4.40E-14 2.00E-13 1.80E-10 4.80E-14 1.30E-13 3.60E-13 1.40E-09

Aug 1–Aug 18, 67 1.07E-09

Jul-67 5.70E-11

Jun-67 6.10E-10

Apr-67 1.20E-09

Mar-67 Negligible 1.40E-10

Feb-67 Negligible 9.80E-11

Jan-67 4.00E-15 2.30E-14 4.80E-13 1.40E-11 4.90E-15 3.00E-14 2.60E-12 8.60E-11

Dec-66 1.30E-10

Nov-66 4.10E-12 6.20E-11 3.30E-11 2.00E-10

Oct-66 2.00E-15 2.70E-14 8.80E-10

Sep-66 3.00E-15 3.50E-14 1.20E-13 2.90E-12 7.00E-14 1.50E-13 1.40E-12 4.30E-09

Aug-66 2.00E-16 2.40E-14 1.30E-13 5.90E-12 6.90E-15 3.40E-14 5.10E-13 8.30E-11

Jul-66 4.00E-14 3.60E-13 1.50E-13 5.00E-13 5.00E-14 4.00E-13 1.30E-11 6.70E-10

Jun-66 3.46E-14 3.65E-13 5.26E-13 2.69E-11 7.90E-14 1.20E-12 2.50E-12 2.25E-10
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 86 of 143

Month and year

Operating gallery Service gallery Operating gallery Service gallery

Monthly average concentration (μCi/ml) Maximum concentration (μCi/ml)

alpha beta alpha beta alpha beta alpha beta

May-66 1.46E-11 1.35E-10 1.20E-12 4.3E-11 1.6E-10 1.6E-09 1.9E-11 6.80E-10

Apr-66 1.20E-14 1.40E-13 1.40E-13 2.90E-09 3.60E-14 5.40E-13 1.60E-13 4.80E-08

Mar-66 6.30E-15 3.70E-14 7.90E-13 5.90E-11 6.80E-14 4.70E-14 1.00E-11 9.30E-10

Feb-66 4.00E-15 3.50E-14 1.70E-13 6.50E-12 4.80E-15 5.70E-14 3.40E-13 4.20E-11

Jan-66 6.00E-15 5.00E-14 2.50E-13 5.90E-12 1.10E-14 1.50E-13 2.00E-12 5.10E-11

Dec-65 6.00E-15 2.80E-14 1.70E-13 5.66E-12 1.00E-13 5.50E-14 5.40E-13 2.50E-11

Nov-65 6.10E-15 2.20E-14 3.10E-13 2.20E-12 1.20E-14 3.10E-14 2.70E-12 1.30E-11

Oct-65 3.70E-15 2.70E-14 1.60E-13 1.28E-11 4.80E-15 3.10E-14 2.70E-13 8.10E-11

Sep-65 8.10E-15 4.10E-14 2.10E-13 2.10E-11 3.40E-14 6.30E-14 7.40E-13 1.80E-10

Aug-65 1.30E-14 7.30E-14 2.40E-13 1.20E-10 2.40E-13 3.30E-13 1.10E-12 1.10E-09

Jul-65 2.00E-14 1.10E-13 1.90E-13 3.50E-11 1.10E-13 4.70E-13 5.50E-13 1.50E-10

Jun-65 1.50E-14 1.70E-13 1.43E-13 2.40E-11 7.00E-14 7.90E-13 2.40E-13 2.20E-10

May-65 1.40E-14 2.40E-13 2.90E-13 4.23E-11 6.00E-14 1.30E-12 1.90E-12 3.90E-10

Apr-65 7.00E-15 9.30E-14 2.10E-13 9.30E-11 1.00E-14 2.70E-13 8.10E-13 7.50E-10

Mar-65 9.20E-15 8.50E-14 4.10E-13 1.68E-10 2.00E-14 2.90E-13 3.70E-12 2.20E-09

Feb-65 1.40E-14 6.20E-13 1.49E-13 8.49E-12 7.00E-14 6.20E-13 2.50E-13 3.30E-11

Jan-65 2.00E-14 3.40E-13 2.40E-13 5.49E-11 1.40E-13 3.40E-13 1.90E-12 4.70E-10

Dec-64 1.45E-14 1.24E-13 5.08E-13 9.76E-11 2.00E-14 1.60E-12 5.08E-13 9.76E-11

Count 3.90E+01 3.90E+01 3.90E+01 3.90E+01 3.80E+01 3.80E+01 3.90E+01 4.70E+01

Average 1.41E-14 1.24E-13 3.22E-13 1.19E-10 5.51E-14 5.03E-13 2.30E-12 1.50E-09

Max 6.60E-14 6.20E-13 4.10E-12 2.90E-09 2.70E-13 3.50E-12 3.30E-11 4.80E-08

Median 1.20E-14 8.50E-14 1.70E-13 2.10E-11 3.50E-14 2.10E-13 5.70E-13 2.20E-10

a. Source: Hughes et al. (1964–1968). 

Table 6-2. HCF debris sampling results and remediation period alpha intakes.a

Radionuclide Sample concentration (μCi/g) Relative alpha fraction Daily intake rate (pCi/d)

b

U-232 4.50E-01 4.94E-02 9.35E-01

U-233 3.80E-01 4.17E-02 7.89E-01

U-234 7.90E-02 8.68E-03 1.64E-01

U-235 8.00E-04 8.79E-05 1.66E-03

U-236 1.20E-03 1.32E-04 2.49E-03

U-238 5.00E-04 5.49E-05 1.04E-03

Th-228 3.10E-01 3.40E-02 6.44E-01

Th-229 2.40E-02 2.64E-03 4.99E-02

Th-232 2.00E-03 2.20E-04 4.16E-03

Pu-238 5.70E+00 6.26E-01 1.18E+01

Pu-239/240 5.00E-01 5.49E-02 1.04E+00

Pu-241c5.50E-04 6.04E-05 5.19E+01

Pu-242 NDd

Am-241 7.10E-01 7.80E-02 1.48E+00

Am-243 1.60E-02 1.76E-03 3.32E-02

Cm-242 2.00E-03 2.20E-04 4.16E-03

Cm-243 4.00E-03 4.39E-04 8.31E-03

Cm-244 9.20E-01 1.01E-01 1.91E+00

Cm-245 NDd

Cm-246 NDd

Cm-247 1.00E-03 1.10E-04 2.08E-03

Cm-248 5.00E-04 5.05E-05 1.04E-03

Cm-250 1.40E-02 1.23E-04 2.91E-02

Cf-248 8.00E-04 8.79E-05 1.66E-03

Cf-250 3.00E-04 3.30E-05 6.23E-04

Cf-252 1.00E-03 1.07E-04 2.08E-03

Total 9.12E+00 1.00E+00

a. Adapted from Sills (1994). 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 87 of 143

b. The results are 365-day intake rate estimates based on a 250-day work year, 10 hours per work day. 

c. Based on 0.0022% alpha decay (which would have been detected by gross alpha air sample analysis) from 25 μCi/g of 

Pu-241 determined by beta analysis (Sills 1994). 

d. ND = not detected. 

e. Use the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004), to derive estimates of ingestion 

intakes from the inhalation intake. 

Table 6-3. HCF debris sampling results and remediation period beta intakes.a

Radionuclide Sample concentration (μCi/g)

Relative beta fraction 

detected on air sample Daily intake rate (pCi/d)b

H-3

c

2.00E-01 1.40E-04 1.82E+00

Fe-55c4.30E+01 3.00E-02 3.92E+02

Ru-106c6.00E+00 4.19E-03 5.47E+01

Ni-63c3.30E+01 2.31E-02 3.01E+02

Rh-106c,d 6.00E+00 4.19E-03 5.47E+01

Co-60 7.50E+01 5.24E-02 6.84E+02

Sr-90 3.80E+02 2.65E-01 3.46E+03

Y-90e3.80E+02 2.65E-01 3.46E+03

Nb-94 4.00E-01 2.79E-04 3.65E+00

Sb-125 2.00E+00 1.40E-03 1.82E+01

Cs-134 2.10E+02 1.47E-01 1.91E+03

Cs-137 2.70E+02 1.89E-01 2.46E+03

Ce-144 2.00E+01 1.40E-02 1.82E+02

Eu-154 4.00E+00 2.79E-03 3.65E+01

Eu-155 1.40E+00 9.78E-04 1.28E+01

Ta-182 3.00E-01 2.10E-04 2.73E+00

Th-231f8.00E-04 5.59E-07 Not applicable

Total 1.43E+03 1.00E+00

a. Adapted from Sills (1994). 

b. The results are 365-day intake rate estimates based on a 250-day work year, 10 hours per work day. 

c. These radionuclides are assumed to have been undetected by gross beta air sample counting. Intake rates for H-3, 

Fe-55, Ni-63, and Ru-106 were calculated based on their proportional sample concentration. 

d. Assumed to be in secular equilibrium with Ru-106. 

e. Assumed to be in secular equilibrium with Sr-90.

f. Organ doses due to intakes of Th-231 are insignificant in comparison to other assessed intakes [15].

g. Use the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004), to derive estimates of ingestion 

intakes from the inhalation intake. 

If no uranium detection threshold is available with the bioassay for this period, the detection threshold 

can be assumed to be 2 times the 2-sigma error. If no error is reported, the detection thresholds in 

the internal dose section can be used. 

6.3.1 Ingestion

When intake estimates are based on other than bioassay results (as in Section 5.0), the ingestion 

pathway must be considered in addition to the inhalation pathway. Dose reconstructors should use 

the guidance in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 2004), to derive estimates of 

ingestion intakes from the inhalation intake estimates in Tables 6-1 and 6-2.

6.4 ESTIMATION OF EXTERNAL EXPOSURE FROM REMEDIATION 

Interpretation of external dosimetry records should be straightforward for the remediation period. The 

assumptions for external exposures in Table 5-5 can be used to estimate external doses for this 

period. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 88 of 143

7.0 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database. 

[1] Olsen, Bernard M. Senior Health Physicist. ORAU Team. August 2007. 

This information is based on a telephone conversation former ORAU Team member Kenneth 

Fleming had with GA employee John Greenwood on March 10, 2006. 

 [2] Olsen, Bernard M. Senior Health Physicist. ORAU Team. August 2007. 

This information is based on a telephone conversation former ORAU Team member Kenneth 

Fleming had with GA employee Bill LaBonte on March 10, 2006. 

 [3] Bloom, Cindy W. Senior Health Physicist. ORAU Team. March 2006. Olsen, Bernard M. 

Senior Health Physicist. ORAU Team. July 2007. 

This information is based on reviews of site personnel bioassay history cards, reports, and 

monitoring records. 

[4] Olsen, Bernard M. Senior Health Physicist. ORAU Team. June 2008. 

The “typical” data on individual total body or lung count data sheets, and the in-vivo data 

typically contained within the “Internal Radiation Monitoring Report” records in a GA dosimetry 

record were arrived at by observation after reviewing several GA dosimetry records covering 

several decades of data. 

[5] Olsen, Bernard M. Senior Health Physicist. ORAU Team. April 2008. 

The specific activities of 233U and 232U in Table 3-2 are 22,000,000 pCi/μg for 232U; and 

9,800 pCi/μg for 233U. Uranium-232 impurity levels in the 233U at GA may have been up to 

8 ppm (Section 2.3). The ratio of 233U to 232U at that level of 232U impurity is 9,800 pCi/μg 

divided by the quantity [22,000,000 pCi/μg X (8 ppm / 1 E+06)], which is 55.7 to 1. 

[6] Olsen, Bernard M. Senior Health Physicist. ORAU Team. July 2007. 

The specific activities and half-lives of the listed radionuclides (presented here only to two 

significant figures) are listed in any standard handbook of radiation protection quantities [e.g. 

The Health Physics and Radiological Health Handbook (Schleien, Slaback, and Birky 1998)], 

or as presented in the application, Integrated Modules for Bioassay Analysis (IMBA). 

[7] Olsen, Bernard M. Senior Health Physicist. ORAU Team. July 2007. 

The data for natural, enriched, and depleted uranium are as presented in IMBA. 

[8] Olsen, Bernard M. Senior Health Physicist. ORAU Team. June 2008. 

A review of some of the data reported on bioassay cards (GA 1963-1967 and GA 1963-1970) 

was performed. In that review, it was noted that, in general, when the value of 0.4 pCi/24 hr 

appeared on the bioassay card for an α analysis, a “<” symbol accompanied that value: if the 

card also contained a column titled “%MPL,” then the “<” symbol appeared in that column; 

when the bioassay card contained no column titled “%MPL,” then the “<” symbol appeared in 

front of the 0.4. For this attribution, two examples of this demarcation are noted: (a) on p. 

1519 of GA 1963-1970, values of 0.6 pCi/24 hr in February 1965 and <0.4 pCi/24 hr in May 

1965 are recorded; and (b) on p. 3915 of GA 1963-1970, the March 1969 value of 0.1 pCi/24 

hr is indicated as being “<1.4 % MPL,” and the June 1969 value of 0.4 pCi/24 hr is indicated as 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 89 of 143

being “1.0 % MPL.” The value of 0.4 pCi/24 hr is equivalent to 0.89 dpm/24 hr (0.4 pCi/24 hr X 

2.22 dpm / pCi). 

A review of the data reported in the dosimetry records of some of the claimants was also 

performed. In that review, it was noted that the “<” notation may not have been always 

transcribed over from the bioassay cards to the computer printout records. For example, in 

the dosimetry records of one claim, four uranium urine results in 1968-69 are recorded as a 

value lower than 0.4 dpm/24 / hr, including one at 0.08 dpm/24 hr. The “<” notation does not 

appear on this record. It is doubtful that a value as low as 0.08 dpm/24 hr could have been 

detected; it is assumed that the “<” notation from the bioassay card record did not get 

transcribed. It is prudent to be favorable to claimants and use the rounded-up value of 1 

dpm/24 hr (from 0.89 dpm/24 hr) for detection sensitivity of uranium in urine for that time 

frame. 

[9] Olsen, Bernard M. Senior Health Physicist. ORAU Team. June 2008. 

A range of minimum detectable activities (by mass) for 235U in the lung is observed or inferred 

in the GA records. In January 1968, a 14 μg result is reported as positive after correcting for 

chest thickness (p. 16 of HNS 1968). That same result of 14 μg appears as a positive result in 

the bioassay cards for January 17, 1968, with the notation, “Valid Results Using New Curve Fit 

technique.” (GDC 1966-1970, p. 69). In 1969, the typical counting error was listed as 0.02 mg 

235U (p. 56 of HNS 1969a). In 1970, positive results as low as 50 μg 235U in the lungs was 

reported (HNS 1970). In 1971, positive results as low as 33 μg 235U in the lungs was reported 

(HNS 1971). The typical numerical values for minimum sensitivities of detection of 235U in the 

lung was characterized in 1976 as being 30 to 60 micrograms (HNS 1976). With a referenced 

range of 30 to 60 μg, and with positive values being reported as low as 14 μg since 1968, it is 

favorable to claimants to assume a nominal minimum sensitivity of 60 μg for 235U in the lung, 

unless there are reported specific data to use instead of that nominal value. 

Since some GA chest count results for 235U are presented in the records as a percentage of 

MPLB, the following is provided. GA (via HNS) was using 255 μg as 100% of MPLB for 235U in 

1968 (p. 16 of HNS 1968). In 1969, the MPLB determination was explained with more detail: 

“The MPLB is a function of the percentage of uranium-235 in the uranium fuel and the ratio of 

the uranium-234 to uranium-235. For fully enriched material the MPLB is 0.245 mg.” (Note (2) 

on p. 51 of HNS 1969b). 

[10] Olsen, Bernard M. Senior Health Physicist. ORAU Team. June 2008. 

Uranium-235 chest counts were available for some workers beginning in 1966, but are not 

considered valid for dose reconstruction purposes until 1968 based on an entry on a wholebody monitoring data card (GDC 1966–1970, p. 69), which states that results are valid using a 

“new curve fitting technique.” A review of approximately 1,000 whole-body monitoring data 

cards (GDC 1966–1970) indicates that before January 1968, 235U chest counts were “not valid” 

(e.g. GDC 1966–1970, pp. 564–567). The reason for the “not valid” determination in the pre1968 counts is explained by HNS as constant bias in the 235U data on p. 45 of HNS 1967. By 

mid January 1968, HNS had solved the bias issue in counting for 235U (HNS 1968). In 

addition, a review of the in-vivo data contained within several GA dosimetry records covering 

several decades of data indicate that the “not valid” phrase does not typically appear for data 

after January 17, 1968. If the phrase “not valid” or “not corrected for chest wall thickness” 

does not appear on the record of a positive in-vivo count performed after 1967, the reported 

result can be used in the dose reconstruction. This conclusion is further supported by the 

detailed counting and analyses procedures provided in the later Helgeson reports (HNS 1968, 

HNS 1969c, chest thickness correction is discussed in detail, e.g., beginning on p. 6 of HNS 

1976). 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 90 of 143

[11] Olsen, Bernard M. Senior Health Physicist. ORAU Team. June 2008. 

A review of some of the data reported on bioassay cards (GA 1963-1967 and GA 1963-1970) 

was performed. In that review, it was noted that except for a few entries in 1966, the “<” 

notation does not appear for a beta (“β” or “MFP”) result. Positive results as low as 6 to 7 

pCi/24 hour for β analyses were observed throughout these records (e.g., p. 1087 of GA 1963-

1967, p. 3849 of GA 1963-1970). The few “<” results were recorded in the 13 to 15 pCi/24 

hour range (e.g., p. 3857 of GA 1963-1967, p. 72 and 87 of GA 1963-1970). Similar 

observations were made in a review of the data reported in the dosimetry records of some of 

the claimants. In that review, the lowest recorded β result observed was 14 dpm/24 hour (6.3 

pCi/24 hour). 

A review of the urine sample calculation sheets indicates that various standards (Sr-90 and 

Cs-137 for example) were used to convert beta count rate results to activity. A January 1964 

result (p. 5 of GDC 1965a) indicating no detectable activity (at “<3 net cpm”) is converted to 

“<69 pCi as Cs-137/24 hr” using the 137Cs “standard” for that count of 8.75 pCi/net cpm. An 

April 1965 result (p. 10 of GDC 1965a) indicating 1.3 net cpm is converted to 12 pCi as Sr90/24 hr using the Sr-90 “standard” for that count of 1.55 pCi/net cpm. From a review of the 

calculations in GDC 1965a and 1965b, it appears that 90Sr was used as the calibration 

standard when converting to a gross beta result, and 137Cs was used as the calibration 

standard when converting to a mixed fission product result. A nominal range for the minimum 

sensitivity of a gross beta result (without normalization to time or urine volume) would be 5 to 

10 pCi and would be on the order of 70 pCi for a mixed fission product result based on this 

review. 

[12] Bloom, Cindy W. Senior Health Physicist. ORAU Team, March 2006. 

This information is based on correspondence with the current GA Radiation Safety Officer. 

[13] Olsen, Bernard M. Senior Health Physicist. ORAU Team. July 2007. 

The methodology is that used to evaluate residual radioactivity over time as presented in 

Section 5.0 of OCAS-TKBS-0002, Technical Basis Document for Atomic Energy Operations at 

Blockson Chemical Company, Joliet, Illinois (NIOSH 2007c).

[14] Olsen, Bernard M. Senior Health Physicist. ORAU Team. February 2008. 

The ratio of 232Th to 228Th is variable and ranges from equal activity concentrations to the 

minimum condition of 228Th being present at approximately 40% the activity concentration of 

232Th (West 1965; GA 1971b). The relative ratio of 232Th to 228Th is significant for dose 

reconstruction. Organ dose that is determined by applying the dose conversion factors of 

ICRP Publication 68 (ICRP 1990) is maximized for nonrespiratory cancers by assuming all 

thorium is 232Th, while the dose is maximized for respiratory cancers by assuming that all 

thorium is 228Th. For a nonrespiratory cancer, it is favorable to the claimant to assume that all 

the thorium is 232Th. For a respiratory cancer, it is favorable to the claimant to assume that all 

the thorium is 228Th. For a best-estimate evaluation, dose reconstructors should assume that 

the ration of 232Th to 228Th is 1:1 for a respiratory cancer, or 7:3 (the minimum condition of 

228Th present at 40% that of 232Th) for a nonrespiratory cancer. 

[15] Olsen, Bernard M. Senior Health Physicist. ORAU Team. February 2008. 

Thorium-231 adds less than 1 millirem total dose to both the inhalation and ingestion 

calculations for the residual period, and it is not applied for the remediation period. Based on 

Tables 5-2 and 6-2, the maximum 1-year inhalation intakes of 231Th would be less than 2 pCi 

during the residual period and less than 250 pCi during the remediation period. Applying the 

maximum inhalation and ingestion dose coefficients from ICRP Publication 68 to those 

estimates on intake results in less than 1 mrem each to the maximally exposed organ (ICRP 

1990). 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 91 of 143

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Radiological Health, Third Edition, Williams & Wilkins, Baltimore, Maryland, January. 

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SNM (Special Nuclear Materials) license 69 (Docket 70-72) – Amendments 41 through 62, multiple 

dates, General Atomics, La Jolla, California, multiple dates. [SRDB Ref ID: 19013] 

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Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 106 of 143

The following is a general description of the fuel fabrication from License SNM-696 for the Fuel 

Fabrication Facility, which began operations in 1962. Similar operations were occurring in other GA 

facilities before 1962. This description does not necessarily include all activities involving fuel 

fabrication. 

A1.0 FABRICATION OF PYROLYTIC URANIUM OXIDE OR CARBIDE OR URANIUMTHORIUM OXIDE OR CARBIDE PARTICLES

Station A1 – Raw Material Storage. Raw material storage was in the raw material storage vault. Any 

enrichment greater than 93.5% was subject to the same rules as fully enriched uranium. The uranium 

was stored in either 1-gal or 3.6-L containers in such a way as to avoid criticality (SNM 696, pp. 30–

31). 

Station A2 – Batch Weighing. Uranium dioxide powder was received in stock containers and weighed 

into individual portions. The weighing occurred in a glovebox. Horizontal racks held the stock 

containers on centers of 16 in. or greater. Each rack was limited to four stock containers so the rack 

held a total of 30 kg of 235U or less, the station's mass limit. The material was then weighed and 

placed into batch containers with screw caps for removal from the station. These batch containers 

were placed on another rack with centers at least 16 in. from each other, the scale, and other stock 

material. Each container was loaded with 7.3 kg or less and was sealed to prevent water entrance 

(SNM 696, p. 32). 

Station A3 – In-Process Storage. The material was stored in closed containers 5 in. or less in 

diameter on steel shelves. The containers were placed in an array of 16 and centered horizontally but 

set as close as possible vertically. The effect was to make a linear array of 5-in.-diameter vertical 

cylinders on 16 centers with surface-to-surface spacing of 11 in. or more. Each bottle was held firmly 

in a storage location by a bar or chain across the opening, and intermediate spacers prevented the 

bottle from tipping over. The 235U mass density was always 1.8 kg/L or less (SNM 696, p. 32). 

Station A3A – In-Process Storage. This station was similar to Station A3 except that material was 

stored in 3.6-L containers with a maximum of 3.6 kg of 235U (SNM 696, p. 32–33). 

Station A4 – Weighing. The station was used to weigh all material other than uranium (SNM 696, 

p. 33). 

Station A5 – Blending. A weighed batch of uranium from dioxide from Station A3, carbon, a binder 

and, if required, thorium dioxide using a solvent weighed at Station A4 were blended at this station. 

The maximum amount of 235U at this station was 750 g. The material was mixed in an 8-quart blender 

inside a glovebox (SNM 696, p. 33). 

Station A5A – Blending. This station was similar to Station A5 except 235U was limited to less than 

5 kg. The blending is done in a blender with a maximum ability to mix 5 L in accordance with the 

1 kg/L for 235U (SNM 696, p. 33). 

Station A6 – Drying. The mixture of carbon, binder, solvent, uranium dioxide, and thorium dioxide 

from Station A5 was delivered to Station A6 in trays and heated in an oven at approximately 140°F. 

The mass limit for this station was the blending batch limit of 750 g of 235U or less split into three trays. 

The oven had three open shelves, each large enough for only one tray. Unused shelves were 

covered with aluminum foil (SNM 696, p. 34). 

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Station A6A – Drying. The mixture of uranium dioxide, thorium dioxide, if used, carbon, binder and 

solvent from Station A5 or A5A was brought in trays one at a time and dried in an oven at 

approximately 150°F. A total of six trays were placed in the end of the oven one after another and run 

through on a conveyor to the other end. The combined mass of trays could contain no more than 8 kg 

of 235U, the station mass limit. The trays were removed and taken one at a time to Station A7. 

The heating was supplied by an electric radiant heating element that ran the length of the oven. The 

oven was an insulated, closed-box structure that was connected by a duct to the facility hood 

ventilation system to remove fumes and dust. Wire cloth restricted the area above the tray conveyor 

to a height to accommodate only one tray (SNM 696, p. 34). 

Station A7 – Sizing and Weighing. The dried mixture of carbon, binder, uranium dioxide, and thorium 

dioxide, if used, was brought to the station in one oven tray at a time. The material was transferred 

into a 3.6-L container. Additional trays would be added until the container was full. This container 

was emptied using a smaller container with a limit of no more than 0.75 L and placed on a set of 

screens and sized into the required size fractions. The screens were unloaded one at a time and the 

material fraction was removed from the screen and placed in a separate 3.6-L container. This 

procedure was repeated until the entire container was emptied. The sized fractions were weighed in 

their separate containers and put in in-process storage. 

The entire station was enclosed in a glovebox connected by a duct to the facility hood ventilation 

system. A special storage fixture was provided in the glovebox so there was a minimum of 16 in. 

between centers for the containers. No more than four containers and one screen set were present at 

the station any one time. The total mass limit was 10 kg of 235U (SNM 696, pp. 34–35). 

Station A8 – In-Process Storage. This station was the same is Station A3 or A3A (SNM 696, p. 35). 

Station A9 – In-Process Storage. This station was the same is Station A3 or A3A (SNM 696, p. 35). 

Station A10 – Batch Weighing. This station was similar to Station A2. The dried mixture of carbon, 

binder, uranium dioxide, and thorium dioxide sized at Station A7 was weighed for blending at Station 

A11 or for loading directly into the graphite crucible in Station A11A. The station mass limit was 10 kg 

of 235U (SNM 696, p. 35). 

Station A11 – Blending and Loading the Conversion Crucible. Weighed portions of the dried, sized 

mix from Station A10 were packed into a crucible. The crucible was a graphite sleeve closed at one 

end with an inside diameter under a 5 in. or less. The filled crucible was loaded to 800 g or less of 

235U, which was a station mass limit. The operation at this station was performed in a glovebox (SNM 

696, p. 36). 

Station A11A – Conversion Crucible Loading. Individual containers of the dried mix and Station A10 

were brought to this glovebox and opened and material was transferred into a single graphite 

conversion crucible similar to what happened at Station A11. The inside diameter of the conversion 

crucible was 4.5 in. or less. The mass limit for the crucible was 6.5 kg of 235U, which was also the 

station mass limit (SNM 696, p. 36). 

Station A12 – Conversion. Converter crucibles loaded at Station A11 or A11A were placed in an 

induction furnace and heated to a maximum temperature between 2,000°C and 2,800°C under a high 

vacuum or in an inert gas atmosphere. The crucibles were then cooled and removed from the 

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converter to an inert atmosphere glovebox for unloading. The converter furnace consisted of a large 

quartz cylinder heated by water-cooled induction coils. One or more crucibles were placed inside a 

close-fitting cylindrical graphite susceptor. The susceptor was secured inside the quartz cylinder and 

insulated for the quart cylinder with carbon insulation. 

One or more loaded crucibles were brought to the conversion station with up to 6.5 kg of 235U, which 

was the mass limit for the station (SNM 696, pp. 36–37). 

Station A12A – Conversion. Weighed batches of the dry, sized mix from Station A10 or the dicarbide 

product from Station A12 was loaded into the hopper of the automatic feeder for the converter at 

Station A13. The material hopper was a 4-in.-diameter by 6-in. stainless-steel canister. The loaded 

hopper was sealed and the interior charged with an inert gas atmosphere. The presized material was 

automatically fed into the furnace section of the converter. The furnace section consisted of an 8-in.-

diameter graphite tube surrounded by 4 in. of lampblack and a thick quartz tube. Water-cooled 

induction heaters were placed around the quartz tubes, and protective screens surrounded the entire 

furnace section. 

A cooling section in series with the furnace section consisted of a 9-in. inside diameter, 0.125-in. 

walled stainless-steel tube that was surrounded by cooling coils. The converted dicarbide material 

entered the cooling section continuously from the furnace section and was discharged into a 5-in.-

diameter stainless-steel receiver. 

The furnace section was heated to between 2,000° C. and 2,800° C. The material hopper, automatic 

feeder, furnace section, cooling sections, and receiver were continuously purged with an inert gas. 

The outlet purge gas of the converter consisting of the inert gas and carbon monoxide was conducted 

to an additional high-efficiency absolute filter. The outlet gas from the filter was diluted with large 

quantities of air and injected into the glovebox and hood ventilation system (SNM 696, pp. 37–38). 

Station A12B – Conversion. This station was the same is Station A12B with the exception that all 

parts of the furnace, the material feed hopper, furnace section, cooling section, and the material 

receiver had an inside diameter of 5 in. or less. The SNM was loaded into the sealed inert-gas-filled 

hopper at Station A13 and brought to this station. The mass limit of this station was 4 kg of 235U (SNM 

696, p. 38). 

Station A12C – Conversion. The converter at this station was a horizontally oriented furnace that 

consisted of a resistance preheater a quartz cylinder that was heated by water-cooled induction coils 

with a water-coil cooling section. The material to be converted was moved through a 5-in. inside 

diameter cylinder lining the three sections of the converter and was introduced into the preheater end 

of the converter in individual graphite boats, each with a mass limit of 2 kg of 235U. The boats were 

pan-shaped (1 in. deep by 4 in. wide by 12 in. long) and flat-bottomed with a vertical piece shaped like 

a segment of a circle and each end was sized to fill the converter cylinder and partition the boats from 

each other. The boats were placed in the liner cylinder one behind the other with each subsequent 

boat pushing the previous boats through the converter. A mechanical feed mechanism propelled the 

boats through the cylinder. The volume of the boat (0.8 L) and the bulk density (2 g/cm3) of the 

material prevented filling a boat in excess of the 2 kg of 235U mass limit (SNM 696, p. 39). 

Station A13 – Crucible Unloading, Feed Hopper Loading, and Receiver Unloading. The crucible from 

Station A12 was brought to this inert-gas-filled glovebox. The material was unloaded, passed through 

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a screen to break up any large chunks, and loaded into a special bottle that fit into the separator, 

Station A14. 

The material feed container or the receiver for Station A12A or A12B was brought to this glovebox for 

loading or unloading. Feed material was transferred directly from individual storage containers into 

the feed hopper for Station A12A or A12B; the station mass limit was 3.6 kg of 235U (SNM 696, p. 40). 

Station A14 – Separation and Weighing. The carbon and uranium-thorium dicarbide or uranium 

dicarbide mixture from the converter came to this station in a feed bottle that was attached to the 

separation apparatus. It was screw-fed to the classifying chamber where a flow of inert gas carried 

away the carbon dust. The product particles of uranium dicarbide and thorium dicarbide fell into the 

product bottle. The mass limit for the station was 3.6 kg of 235U. Before another batch was charged to 

the classifier, a material balance check was performed to guard against holdup material in the 

classifier (SNM 696, p. 40). 

Station A15 – In-Process Storage. This station was similar to Station A3 or A3A (SNM 696, p. 40). 

Station A16 – In-Process Storage. This station was similar to Station A3 or A3A (SNM 696, p. 41). 

Station A17-1 – Weighing, Screen and Loading. This station was a controlled-atmosphere, sealed 

glovebox of the same kind as that at Station A13 and had a mass limit of 3.6 kg of 235U. Containers 

were opened and the material was screened, sampled, weighed, and placed in the coater loading 

fixture. Weighed batches of material with 3.6 kg of 235U or less were placed in a loading fixture that 

was then sealed and removed from the glovebox (SNM 696, p. 41). 

Station A17A – Coating. The coater was an insulated cylinder that was heated by resistance 

elements or water-cooled induction coils and was vented to the exhaust system. In operation it was 

heated, flushed with inert gas, and loaded with particles. The coater tube, the portion holding SNM, 

was 5 in. inside diameter or less. Material was brought to the station in a sealed loading fixture, 5 in. 

inside diameter or less with up to 3.6 kg of 235U, the station mass limit. All of the uranium-containing 

material was placed into the coater where appropriate coating gases were introduced (SNM 696, 

p. 41). 

Station A17B – Coating. This dry coater was an insulated cylinder that was heated by resistance 

elements and was vented to one of two exhaust systems. The coater was cooled using halocarbon 

rather than water. A series of graphite inner liners was used to control thermal distributions, gas, and 

product flow parameters. 

The material was brought to the station in sealed containers each having up to 3.6 kg of 235U. The 

maximum quantity of particles in the coater did not exceed 20 kg of 235U, and enrichments did not 

exceed 93.2%. The form of the materials was one of the following: UC2 kernels, UC2 LTI particles, 

UC2 SiC coated particles, UC2 outer LTI particles, or uranium-thorium mixtures of the above. The 

carbide materials were typically converted from oxide materials in a furnace at 2,200°C to drive off all 

hydrogen present. The carbide-converted materials were offloaded from the conversion furnace and 

stored under an inert atmosphere to prevent back-conversion to the oxide form by hydrolyzing. 

After coating, the hot particles were transferred by gravity to one or both of 25-in.-diameter, 4-ft.-long 

cylindrical cooling hoppers, each capable of containing 10 kg of 235U. After cooling the material was 

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meted out of the cooling hoppers into several 3.6-L containers with a limit of 3.6 kg of 235U each. An 

inert atmosphere was maintained during all material transfer operations (SNM 696, p. 42). 

Station A18 – Screening, Weighing, and Packaging. Carbon-coated particles of uranium dicarbide 

and thorium dicarbide were screened in a hood and the product and oversized materials were 

packaged for storage. Material arrived in the cylindrical receiving container from the coater. Portions 

were placed on the screen set and shaken manually. The size fractions were removed to storage 

cans, sealed, weighed, and removed. A maximum of 10 kg of 235U was allowed at the station at any 

time (SNM 696, p. 48). 

Station A19 – In-Process Storage. This station was similar to Station A3 or A3A (SNM 696, p. 48). 

Station A20 – In-Process Storage Vault. The storage vault was a closed installation approximately 

30 ft. square with 8-in.-thick reinforced concrete walls and a 6-in.-thick reinforced concrete ceiling. All 

entrances to the vault were secured by doors with combination locks except the east wall double 

doors, which were barred on the inside to provide an emergency escape. Storage racks for the 

material were laid out in arrays in the vault. 

The SNM in the vault was stored in steel racks that were securely fastened to the walls or partitions. 

Individual storage locations were provided by specially installed metal fixtures or spacers that were 

designed according to the type of storage unit. Vertical shelf spacing was at least 16 in., and material 

was stored on the shelf was not less than 16 in. center-to-center spacing (SNM 696, pp. 48–49). 

Station A21 – Oversized Material Ballmilling, Screening. Oversized dicarbide material from Station 

A13 was processed at this station by ballmilling to obtain the proper material size fractions. One or 

two ballmill units were used at this station as well as a screen set and as many as four containers at a 

time. Each unit had a volume of no more than 3.6 L, and the total station mass limit was 3.6 kg of 

235U. Racks were installed in the glovebox for each material container to provide a minimum of 16 in. 

center-to-center spacing and 12-in. surface for spacing between each container, each ballmill, and the 

screen set. All operations at the station were done in a closed glovebox to contain any dust. 

At any one time, only a sealed container of oversized dicarbide material with up to 3.6 kg of 235U was 

brought to the station; the container had a volume of 3.6 L or less. The material was transferred from 

the feed container into one or both of the ballmills. As product was formed it was removed and 

screened into three size fractions. Each size fraction, one at a time, was removed from the screen 

and placed in an individual container of 3.6-L volume or less. The material from each ballmill was 

handled alternately until the entire batch of the material was sized and put in one of the three size 

fractions bottles. All material of the given batch was removed before any additional material was 

charged to the station because the mass limit was 3.6 kg 235U (SNM 696, p. 49). 

Station A22 – Splitting. Batches of the dicarbide particle material from Station A13 or the coated 

particle material from Station A18 were blended at this station by successive splitting. Five splitters in 

a special fixture were set up in a series-parallel arrangement so, as the material passed through the 

first unit, it was split into two equal fractions. The fractions entered separate units simultaneously and 

were then split into two more fractions. Each of these fractions entered a splitter to be separated into 

two more fractions. Each of these fractions was then passed into an individual receiver container. A 

second batch of material was passed through the splitters and the eight fractions of the second batch 

were added to those of the first. The splitting and blending step was sometimes repeated for several 

batches of material with a total 235U content of up to 28.8 kg, the station mass limit. All of the blending 

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process occurred in a closed glovebox to contain any dust. As with all gloveboxes, it was connected 

to the facility ventilation system and absolute filters so when the glovebox material pass-through was 

open an inward airflow was provided to control air contamination outside the box. 

The material was brought to this station in closed containers with a volume of 3.6 L or less and a 

mass of 2.6 kg 235U or less. The container was opened inside the glovebox and then passed through 

the splitters (SNM 696, p. 49). 

A2.0 FABRICATION OF COMPACTS COMPOSED OF COATED PARTICLES

Station B1-1 – Splitting. Uranium and/or uranium-thorium particles were brought to this station in a 

container whose volume was limited to 3.6 L and whose mass was limited to 3.6 kg of 235U for 

blending by a splitting technique. The particles were poured into the splitting fixture such that the 

material was split into as many as 10 equal fractions. Each of the fractions flowed by gravity through 

tubing into as many as 10 individual receiver containers, each of which were limited to 3.6 L and 

3.6 kg of 235U. The splitter and tubing were emptied before another batch of particles was introduced. 

Once splitting was accomplished, the filled receiver containers were removed from their positions at 

the station one at a time. 

There were actually two B1-1 splitting stations. The B1-1A station was in the vicinity of the A8 and 

A17 stations. The B1-1B station was in the vicinity of the A18 and A19 stations (SNM 696, p. 50). 

Station B1 – Weighing. Coated uranium-thorium dicarbide particles were brought to this station and 

split and weighed into individual compact charges. The mass limit of this station was 0.35 kg of 235U 

(SNM 696, p. 51). 

Station B2 – Mixing. Several individual compact charges containing uranium-thorium dicarbide were 

brought to this station and mixed in several blenders with graphite, a binder, and the hydrogenous 

solvent. After mixing was completed, the single compact mixes were transferred to boats and placed 

in a drying oven. The mass limit of this station was 0.35 kg of 235U (SNM 696, p. 51). 

Station B3 – Drying. Boats that contained single compact mixes from Station B2 were placed in a 

drying oven and heated until the solvent evaporated. The mass limit of this station was 0.35 kg of

235U (SNM 696, p. 52). 

Station B4 – Transfer Cart. The material was transferred among stations on standard four-wheel 

carts with a special top to hold the various containers and compacts. Each cart contained 0.35 kg of 

235U or less. No more than one cart was allowed in any one station at a time other than Station B9, 

where there could be a maximum of three (SNM 696, p. 53). 

Station B5 – Hot Pressing. Batches of carbon, binder, and uranium-thorium dicarbide were placed in 

the dies of a hydraulically operated, electrically heated press and heated to approximately 700°C at 

about 5,000 psi. The mass limit for this station was 0.35 kg of 235U (SNM 696, p. 53). 

Station B6 – Sintering Furnace Load Box. Compacts that were to be sintered were brought to this 

station on a transfer car, and entered an inert-gas-filled glovebox through an airlock. The compacts 

were then placed on graphite trays for introduction into the sintering furnace, Station B7 (SNM 696, 

p. 53). 

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Station B7 – Sintering. Pressed fuel compacts were heated in a horizontal furnace to temperatures of 

up to 2,000°C. The furnace consisted of a resistance preheater, a quartz cylinder heated by watercooled induction coils, and a water coil cooling section. 

The compacts were placed on graphite trays that were introduced into the preheater end of the 

furnace, one behind the other, with each subsequent tray pushing the previous trays through the three 

sections of the furnace. A feed mechanism propelled the trays. The trays were pan-shaped with a 

holed insert such that eight compacts were equally spaced on the tray (SNM 696, p. 54). 

Station B8 – Sintering Furnace Unload Box. Compacts that were to be sintered were brought to this 

station on a transfer cart, and entered an inert-gas-filled glovebox through an airlock. The compacts 

were then placed on graphite trays for introduction into the sintering furnace, Station B7 (SNM 696, 

p. 54). 

Station B9 – Inspection. The loaded transfer cart with sintered compacts was brought to this 

inspection station. The compacts were removed from the cart one at a time, inspected, and placed on 

either off-specification compact transfer carts or acceptable compact transfer carts. The mass limit for 

this station was 0.35 kg of 235U (SNM 696, p. 55). 

Station B10 – Tube Loading. Compacts that passed final inspection were brought to this station on 

transfer carts for loading into special storage tubes. No more than one tube was loaded at a time. 

This station had a mass limit of 0.35 kg of 235U (SNM 696 p. 55). 

Station B11 – In-Process Storage. Storage tubes that were loaded with compacts at Station B10 

were transferred, one at a time, to a special storage rack at this station. The storage rack was 

constructed of steel and secured to the building wall. Special hangers that would hold only one tube 

were attached to the rack (SNM 696, p. 55). 

A3.0 FUEL ELEMENT FABRICATION

Station C1 – Fuel Element Assembly. Compacts were brought to this station on a transfer cart and 

placed into a fixture that contained three horizontally oriented V-trays in a vertical plane. Each V-tray 

could contain up to 30 compacts. Three partially assembled fuel elements were placed in the same 

vertical plane as an extension of the V-trays and the compacts were loaded into the elements. 

Each compact contained up to 10 g of 235U and was 3 in. in length and 3 in. in diameter (SNM 696, 

p. 56). 

Station C2 – Element Curing and Inspection. The partially assembled elements were removed from 

Station C1 one at a time and placed in a horizontal plane array. The remaining graphite components 

were joined to the elements and the cemented joints were cured in curing ovens. Movable tables 

were used to accumulate elements while other elements were curing (SNM 696, p. 56). 

Station C3 – Storage. For storage, assembled fuel elements were placed into cans inside a specially 

designed container. The container held 25 fuel elements with a maximum weight of 7.2 kg of 235U. 

The individual cans had a 4-in. outside diameter with a 0.5-in. wall thickness. The cans were placed 

in the container in a square array with a 4-in. pitch. The container was constructed of 0.25-in.-thick 

steel side plates with a plating of cadmium metal 0.0025 in. thick in the inside surfaces (SNM 696, 

p. 56). 

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Station D1 – Weighing. SNM in a container with limits of 3.6 L and 3.6 kg of 235U was sometimes 

withdrawn from the process or in-process storage for weighing, measuring, and sampling, and 

inspection (SNM 696, p. 61). 

Station D2 – Chemical and Physical Laboratories. SNM was sometimes withdrawn from various 

steps of the process for chemical and physical analysis and for laboratory-scale processing and 

process investigation. The total amount of 235U maintained at this station was limited to 350 g (SNM 

696, p. 61). 

Station E1 – Blending. Uranium dioxide was brought to this station from Station E2 or in-process 

storage in a batch that contained up to 790 g of 235U, the station mass limit. The uranium was mixed 

in a batch-type blender with thorium dioxide, graphite flour, ethyl cellulose, and trichloroethylene. The 

thorium-to235U ratio was not less than 3.6 to 1 (SNM 696, p. 63). 

Station E2 – Rolling. The mixed material was processed through a rolling mill to achieve 

homogeneity. Adjoining rolls turned in opposite directions and the material moved from the first roller 

to the last roller where it came off into a container and was returned to the mixing process. The mass 

limit this station was 790 g of 235U (SNM 696, p. 63). 

Station E3 – Drying. The mixture of uranium dioxide, thorium dioxide, graphite flour, and ethyl 

cellulose from Station E2 was poured into 1-in.-deep trays and the trays were placed into a horizontal 

drying oven, one behind the other, and conveyed to the opposite end of the oven. The oven when full 

could contain up to 8 kg of 235U, the station mass limit. The trays were removed from the oven one at 

a time and taken to Station E4. 

Heating was supplied by electric radiant heating that ran the length of the oven. Angle-iron framing on 

both sides of the tray conveyor restricted the areas such that trays could not be stacked (SNM 696, 

p. 63). 

Station E4 – Unload and Weigh. The dried mixture of uranium dioxide, thorium dioxide, graphite flour, 

and ethyl cellulose was brought to this station from the drying ovens in individual pans. The pans 

were unloaded and the material weighed into a container with limits of 3.6 L and 3.6 kg of 235U, the 

station mass limit. When a container was filled, it was transferred in single units by an elevator that 

ran through the unload scuttles to the scuttle of the Grinding-Screening Station, E5 (SNM 696, pp. 

63–64). 

Station E5 – Grinding–Sizing. Individual containers of the dried mixture from Station E4 were brought 

to this station and placed in a glovebox and a special storage fixture. The station mass limit was 

8.5 kg of 235U. The special storage fixture in the box provided for each container to have a minimum 

of 16 in. center-to-center spacing and 8 in. surface-to-surface spacing from any other container. 

The dried material was poured from the containers, one at a time, to the funnel feeder, which was 

connected by a 3-in. pipe through the bottom of the glovebox to a rotary valve that fed a 5.0-L 

mechanical grinder. The ground material passed through a flexible tube to a screen separator unit to 

size the particles. The material was cascaded through a series of screen separators, each connected 

by 3-in. internal diameter tubes, until the desired material fractions were obtained. The material 

fractions (fines, oversize, product) were collected in individual containers of 3.6-L capacity or less 

below the final screen separator (SNM 696, p. 64). 

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Station E6 – Conversion Crucible Loading. Individual containers of the dried mixture from in-process 

storage or Station E5 were brought to this glovebox and opened, and the material was transferred into 

a conversion crucible. More than one batch of material was sometimes required for a full crucible 

load; however, only one container was handled at a time at this station. All material transfers took 

place in the glovebox, so any dusting was carried off in the facility ventilation and ducting system. 

The station mass limit was 3.6 kg of 235U (SNM 696, pp. 64–65). 

Station E7 – Conversion. Converter crucibles that were loaded in Station E6 were placed in a furnace 

and heated to a maximum temperature of 2,800°C under a high vacuum or an inert gas atmosphere. 

The crucibles were then cooled and removed from a furnace to the Station A13 inert atmosphere 

glovebox for unloading. The crucible inside diameters were 5 in. or less. The station mass limit was 

3.6 kg of 235U (SNM 696, p. 65). 

Station E8 – Screening. Coated particles of uranium-thorium dicarbide were brought to this station in 

a 5-in. inside diameter receiver that contained up to 790 g of 235U, the station mass limit. The material 

was placed in a screen set and shaken into sized fractions. The sized fractions were caught in 

individual containers that were then removed from the station and weighed (SNM 696, p. 65). 

Station F-1 – Rod Batching. A batch of coated uranium-thorium fuel particles was brought to the 

Station in a 5-in. inside diameter containers with a limit of 3.6 kg of 235U. The batch container was 

connected by a 1-in. inside diameter pipe to a rotating metering drum. The metering drum filled with 

fuel particles contained up to a maximum of 20 g of 235U. A conveyor below the metering drum 

extended between two 30-in.-diameter revolving tables. Empty vials were placed on a revolving table 

and were then fed onto the conveyor in a single line. The single line of vials was moved beneath the 

metering drum and the fuel particles in the drum orifices were emptied into the vials (up to 0.4 g of 

235U per vial) as they passed below the drum in synchronization with the drum movement. The loaded 

vials continued on the conveyor to the revolving take-up table. The vials were then manually placed 

into a metal loading board that had 30 locations in a 5-by-6 matrix with 1-in. center-to-center spacing. 

After the board was loaded with vials, it was placed on one of the transfer carts in the station or into 

Station F-2. The station mass limit was 4.2 kg of 235U (SNM 696, pp. 70–71). 

Station F-1A – Alternate Rod Batching. A batch of coated uranium-thorium fuel particles was brought 

to this alternate rod batching station in a 5-in. inside diameter container with a limit of 3.6 kg of 235U. 

The batch container was placed on a fixture with a tube that extended downward into a preset 

metering, mixing, and splitting device. The splitting portion was connected by tubes that extended 

downward to vials on a metal loading board similar to the one at Station F-1. Volume of the tubes, 

device, and board was such that no more than 350 g of 235U were present. When the operation at this 

station was completed, the loading board was placed in a transfer cart or into Station F-2. The station 

mass limit was 4.0 kg of 235U (SNM 696, p. 71). 

Station F-2 – In-Process Storage. The metal loading boards from Stations F-1, F-1A, and F-3 were 

placed in layers that had a minimum vertical spacing of 5 in.; the storage rack was constructed of 

metal and securely attached to a building member (SNM 696, p. 71). 

Station F-3 – Fuel Rod Injection. The loading boards were brought to this station one at a time and 

placed in a fixture that transferred the fuel particles from the vials into a fuel rod die-mold. A mixture 

of pitch and graphite was then injected into the die-mold. The resultant fuel rods were injected into a 

metal loading board dimensionally like the particle loading board. The boards were placed in transfer 

carts or into Station F-2 (SNM 696, p. 72). 

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Station F-4 – In-Process Storage. Containers of fuel rods were placed in the special storage rack 

such that fuel rods in the containers formed a 4-in. planar array. The storage rack was constructed of 

metal and securely attached to a building member (SNM 696, p. 72). 

Station F-5 – Rod Transfer Cart. Containers of fuel rods were placed on a cart for transfer between 

stations. The containers of fuel rods formed a 4-in.-thick slab in a horizontal planar array (SNM 696,

p. 72). 

Station F-6 – Inspection. Fuel rods were brought to the station and subjected to various inspection 

procedures. The operations of the station were such that fuel density in the station did not exceed 

172 g of 235U in any 1 ft2 of aspect area as viewed perpendicular to the surface of the plane of the 

array (SNM 696, p. 72). 

Station F-7 – Fuel Rod Low-Temperature Heat Treating. Transfer carts from Station F-3 were moved 

to this process level, where the rod loading boards were removed from the cart one at a time and the 

fuel rods were placed in graphite boats that were then introduced into a horizontal furnace. The 

furnace was constructed in such a manner that boats could not stack one on top of the other. 

The graphite boats exited the furnace cooling section and were placed in a hood where the rods were 

removed from the boat, cleaned, and placed in a container. As each container was filled with fuel 

rods, it was removed from the station and placed in Station F-4, In-Process Storage, or moved to 

Station F-8, Fuel Rod High-Temperature Heat Treating (SNM 696, p. 73). 

Station F-8 – Fuel Rod High-Temperature Heat Treating. Rods in graphite boats were introduced into 

another horizontal furnace. The furnace was constructed in such a manner that boats could not stand 

one on top of the other. The boats exited from the furnace cooling section, where the rods were 

removed from the boats and placed in other containers (SNM 696, p. 74). 

Station F-9 – Fuel Element Assembly. Fuel rods in containers were brought to this station on the rod 

transfer cart. The fuel rods were inserted into the holes in the graphite element block. Single 

elements were vertically oriented on carts that restricted edge-to-edge spacing between adjacent 

elements to a minimum of 12 in. 

The operations performed at this station included fuel rod loading, whole plugging, plug curing, and 

inspection (SNM 696, p. 74). 

Station F-10 – Shipping Container Loading. Completed fuel elements were prepared for shipping and 

then loaded into the shipping containers. One element was handled at a time. After the element was 

loaded into the shipping container, the container was removed to storage to await shipment. 

Containers were stored in planar arrays no more than three layers thick (SNM 696, p. 76). 

Station F-11 – Fuel Rod Cleaning. Fuel rods were brought to this station in graphite crucibles of 5-in. 

inside diameter or less and placed one crucible at a time, one behind the other, into the conversion 

furnace described as Station A12B. When loaded with crucibles, the furnace was heated and a 

mixture of gases that contained chlorine was introduced into the furnace to accomplish fuel rod 

cleaning through vaporization of heavy metal chlorides. Outlet gas was conducted to a fuel scrubber 

before it was vented (SNM 696, p. 76). 

ATTACHMENT A

FUEL FABRICATION PROCESS

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Recovery Operations. Recovery of unirradiated 235U was performed on GA-licensed and -contracted

SNM, including that for AEC. These operations were performed in the recovery operations areas on 

the main floor and on the South mezzanine of the Fuel Fabrication Facility (see Figures 2-4 and 2-5). 

At no time were GA-licensed and -contracted SNM mixed or combined in a single process. The 

following three recovery processes were performed (SNM 696, p. 160). 

 Conversion to Oxides. Recovery of contained SNM mixed with carbonaceous or other 

materials was done by oxidation. The combustible and volatile materials were burned off and 

residue oxides were recovered in the high-level furnaces in the fluid bed furnace. The oxides 

were sometimes further processed by grinding in the ballmill (SNM 696, p. 168). 

 Uranium and Uranium-Thorium Dicarbide Size Reduction. Oversize dicarbide particles were 

reduced to the desired size range by grinding. This size reduction was performed with the 

ballmill in a controlled-atmosphere glovebox (SNM 696, p. 168). 

 Burning Contaminated Carbon. In the fabrication of uranium carbide fuels, the carbonaceous 

materials and combustible materials associated with the manufacturing became contaminated 

with uranium. This material was reduced to an ash by burning in the low-level furnace and the 

ash was collected for recovery of the uranium. Only dry uranium-carbon or uranium-thorium 

carbon materials were processed in this operation (SNM 696, p. 168). 

Station R1 – In-Process Storage. Storage of in-process material was in a metal storage rack section 

made from module assemblies. Each module consisted of a sheet-metal assembly that was 

fabricated to form shelf units on a common vertical axis. The shelf units were designed to have a 

center-to-center spacing of 18 in. A completed storage rack was made by connecting module 

assemblies to form a two-dimensional plane array with 8-in. surface-to-surface and 16-in. center-tocenter spacing between maximum storage units. Nuclear material was stored in a rack in containers 

with a volume of 3.6 L or less and a mass limit of 3.6 kg of 235U (SNM 696, p. 162). 

Station R3 – High-Level Material Transfer. The transfer of all dry in-process material for the high-level 

furnaces from storage containers to process equipment and from process equipment back into 

storage containers occurred in this glovebox. The glovebox was coupled directly to the loading flange 

of the high-level furnace and thus effectively controlled the spread of dry or dusty materials. The 

glovebox was connected to the facility ventilation duct system and absolute filters. A minimum inward 

air velocity of 150 ft/min was provided by the ventilation system when the loading door of the glovebox 

was open. The station had a mass limit of 3.6 kg of 235U (SNM 696, p. 162). 

Station R4 – High-Level Uranium Oxidation and Reduction. The oxidation of uranium carbide or 

uranium-thorium carbide material of 10% to 90% 235U was carried out in a controlled oxygen-air 

atmosphere. Reduction of the oxidized material was carried out in a hydrogen atmosphere. The 

nuclear material was placed in five steel material trays that were in the central 18 in. of the furnaceheated zone. A tray rack fixed the trays in the 7-in.-diameter liner. Heat was supplied by the 

cylindrical electric heating element. The gas atmosphere of the furnace was supplied through the inlet 

gas tube. Cylindrical blocks of solid silica insulation were placed in each end of the liner to control 

heat loss. 

During operation, the temperature in the furnace was controlled to around 1,800°F during the 

oxidation stage and 1,200°F during the reduction stage. The inlet gas to the furnace was 

ATTACHMENT A

FUEL FABRICATION PROCESS

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automatically controlled, and alarms sounded if there was a failure of the heating element, nitrogen 

gas pressure, or diluting airflow. 

The sequence of furnace operation was as follows: 

1. Oxidation at 1,800°F with air and excess oxygen 

2. Furnace cooling with nitrogen purge gas to room temperature when the oxidation product 

was desired or to 1,200°F when reduction of the oxidized material was desired 

3. Reduction at 1,200°F with hydrogen gas atmosphere 

4. Furnace cooling with nitrogen purge gas to room temperature 

The furnace was mounted horizontally on a steel table with the centerline of the furnace liner about 

36 in. above the floor. A portable hood was provided for servicing the ash filter. The hood was 

ventilated such than a minimum air velocity of 150 ft/min was provided. The mass limit for the station 

was 3.6 kg of 235U (SNM 696, pp. 162–163). 

Station R4A - Fluid Bed Furnace. An alternative fluid bed furnace was used for the oxidation or 

reduction of dry material that contained recoverable SNM. The fluid bed furnace was a vertical 

cylinder with an external heat source. Material to be burnt was brought to the station in a closed 

container with a 5-in. or less inside diameter that was attached to the feed mechanism of the furnace. 

As material was fed into the furnace, it was levitated into the hot zone by air-oxygen gas. When the 

burning was completed, the impure oxide was removed from the furnace or in some cases reduced by 

the introduction of hydrogen gas to the furnace. The station mass limit was 4.0 kg of 235U (SNM 696,

p. 164). 

Station R4B – Calcining. Calcination of uranium precipitate took place at this station. The precipitate 

was changed to each furnace action in a boat with the capacity of no more than 1.6 L and a maximum 

uranium capacity as precipitate of no more than 750 g. This precipitate was heated to 900°C in air to 

oxidize the uranium to U3O8. The product was cooled and removed from the boats and placed in 

polyethylene bottles with a volume of 3.6 L or less. Each bottle held a maximum of 3 kg of 

unscreened U3O8. The station mass limit was 3.6 kg of 235U. The boats were loaded, charged to the 

furnace, unloaded, and emptied one at a time to preclude the possibility of stacking loaded boats on 

top of one another (SNM 696, p. 165). 

Station R5 - Low-Level Uranium Oxidation. Oxidation of uranium dicarbide material of 10% 235U or 

less was carried out in an excess oxygen atmosphere similar to that at Station R4. The material 

charged to the low-level furnace was limited to 300 g of 235U or less. A combustible charged container 

was loaded and placed in the furnace. No external source of furnace heating or cooling was required 

at the station. The charge material was ignited inside the furnace, the furnace closed, and oxygen 

supplied through the nozzle to support combustion and provide the excess oxygen atmosphere. The 

low-level furnace was mounted inside the hood box. The discussions at Station R4 for hood 

ventilation, inlet gas control, and only gas handling were directly applicable to this station. 

After each batch of material was burned, the furnace was cooled and the ash was removed and 

placed into a polyethylene bottle with a volume of 3.6 L or less. The furnace was checked visually 

and with the counter for any accumulation of ash. If the counter indicated an accumulation of 50 g of 

ATTACHMENT A

FUEL FABRICATION PROCESS

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235U in the furnace, the furnace was torn down and rebuilt. The relationship between accumulated 

235U and the counter reading was established by calibration. Mass limit for the station was 350 g of 

235U or less (SNM 696, p. 165). 

Station R6 - Size Reduction. The reduction of the particle size of dry uranium-carbon material was 

carried out in a sealed glovebox provided with a controlled gas atmosphere. Nuclear material was 

brought to the station a single sealed polyethylene bottle with the limit of 3.6 kg of 235U or less and a 

capacity of 3.6 L or less. 

The nuclear material was ground in a standard ballmill. Product from the ballmill was screened and 

divided into specification and off-specification material. The screened material was placed into 

separate bottles with limits of less than 3.6 kg of 235U and 3.6 L (SNM 696, pp. 165–166). 

Station R7 – Grinding. Dry uranium-bearing material was brought to the station in a container of 3.6-L 

capacity or less with up to 3.6 kg of 235U, the station mass limit. The material was fed to an enclosed 

grinding mill with a total volume, including the feed funnel and catch receiver, of less than 3.6 L (SNM 

696, p. 166). 

Station R17 – Higher Level Uranium Oxidation. An alternate station for burning uranium-bearing 

materials consisted of four horizontal furnaces in a planar array on a 16-in. center-to-center spacing. 

The dimension of each furnace was 2.5 in. high by 8 in. wide by 14 in. deep. This was brought to the 

station in a container with a maximum volume of 3.6 L and up to 3.0 kg of 235U, the station mass limit. 

The material was charged to each furnace action in a boat with up to 750 g of 235U. No more than one 

boat that actually contained material was outside a furnace at any one time. After the material was 

burned and cooled, the boats were removed one at a time and the contents were emptied into a 

container with a volume of 3.6 L or less (SNM 696, p. 166). 

Station R18 – Blending. Dry uranium-bearing material was brought to the station in a container with a 

volume limit of 3.6 L and containing up to 3.6 kg of 235U, the station mass limit. The material was fed 

into blender with a volume of 3.6 L or less (SNM 696, p. 167). 

Station R19 – Grinding. Dry uranium-bearing material was brought to the station in a container with 

limits of 3.6 L and 3.6 kg of 235U. The material was fed to a hopper and grinder with a total volume 

less than 3.6 L. The ground material passed through a pipe through the bottom of the glovebox to a 

sealed receiver hopper with a volume limit of 3.6 L or less (SNM 696, p. 167). 

ATTACHMENT A

FUEL FABRICATION PROCESS

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Station P1 – In-Process Storage. SNM in-process was stored in containers with limits of 3.6 L and 3.6 

kg of 235U. Each storage location of the in-process storage rack was limited to a single container. 

Storage locations were on 16-in. centers (horizontal and vertical) with a minimum of 8-in. surface 

spacing between containers and adjacent locations. The storage racks were constructed of metal and 

securely fastened to the building structure (SNM 696, as renewed, p. 193). 

Station P9 – In-Process Storage. Liquids from the various steps of the uranium purification process 

were stored at the station. These liquids might have been from the dissolution step, organic in 

aqueous extraction steps, or the precipitation and filtering steps. The uranium content of liquids that 

were transferred to this station was determined by chemical analysis or gamma accounting before 

storage. 

Storage containers at this station were 5-in. inside diameter cylinders. Each cylinder was secured in a 

specially constructed metal rack with a minimum 12-in. surface spacing between adjacent cylinders 

(SNM 696, p. 193). 

Station P2 – Weighing. A single closed container of uranium-bearing scrap material was brought to 

this station and weighed into a batch container. The balance was enclosed in a hood box with a 

vertical opening front panel. The hood was connected to facility ventilation duct and absolute filters. 

When the front panel was opened, a minimum inward air velocity of 150 ft/min was provided. 

Nuclear material was handled at the station in two containers each with a volume of no more than 

3.6 L and up to 3.6 kg of 235U. A fixture was provided to space raw material containers with at least 

16-in. center spacing and 12-in. surface spacing from the balance. The batch container was placed 

on the balance of the raw material transferred and weighed into the batch container. More than one 

raw material container might have been needed to make up the batch; however, the first container 

was emptied and removed before a second container was brought to the station (SNM 696, pp. 194–

195). 

Station P3 – Dissolution. A weighed batch of raw material was brought to the station and placed in a 

special container on an electric hot plate. The container was made of special acid-resistant material 

and had a volume no greater than 5 L. While more than one weighed batch of material might have 

been brought to this station at any time and placed in the dissolution container, the total combined 

batch size was limited to 3.6 kg of 235U, the station mass limit. An acid solution was slowly added to 

the raw material in the container. The mixture was heated and allowed to react until all the raw 

material was dissolved or the solids and liquid reached equilibrium. A special temperature control 

was provided on the hot plate to prevent damage to the dissolution container or boiling over of the 

acid solution. However, if either event did occur, a 5-L catch pan in which the dissolution container 

sat prevented spilling. The dissolution station was fully enclosed in a special acid hood and equipped 

with ducts and filters suitable for handling any acid fumes that were generated. Tell-tales that 

indicated the amount of air draft through the hood were installed to indicate plugged duct system 

filters. 

If material residue remained in the dissolution container, the entire contents were filtered to separate 

the residue and the solution. The filter and filtrate receiver were each limited to a volume of 5 L and 

there was no more than 3.6 kg of 235U at the station. Filtrate from this station was transferred either to 

Station P4 or liquid storage in Station P9 (SNM 696, pp. 195–196). 

ATTACHMENT B

URANIUM (AND THORIUM) SCRAP PROCESSING

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Station P3A – Dissolution. Weighed batches of uranium-bearing material and up to two 5-in. inner 

diameter scene hoppers were brought to this alternate dissolution station one at a time and attached 

to two vertical 5-in. inner diameter cylinders on 17-in. center-to-center spacing and 12-in. surface-tosurface spacing. The total combined batch size was limited to 3.6 kg of 235U, the station mass limit. 

An acid solution was added to the columns, and heat was applied until dissolution was complete. 

The dissolved material was then pumped from the dissolver cylinders to an adjacent 5-in. inner 

diameter pressure vessel cylinder where it was pressurized to an adjacent filter to remove the solids 

from the solution. The filter and a volume envelope that was 5 by 5 by 16.5 in.. The filtered solution 

was accumulated in an adjacent 5-in. inner diameter cylinder, then transferred to Station P9, InProcess Storage (SNM 696, p. 197). 

Station P4 – Extraction. The dissolved material from Station P3 was moved to this extraction of 

station one container at a time and transferred to the first stage of the extraction unit. More than one 

dissolution batch might have accumulated in the first extraction stage, but the unit and a total volume 

limit of 5 L and the total extraction batch size was limited to 3.6 kg of 235U, the station mass limit. Up 

to three steps of extraction were carried out at this station – organic extraction of dissolved uraniumbearing scrap, aqueous extraction of the impurities from the organic phase, and an aqueous 

extraction of the uranium from the organic phase. For each type of extraction several stages of 

contacting and separation might have been required. 

Each extraction unit was identical and had a total unit volume of 5 L. Contacting of solutions in the 

extraction unit was accomplished by mechanical or other comparable means of agitation. Phases 

were transferred from one unit to another or two liquid waste containers by gravity flow. Individual 

extraction units were securely mounted to a metal rack that was securely mounted to the building. 

Dissolved material was transferred into the first unit, which contained a premeasured amount of the 

organic extraction agent. The solutions were contacted in the mixture and allowed to stand and 

separate. The uranium selectively went into the organic phase. Separation was accomplished by 

draining off the aqueous phase and leaving the uranium-containing organic phase in the first unit. 

As the aqueous phase flowed from the first extraction unit, it flowed directly into a second unit that 

contained a second premeasured amount of organic extracting agent. The second extraction was 

completed with the aqueous phase sometimes being transferred to a third extraction unit with a 

premeasured amount of the organic extracting agent. This step was repeated as required. When the 

final extraction step was completed, the aqueous phase stripped of uranium was sampled; the sample 

was gamma-counted and the aqueous solution was transferred to liquid storage. 

From each extraction stage, the organic phase that contained the uranium was transferred and 

accumulated in a single extraction unit. At this point in the process, all of the medium that was initially 

charged to the extraction station, except for a trace amount that might have been present in the waste 

aqueous solution, should have been present in a single extraction unit. The second type of extraction 

began at this step with the addition of a premeasured amount of an aqueous scrubbing solution. The 

solution was contacted in the mixture and was then allowed to stand and separate into the light 

organic phase and the denser aqueous phase. The uranium material remained in the organic phase. 

Separation was performed by draining the aqueous phase into a second extraction unit. 

The organic phase that remained in the first extraction unit was contacted a second time with another 

charge of the aqueous scrubbing agent. The aqueous phase was drained into a second extraction 

ATTACHMENT B

URANIUM (AND THORIUM) SCRAP PROCESSING

Page 2 of 4

ATTACHMENT B

URANIUM (AND THORIUM) SCRAP PROCESSING

Page 2 of 4
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 121 of 143

unit and added to the material obtained from the first extraction stage. Sometimes a third extraction 

was performed on the organic solution with the aqueous phase being combined with that of the 

previous extractions. The accumulated aqueous solution was transferred from the second extraction 

unit into the third unit and a premeasured amount of organic extraction agent was added. The 

solution was contacted and allowed to separate. Any uranium that was extracted by the aqueous 

scrubbing solution selectively went into the organic phase. The aqueous phase, stripped of uranium, 

was separated and sampled, and the sample gamma counted in the solution was transferred to a 

liquid storage container at Station P9. The organic phase was drained and added to the organic 

phase from the first extraction unit. 

The clean organic solution in the first extraction unit contained all the uranium that was charged to the 

extraction station except a small amount that might have been present in the waste aqueous 

solutions. The aqueous extraction of the uranium from the organic phase began at this step with the 

addition of a premeasured amount of the aqueous solution. Contacting of the solutions was done and 

the material was allowed to stand and separate into the light organic phase and the denser aqueous 

phase. The uranium material selectively went into the aqueous phase. Separation was accomplished 

by draining the aqueous phase into the second extraction unit. The above steps were repeated until 

all the uranium, except for trace amounts, was contained in the new aqueous solution. This aqueous 

solution was then placed in one of the 5-L extraction containers (SNM 696, pp. 199–200). 

Station P5 – Precipitation. A single container of the aqueous phase that contained the purified 

uranium in solution was brought to this station from the final stage of the extraction process. The 

container of product from Station P4 was limited to 5 L and no more than 3.6 kg of 235U, the station 

mass limit. The container of solution was placed on a hot plate, heated, and maintained at a 

temperature of about 90°C. At this point, a basic reagent was added that resulted in the precipitation 

of the uranium. 

A special temperature control was provided on the hot plate to prevent damage to the container or 

boiling over of the solution. The precipitation station was fully enclosed in a special acid hood with 

ducts and filters suitable for handling the acid fumes that were generated (SNM 696, p. 200). 

Station P6 – Filtration. A container of aqueous solution with a precipitated uranium salt was brought 

to this filtration station. The uranium was separated from the aqueous solution by vacuum filtration. 

While more than one container with the precipitated material was allowed at this station at any given 

time with their contents transferred to the filter, the total batch accumulated in the filter was limited to a 

volume of 5 L and 3.6 kg of 235U, the station mass limit. The filtrate was transferred into a liquid 

storage container at Station P9. 

The filter cake that contained the uranium precipitate was air dried, removed from the filter, and 

placed in a close container with a maximum volume of 3.6 L. While it was possible that more than 

one container was required for a batch of the filter material, no more than one container was handled 

at this station at a time and it was limited to 3.6 kg of 235U. As each container was loaded, it was 

closed and placed in a storage location in an in-process storage rack at Station P1. The material was 

taken from Station P1 to be calcined at Station R4B (SNM 696, pp. 201–202). 

Station P10 – Large Batch Extraction-Precipitation. At this station to dissolve batch material from 

Station P9 with up to 3.6 kg of 235U was pumped to the first of nine vertically oriented columns in a 

single planar array. The columns had a length of up to 11 ft and inside diameter of 5 in. in a surfaceto-surface spacing of 12 in. 

ATTACHMENT B

URANIUM (AND THORIUM) SCRAP PROCESSING

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Organic extracting agents were pumped to the first column, and the solution was then circulated 

between the first and second columns through a 2-in. inner diameter pipe that was connected to the 

top until extraction was completed. 

The organic phase that contained the extracted uranium was pumped from the first two columns to 

the third column where scrubbing agents were added to remove impurities. 

The pregnant organic phase was then pumped from the third column to the fourth column where 

scrubbing agents were added to separate the uranium from the organic phase. 

The resultant pregnant aqueous phase was then pumped from the fourth column to the first column 

where washing agents were added to remove any residual impurities. Agitation of the solution was 

accomplished in the third fourth and fifth columns by air sparging. 

The pregnant aqueous phase was then pumped from the fifth column to one or more precipitation 

columns where a basic reagent was added that resulted in the precipitation of the uranium. 

The discard solution was drained from the bottom of each column into an 11-L bottle (SNM 696, pp. 

202–203). 

Station P6A – Filtration. The aqueous solution with a precipitated uranium salt and the precipitation 

columns was pumped to this alternate filtration station. The uranium was separated from the aqueous 

solution by a filter method. A total accumulated batch on the filter did not exceed 4.8 L, the station 

volume limit, or 3.6 kg of 235U, the station mass limit. The aqueous discard was drained to a critically 

safe container. No more than one batch was processed through the filter at one time and all solids 

and liquids were removed from the station before a second batch was filtered. The uranium in the 

filter cake was placed in a container with a maximum volume of 3.6 L and transferred to Station P11, 

Calcining (SNM 696, p. 203). 

Station P11 – Calcining. The filter cake from Station P6A was brought to this station in a container 

with a maximum volume of 3.6 L and up to 3.0 kg of 235U, the station mass limit. This station 

consisted of four horizontal furnaces oriented in a planar array on 16-in. center-to-center spacing. 

The dimensions of each furnace were 3.5-in. high by 8 in. wide by 8 in. long. 

The filter cake was charged to each furnace section in a boat with up to 750 g of 235U. No more than 

one boat that actually contained material was outside the furnace at any time. After the material was 

burned the boats were removed one at a time and transferred to Station P12, Grinding-SizingBlending (SNM 696, p. 204). 

Station P12 – Grinding-Sizing-Blending. Uranium-bearing material was brought to this station in a 

container with a maximum volume of 3.6 L and up to 3.6 kg of 235U, the station mass limit. The 

material then underwent separate grinding, sizing, blending, or a combination (SNM 696, p. 204). 

ATTACHMENT B

URANIUM (AND THORIUM) SCRAP PROCESSING

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ATTACHMENT C 

ALLOWABLE QUANTITIES OF RADIOACTIVE MATERIAL AT GENERAL ATOMICS 

Page 1 of 3 

Table C-1 provides a snapshot of the allowable quantities of radioactivity at GA. The SNM licenses 

were in effect before 1973. Radionuclides and allowable quantities might have changed over time 

but, based on site descriptions, health physics reports, and inventory reports, the materials in 

Table C-1 seem generally representative of the variety and the upper limits on source term during the 

covered period. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 124 of 143

ATTACHMENT C 

ALLOWABLE QUANTITIES OF RADIOACTIVE MATERIAL AT GENERAL ATOMICS 

Page 2 of 3 

Table C-1. Summary of licensed source terms. 

Date License Material Description Form Quantity Total Use

ca. 1960 AEC License No. 4-1611-

11 (J-60)

Byproduct materials, atomic 

numbers from 3 through 83

Any 10 Ci each 500 Ci

H-3 200 mCi

Sr-90 Sealed 1 Ci

Sr-90 Unsealed 100 mCi

Po-210 Sealed 50 Ci

Po-210 Unsealed 100 mCi

Tm-170/171 50,000 Ci

Dec. 1960 AEC License No. 4-1611-

11 (J-60), proposed 

increase

Byproduct materials, atomic 

numbers 3 through 83

Any 100,000 Ci each 500,000 Ci

Proposed addition Fission products x 1,000,000 Ci

1973 SNM-696, Amendment No. 

45

U Up to fully 

enriched in U-235

985 kg Fabrication of fuel materials 

and fuel elements

U Up to 100% 

enriched in U-235

9.5 kg Fabrication of fuel materials 

and fuel elements

U Up to fully 

enriched in U-233

2.5 kg Fabrication of fuel materials 

and fuel elements

Pu 3.0 kg Fabrication of fuel materials 

and fuel elements

ca. 1973 SNM-69 U-235 Up to fully 

enriched

Metal, metallic alloy, or 

compounds

1,600 kg

U-235 Up to 100% 

enriched

Metal, metallic alloy, or 

compounds

10 kg

U-235 100% Metal, metallic alloy, or 

compounds

2.5 kg

U-236* 100% Metal, metallic alloy, or 

compounds

300 g *As residue part of material 

enriched U-235 

Pu 100% Metal, metallic alloy, or 

compounds

13 g

ca. 1973 SNM-696 U-235 Fully enriched Metal, metallic alloy, or 

compounds

2,500 kg

U-236* 100% Metal, metallic alloy, or 

compounds

1 kg *As residue part of material 

enriched U-235 

U-233 Fully enriched Metal, metallic alloy, or 

compounds

500 g

U-233 Encapsulated Metal, metallic alloy, or 

compounds

10 g

U-235 100% Metal, metallic alloy, or 

compounds

10 kg

c. 1973 State of California Any radioactive material 

(except source and SNM)

Any 30 Ci each nuclide 1,500 Ci
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 125 of 143

ATTACHMENT C 

ALLOWABLE QUANTITIES OF RADIOACTIVE MATERIAL AT GENERAL ATOMICS 

Page 3 of 3 

Date License Material Description Form Quantity Total Use

Any radioactive material or 

activated material (except 

source and SNM)

Any 100,000 Ci each 

nuclide

500,000 Ci

Thorium (natural) Any 200,000 lb

Uranium (natural or depleted) Any 400,000 lb C-14 Sealed sources 50 mCi

Cs-137 Sealed sources 40 Ci

Co-60 Sealed sources 20 Ci per source

Sr-90 Sealed sources4,800 Ci

Ir Sealed sources

Po-210 Sealed sources 50 Ci

Ra-226 Sealed sources 47 mCi

U-235 Sealed sources 50 g

Pu-238 and 239 Sealed sources30 g

Am-241 Sealed sources 0.1 Ci

Cf-252 Sealed sources 70 Ci

Co-60 Metal discs 100,000 Ci

ca. 1973 AEC R-38 (TRIGA Reactor)U-235 SNM in fuel elements 

and experiments

5 kg

ca. 1973 AEC R-67 (TRIGA Reactor)U-235 SNM in fuel elements 30 kg

Pu SNM in sources 1 g U-235 SNM in experimental 

elements

300 g

ca. 1973 AEC R-100 (TRIGA 

Reactor)

U-235 SNM in fuel elements 10 kg

Pu SNM in fuel elements 30 g

Po-210 Be Sealed sources 15 Ci

Sb-124 Be Sealed sources 150 Ci

U-235 SNM in experimental 

elements

806 g
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 126 of 143

ATTACHMENT D 

LABORATORY AND FACILITY THROUGHPUT INFORMATION FOR 1972 

Page 1 of 3 

Table D-1 provides a snapshot of the SNM that might have been used in the Science Laboratory in 

1972. It is assumed that throughput means the amount of material used in 1972, and that the total 

amount was the amount on hand at the end of 1972. 

Table D-1. Science Laboratory 1972 annual inventory information.a

Location 1972 throughput (g) 1972 total amount (g)

Laboratory U-235 U-233 Pu U-235 U-233 Pu

L-122 200 90

L-134 100 100

L-137 3,000 499

L-139 250 1,000 10

L-141 3,000 499

L-143 3,000 499

L145 <200 Not to exceed 100

L-147 4,000 499

L-154 4,000 499

L-218 80 1 1 10 0.1 0.1

L-230 90 500

L-232 4,000 499

1,000 0.20 0.02 360 0.20 0.02

L-234 4,000 <1

1,150 499

1,000 499

0.20 0.02 350 0.20 0.02

L-236 4,000 499

1,150 499

1,000 0.20 0.02 350 0.20 0.02

L-240 1 1 10-310-4

3,000 499

L-242 7 100

1,000 100 <1 10

L-307 <200 Not to exceed 100

7 1,000

250 3 350 10

1,000 0.20 0.02 <1 0.20 0.02

L-311 250 1,000 10

L-313 1 0.1 10-310-4

3,000

L-315 4,000 499

L-317 1 0.1 10-310-4

3,000 499

L-331

L-333

L-347 through 

L-361

80 80

L-359 and L-361

L-401 and L-403 4,000 499

L-405 4,000 499
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 127 of 143

ATTACHMENT D 

LABORATORY AND FACILITY THROUGHPUT INFORMATION FOR 1972 

Page 2 of 3 

Location 1972 throughput (g) 1972 total amount (g)

Laboratory U-235 U-233 Pu U-235 U-233 Pu

L-407 4,000 499

1,000 0.20 0.02 350 0.20 0.02

L-409 4,000 499

L-411 4,000 499

L-413 3,500 350

250 0.500 1,000 10

L-415 250 1,000 10

L-417

L-419

L-421

500 500

L-425

L-427

500 6,000

L-429 4,000 500

<1

L-441

L-443 700 350

L-530

L-540 7 <1

5

L-543 3,000 350

L-545 50 10

L-548 50 10

L-550 50 10

L-552 50 10

L-554 3,000 300

L-556 3,000 300

7

25 300

L-558 3,000 300

L-635

L-637

L-639

300 600

L-647 3,000 350

Activation 

Analysis

160 g

270 kg 20 g 220 g 270 kg 20 g 220 g

8,000 g 350 g

10 kg 5 g 400 g 5 g

Experimental 

Building

E-141

50,000g 50,000 g

40 curies 40 curies

3,000 g 350 g

TRIGA <200 g Not to exceed 100 g

3 kg 350 g

3,000 g 499 g

190 g 1 g 1 g 900 g 5 g 5 g

190 g <190 g
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 128 of 143

ATTACHMENT D 

LABORATORY AND FACILITY THROUGHPUT INFORMATION FOR 1972 

Page 3 of 3 

Location 1972 throughput (g) 1972 total amount (g)

LINAC

Bldg. 512

10-7g

0.15 g 600 kg

<1.5 g 725 g <1.5 g

15 kg 1000 g 1,000 g 725 g 299 g 299 g

604 g 604 g

Waste Yard 24 kg 50 g 10 g

24 kg

5,000 g 4,000 g

15,000 g 500 g

SVA-Tunnel 

Glovebox

3.6 kg NAb

SVA-Quality 

Control/Fuels

NA 350 g

350 g

SVB 1,000 g 350 g

2,500 g 700 g

3,600 g 1,750 g

500 g 5,000 g

5 kg 500 g

350 g 350 g

Hot Cell 2,000 g 100 g 100 g 2.1 kg 100 g 100 g

<200 g Not to exceed 100 g

Hot Cell 

Radiography 

Room

5 g 100 g

HTGR 87 kg 2.52 kg

Convoy Court 350 g 10 kg

Machine Shop 350 g 10 kg

Barrel Storage 

Yard

700 kg NA

a. GA (1972). 

b. NA = not available. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 129 of 143

ATTACHMENT E 

EXAMPLES OF ACTIVITIES AND AIR MONITORING IN THE PLUTONIUM LABORATORY 

Page 1 of 2 

Table E-1 provides information on work and associated monitoring in the Plutonium Laboratory. No 

evidence of plutonium bioassays have been found to date. Well after 1970, lung counts might have 

been able to detect plutonium or 241Am. 

Table E-1. Plutonium Laboratory workplace air sample results. 

Period

Alpha

(μCi/ml)

Beta

(μCi/ml) Comments

November 

1968

<5E-15 <6.4E-13 On November 8, a survey following classified Pu-contaminated fuel 

particle work indicated <15 pCi beta removable and <100 cpm alpha 

on a PAC-1A air proportional survey meter, but on November 21, a 

masslinn sweep detected 800 pCi beta on the floor in the room behind 

the LLC.

a

October 1968 <3.6E-14 <7E-15 A masslinn sweep on October 16 detected 4,000 cpm tactile 

contamination on the floor that was subsequently decontaminated to 

<15 pCi beta and <3 pCi alpha.

b

January to 

September 

1968

No operations involving Pu were performed. 

December 

1967

No operations involving Pu. Continuous air samples showed no 

detectable airborne radioactivity. 

November 

1967

Limited operations in the Plutonium Laboratory.

October 1967 Sintering, sectioning, and counting graphite Pu samples in glovebox 

complex continued throughout month. No samples were removed 

from “enclosures” in October. Air and wipe samples indicated “no 

detectable” contamination.c

September 

1967

Sintering, sectioning, and graphite Pu sample counting was performed 

in glovebox complex. No samples were removed from “enclosures” in 

October. Air and wipe samples indicated “no detectable” 

contamination.d

August 19 to 

September 1, 

1967

No operations involving Pu were performed. 

August 1 to 

August 18, 

1967

Very little work was performed. Air and wipe sampling indicated no 

“significant airborne or tactile” contamination.

April 1967 No Pu work was performed during this period.

February 1967 No Pu work was performed during this period. Two g of Pu were still in 

storage in one of the gloveboxes. Periodic masslinn wipes indicated 

no tactile contamination. 

January 1967 2.1E-13 8.8E-14 Four g of Pu-239 O2 was separated into 2-g lots with 1 g going to 

Science Building. Material removed from the gloveboxes was wiped 

and found free from significant contamination.e 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 130 of 143

ATTACHMENT E 

EXAMPLES OF ACTIVITIES AND AIR MONITORING IN THE PLUTONIUM LABORATORY 

Page 2 of 2 

Period

Alpha

(μCi/ml)

Beta

(μCi/ml) Comments

December 

1966

PuO2 was dissolved in acid. Wipe and air samples collected during 

and following the operation indicated no significant transfer of activity. 

100 ml of Pu-239 solution was removed and double-bagged and 

checked for contamination. A wipe sample of the container taken 

inside No. 2 glovebox showed an activity level of 200 cpm gamma 

through the envelope containing the filter paper. The container was 

removed from glovebox No. 3 into two double bags and placed inside a 

heavy receiving bag. The pass-through bag was then removed and 

placed inside the Chemistry hood. The flask inside the container was 

then removed in the hood and placed inside double bags. The outside 

of all bags removed from the plutonium facility were wipe-checked and 

found free of alpha contamination. The gamma level was <1 mR/hr at 

the surface.f 

November 

1966

The Plutonium Laboratory was activated for storage of 115 mg of PuO2

in No. 1 glovebox. Both inside and outside containers were checked by 

wipe sampling and found free of contamination.g 

October 1966 All gloveboxes were pressure-tested, absolute filters were in the 

exhaust system, and radiation detection instruments were in place. 

The area was free of alpha and beta contamination. U-232 would be 

the first isotope used in the facility.h 

a. Cockle (1968a). 

b. Cockle (1968c). 

c. Kesting (1967d). 

d. Kesting (1967e). 

e. Gurren (1967c). 

f. Gurren (1967d). 

g. Gurren (1966d). 

h. Gurren (1966a). 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 131 of 143

ATTACHMENT F 

LINAC RADIOACTIVE MATERIAL PRODUCTION INFORMATION 

Page 1 of 2 

The source term in the LINAC was as varied as the targets that were irradiated. Table F-1 shows 

estimated production rates for the 45-MeV accelerator operating at 25-MeV electron energy (Author 

unknown, undated). The following was assumed: 

 Accelerator is operating at 25 MeV. 

 Tungsten converters are used to produce bremsstrahlung. 

 Estimates, except where noted, are based on a 1-g sample 1.5 in. in front of the photon 

converter.

 Average dose in the 1-g sample is ~80% of the peak surface dose. 

 Activities are given as of shutdown. 

 Actual production conditions had not yet been explored. 

Table F-1. LINAC radioactive material production information (Author unknown 

1970, pp. 4–7). 

Radionuclide Half-life

Irradiation 

timea

Activity 

produced Accuracyb

Pile isotope 

available

Sample 

materialc

H-3 12.46 yr 1 hr 104 pCi 1 Yesd He

C-11 20.4 m S 415 mCi 1 No C

N-13 9.93 m S 208 μCi 1 No N2

N-17 4.14 s S 3.6 μCi 1 No O2

O-15 118 s S 156 mCi 1 No O2

F-17 70 s S 3.6 mCi 1 No LiF

F-18 112 m 1 hr 78 mCi 1 No LiF

Ne-19 18.2 s S 386 μCi 1 No Ne

Na-22 2.60 yr 1 hr 21 μCi 1 No Na

Na-24 15.1 hr 1 hr 2.4 mCi 1 Yesd Na

Na-25 60 s S 47 mCi 1 No Na

Mg-23 11.9 s S 300 mCi 1 Yes Mg

Al-26 6.5 s S 47 mCi 1 No Al

Al-28 2.27 m S 47 mCi 1 Yes Si

Al-29 6.56 m S 26 mCi 1 No Si

Si-27 4.9 s S 31.2 mCi 1 No Si

P-30 2.55 m S 571 mCi

10 mCi

1 No P

S

S-31 3.2 s S 364 mCi 1 No S

Cl-39 55.5 m S 2 mCi 1 No Ar

Ar-39 265 yr 1 hr 520 pCi 1 Yes Ar

Ca-39 1.06 s S 200 mCi 1 No Ca

Sc-47 3.43 d 1 hr 67 μCi 1 No V

V-52 2.6 m S 26 mCi 1 Yes Cr

Cr-49 419 m 1 hr 31 mCi 1 No Cr

Mn-54 310 d 1 hr 182 μCi 1 No Mn

Fe-53 8.9 m S 83 mCi 1 No Fe

Fe-55 2.94 yr 1 hr 47 μCi 1 YesdFe

Co-57 270 d 1 hr 26 μCi 1 No Ni

Co-58 72 d 1 hr 780 μCi 1 Nod Co

Co-61 99 m 1 hr 2 mCi 1 No Cu

Ni-57 36 h 1 hr 16 mCi 1 No Ni
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 132 of 143

ATTACHMENT F 

LINAC RADIOACTIVE MATERIAL PRODUCTION INFORMATION 

Page 2 of 2 

Radionuclide Half-life

Irradiation 

timea

Activity 

produced Accuracyb

Pile isotope 

available

Sample 

materialc

Cu-61 3.33 hr 1 hr 7 mCi 1 No Cu

Cu-62 10.1 m S 1.2 Ci 1 No Cu

Cu-64 12.8 hr 1 hr 68 mCi 1 Yesd Cu

Zn-63 38.3 m S 1.3 Ci 1 No Zn

As-73 76 d 1 hr 125 μCi 1 No As

As-74 17.5 d 1 hr 3.9 mCi 1 No As

As-77 38 hr 1 hr 18 μCi

16 μCi

1

1

No

Yesd

Br

U

Br-78 6.4 m S 830 mCi 1 no Br

Br-80 18 m S 935 mCi 1 Yes Br

Br-80m 4.6 hr 1 hr 57 mCi 1 Yes Br

Br-83 2.33 hr 1 hr 208 μCi

1.9 mCi

1

1

No

Yes

Rb

U

Rb-86 29.5 d 1 hr 1.5 mCi 1 Yesd Rb

Sr-85 65 d 1 hr 93 μCi 1 Yes Sr

Sr-89 53 d 1 hr 39 μCi 1 Yesd U

Y-88 2.0 hr 1 hr 52 mCi 1 No Y

Zr-89 79.3 hr 1 hr 21 mCi 1 No Zr

Zr-89m 4.4 m S 2.1 Ci 1 No Zr

Zr-97 17 hr 1 hr 3.8 mCi 1 Yes U

Nb-92 21.6 hr 1 hr 88 mCi 1 No Nb

Nb-97 72 m 1 hr 52 mCi 3 No Mo

Nb-99 2.5 m S 10 mCi 1 No Mo

Mo-91 75 s S 42 mCi 1 No Mo

Mo-91m 15.5 m S 210 mCi 1 No Mo

Mo-99 67 hr 1 hr 4.2 mCi

880 μCi

1

1

Yes

Yesd

Mo

U

Ru-103 39.8 d 1 hr 36 μCi 1 Yes” U

Ru-105 4.5 hr 1 hr 5.5 mCi 1 Yes U

Rh-102 210 d 1 hr 390 μCi 1 No Rh

Rh-105 36.5 hr 1 hr 0.21 μCi 1 No Ag

Ag-105 40 d 1 hr 100 μCi 3 No Ag

Ag-106 24 m S 1.6 Ci 1 No Ag

Ag-108 2.3 m S 1.6 Ci 1 Yes Ag

Ag-111 7.6 d 1 hr 21 μCi 1 Yesd U

Ag-112 3.2 hr 1 hr 10 mCi 3 No Cd

In-113m 104 m 1 hr 286 mCi 1 No In

In-114 72 s S 730 mCi 1 YesdIn

In-114m 49 d 1 hr 2.9 mCi 1 YesdIn

In-115m 4.5 hr 1 hr 2.3 mCi 1 No In

In-117 2.5 hr 1 hr 5.5 mCi 3 No Sn

Sb-120 16.4 m S 1.6 Ci 2 No Sb

Sb-122 13.0 d 1 hr 21 mCi 1 Yesd Sb

I-126 33 m 1 hr 5.7 mCi 1 No I

Cs-138 85 m S 94 mCi 1 Yes U

Ba-139 33 hr 1 hr 31 mCi 1 Yes U

Ce-143 33 hr 1 hr 1.6 mCi 1 Yes U

Ta-179 600 d 1 hr 21 μCi 1 No Ta

Ta-180 8.15 hr 1 hr 310 mCi 1 No Ta

Au-195 180 d 1 hr 36 μCi 3 No Au

Au-196 5.55 d 1 hr 16 mCi 2 No Au

Au-197m 7.4 s S 21 mCi 2 No Au

Tl-207 4.8 m S 2.1 mCi 1 No Pb

Bi-207 50 yr 1 hr 0.6 μCi 1 No Bi

Th-231 25.64 hr 1 hr 114 mCi 1 No Th

U-237 6.75 d 1 hr 16 mCi 1 No U

a. Either 1 hour (1 hr) or Saturation (S) 

b. 1 = ± 50%, 2 = within factor of 3, 3 = within factor of 10 

c. 1 g of natural element unless noted 

d. Listed in Oak Ridge Isotope catalog, January 1957 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 133 of 143

ATTACHMENT G 

RADIOLOGICAL LABORATORY HAZARD CLASSIFICATION SCHEME 

Table G-1 lists the classifications from “United States of America Standards Institute (USASI) Design 

Guide for Radioisotope Laboratory,” which provided guidelines for radiochemistry laboratories based 

on the four classifications of relative radiotoxicity: very high, high, moderate, and slight (Bold 1970a). 

Recommended maximum activities for each laboratory were given. These guidelines were used at 

GA to assess conditions in the Science Laboratory. 

Table G-1. Classification of isotopes according to relative radiotoxicity per unit of activity.

Class 1

(very high radiotoxicity

Sr-90 + Y-90, Pb-210 + Bi-210 (Ra D + E), Po-210, At-211, Ra-226 + 55%

progeny, Ac-227, U-233, Pu-239, Am-241, Cm-242, plus other transuranic 

isotopes 

Class 2

(high radiotoxicity)

Ca-45, Fe-59, Sr-89, Y-91, Ru-106, I-131, Ba-140 + La-140, Ce-144 + Pr-144, 

Sm-151, Eu-154, Tm-170, Th-234 + Pa-234, natural uranium

Class 3

(moderate radiotoxicity)

Na-22, Na-24, P-32, S-35, Cl-36, K-42, Sc-46, Sc-48, V-48, Mn-52, Mn-54, 

Mn-56, Fe-55, Co-58, Co-60, Ni-59, Cu-64, Zn-65, Ga-72, As-74, As-76, Br-82, 

Rb-86, Z5-95 + Nb-95, Mo-99, Tc-98, Rb-105, Ru-103 + Rh-103, Ag-105, Ag-111, 

Cd-108 + Ag-109, Sn-113, Te-127, Te-129, I-132, Cs-137 + Ba-137, La-140, 

Pr-143, Pm-147, Ho-166, Lu-177, Ta-182, W-181, Re-183, Ir-192, Pt-191, Pt-193, 

Au-196, Au-198, Au-199, Tl-200, Tl-202, Tl-204, Pb-203

Class 4 

(slight radiotoxicity)

H-3, Be-7, C-14, F-18, Cr-51, Ge-71, Tl-201

Radiotoxicity here refers to the potential of a radionuclide to cause damage to living tissue by 

absorption of energy from the disintegration of the radioactive material introduced into the body 

(Table G-2). 

Table G-2. Radioisotope laboratory classification.

Relative 

radiotoxicity 

of isotopes

Minimum significant 

quantity (μc)

Type of laboratory or working place required

C. Good chemical

laboratory

B. Radioisotope

laboratory

A. High-level

laboratory

Very high 0.1 10 μCi or less 10 μCi to 10 mCi 10 mCi or more

High 1.0 100 μCi or less 100 μCi to 100 mCi 100 mCi or more

Moderate 10 1 mCi or less 1 mCi to 1 Ci 1 Ci or more

Slight 100 10 mCi or less 10 mCi to 10 Ci 10 Ci or more
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008

Page 134 of 143

ATTACHMENT H 

INSTRUMENTATION ON HAND AND DESIRED AS OF SEPTEMBER 1960 

Page 1 of 3

Table H-1. Instrumentation on hand and desired as of September 1960.a

Instrument Function

On 

hand

Total 

needed

HP 

surveyors

Field 

crews

Emergency 

truck

Experimental 

building Fusion

Science 

buildings LINAC HTGR MGCR TRIGA FLAIR

Hot 

cell

Be 

facility

Waste 

yard

TransportationService 

loaner Janitors

GMbcounter 

with 

headphones

Beta-gamma 

source 

locating

50 37 9 1 1 1 0 26 2 * * 1 * * 1 2 2 1 2

Alpha 

proportional 

counters with 

phones

Alpha 

contamination 

detecting

26 3 7 0 0 2 0 8 1 * 1 0 0 * 1 2 2 1 1

Alphadetecting ion 

chamber 

(Samson)

Alpha 

contamination 

detecting

15 14 5 0 0 2 0 5 0 0 0 0 0 0 0 0 0 1 2

Beta-gamma 

ion chamber 

(Samson)

Beta dose rate 

measuring

24 10 9 0 1 * 0 7 1 * * * * 2 * 1 2 1 0

Gamma ion 

chamber 

(Radetector)

Gamma dose 

rate 

measuring

50 38 9 4 1 * 0 16 1 2 1 2 1 5 1 3 2 1 1

Gamma ion 

chamber 

(Qtpi)

Gamma dose 

rate 

measuring

4 4 0 0 0 0 0 0 2 1 1 0 0 0 0 0 0 0 0

Fast neutron 

survey meter

Fast neutron 

dose rate 

measuring

4 2 2 2 0 0 0 0 * * * * * 0 * 0 0 0 0

Slow neutron 

survey meter

Thermal 

neutron flux 

locating

4 0 2 2 0 0 0 0 * * * * * 0 * 0 0 0 0

Anemometer Measuring air 

velocity

4 2 4 0 0 * 0 * * * * * * * * * 0 0 0

Air samplers 

(Staplex)

Collecting 

airborne 

particulates

9 7 9 0 1 * 0 * * * * * * * * * 0 0 0

Air samplers 

(Giraffe)

Collecting 

airborne 

particulates

14 10 1 0 0 3 0 4 * 1 1 0 0 1 2 1 0 0 0

Continuous 

air monitors 

(Mobile)

Detecting 

airborne beta 

particulates

9 4 0 0 0 0 0 2 1 1 1 1 1 2 0 0 0 0 0

Continuous 

air monitors 

(Station)

Recording

beta activity of 

atmosphere

3 1 2 0 0 0 0 * * * * * * 1 * 0 0 0 0
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008

Page 135 of 143

ATTACHMENT H 

INSTRUMENTATION ON HAND AND DESIRED AS OF SEPTEMBER 1960 

Page 2 of 3

Instrument Function

On 

hand

Total 

needed

HP 

surveyors

Field 

crews

Emergency 

truck

Experimental 

building Fusion

Science 

buildings LINAC HTGR MGCR TRIGA FLAIR

Hot 

cell

Be 

facility

Waste 

yard

TransportationService 

loaner Janitors

GM frisking 

rate meters

Hand, clothing 

and shoe 

check station

8 5 0 0 0 0 0 3 0 0 0 1 1 3 0 0 0 0 0

GM foot 

counters

Foot checking 

station

7 4 0 0 0 0 0 2 0 1 0 0 0 3 0 0 0 0 0

Alpha-beta 

probe friskers

Hand, clothing 

and shoe 

check station

12 12 0 0 0 2 0 4 1 1 2 0 0 1 1 1 0 1 0

Alpha 

proportional 

hand and foot 

counters

Alpha hand 

and foot portal 

check

4 0 0 0 0 2 0 1 0 0 1 0 0 0 0 0 0 0 0

Criticality 

alarms

Warn of 

radiation pulse

22 12 0 0 0 5 3 2 1 1 2 1 1 1 1 2 1 1 0

Criticality 

alarm 

cameras

Locate 

nuclear 

accident 

victim

7 0 0 0 0 3 1 0 0 1 2 0 0 0 0 0 0 0 0

Accident 

dosimeters

Evaluate 

nuclear 

excursions

60 0 0 0 0 12 5 12 3 3 5 1 1 5 2 6 4 0 0

Thin-window 

GM counters

Beta counting 

in laboratory

2 1

Automatic 

sample 

changers

To feed beta 

counters

2 0

Low 

background 

counter

Count alpha 

and beta 

activity of 

sewage, etc.

2 1

Alpha 

scintillation 

counters

Alpha 

counting in 

laboratory

5 5

Analytical 

balance

Weigh 

samples in 

laboratory

1 1

Minometers Read PICs 3 3

PICs Basic 

personnel 

monitoring

1,600 1,400
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008

Page 136 of 143

ATTACHMENT H 

INSTRUMENTATION ON HAND AND DESIRED AS OF SEPTEMBER 1960

Page 3 of 3

Instrument Function

On 

hand

Total 

needed

HP 

surveyors

Field 

crews

Emergency 

truck

Experimental 

building Fusion

Science 

buildings LINAC HTGR MGCR TRIGA FLAIR

Hot 

cell

Be 

facility

Waste 

yard

TransportationService 

loaner Janitors

Radiation 

dosimeters

Permit 

workers to 

measure own 

dose

80 57

Calibration 

facility

Calibrate 

radiation and 

anemometer 

instrument

1 0

Portable 

generator

Emergency 

and 

environmental 

air sampling 

power supply

2 0

Recording 

weather vane 

and 

anemometer

Record wind 

history

1 1

Remote 

recording 

thermometers

Record 

inversion 

history

6 0

Oscilloscope Test and 

maintain 

laboratory

equipment

1 0

a. Source: Ray (1960g). 

b. GM = Geiger-Mϋller. 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 137 of 143

ATTACHMENT I 

SUMMARY OF SOME GA ACCIDENTS AND INCIDENTS 

Page 1 of 2

Table I-1. Summary of some GA accidents and incidents.

Date Location Description Source

January 1, 1960 Hot cell β exposure – 2 individuals, 136-150 mR gamma Cockle 1960a

February 11, 1960 Hot Chem. Lab No. 164 Radioactive spill Cockle 1960b

April 1, 1960 Radioactive waste spill – 7 individuals, 35,000 cpm Ray 1960h

April 14, 1960 TRIGA reactor Release of airborne radioactivity, Ru-88 – 5 individuals 

maximum 

Bold 1960b

September 2, 1960 Science Building B K-85 release – 5 Ci from September 2 to September 3 Ray 1960i

October 31, 1960 Fire in hot furnace room, Clothing contamination – 150-1,200 

cpm alpha

Bold 1960a

August 30, 1961 Counting Room L-236 TRIGA reactor incident – 900 rep Ray 1961

November 2,01961 Hot machine shop Fire – 10 individuals, 8% uranium alloy, 20% enriched Cockle 1961

March 6, 1962 Metallurgical 

Development Building

3 incidents, including uranium fires and flask of fission 

products explosion, from March 6 to March 12, 1962. No 

radiation exposure was associated with the explosion.

Bonner 1963

March 27, 1962 Mark F reactor Release, Iodine-131,-133,-135 – 0.016 mCi released Stern 1962

September 10, 1962 Hot suite Fire – 8% uranium alloy, 20% enrichment Cockle 1962

October 5, 1962 Hot cell Personnel exposure – 1 individual, <2 rem Bedsaul 1962

May 1, 1963 Experimental Building Fire in the hot furnace room, uranium sludge Keith 1963a

May 22, 1963 L-540 Fire – 150 g normal uranium Bold 1963b

May 23, 1963 LINAC Facility Fire – 40 mR Bold 1963c

June 4, 1963 L-441 Spill – 9 individuals, 50,000 cpm alpha Keith 1963b

September 19, 1963 L-403 Radioactive spill – 1 individual, MFPs, 7 mR/hr Lepper 1963

December 4, 1963 Mark-F Reactor Room Radiation release, xenon and krypton Bold 1963d

January 29, 1964 LINAC Accidental irradiation – 2 individuals, 25 mR Kesting & Cockle 1964

March 11, 1964 Service Gallery Urine bioassays – 2 individuals, max 2,200 pCi/24 hr beta in 

urine, clothing contamination

Hughes 1964b

April 23, 1964 L-307 Bioassay for Cs-134 – 1 individual, 20,000 pCi Andrews 1964b

May 17, 1964 Experimental Building Filter fire Bold 1964c

December 30, 1964 Experimental Building Fire in hot machine shop – 1.5 kg of 8.5% U alloy Bold 1965f

January 1, 1965 Hot cell From January 1 to January 15, release of 100 mCi of I-131 

from cutting of a P-9 capsule

Andrews 1965

July 14, 1965 X-ray intensities from X-ray diffraction machine Lepper 1965

May 17, 1966 Station No. 5 Contamination of hot cell operating gallery, Ta-182, Co-58, 

Cr-51

Bold 1966b

May 17, 1966 Console Room Operational error at the 17-MEV accelerator – 9 individuals, 

0-90 mrem

Cockle 1966

June 27, 1966 TRIGA Overexposure incident – cadmium, 1.8 R gamma, 18.7 rad, 

beta

Gurren 1966c

July 16, 1966 L-427 Fire – UZr pellet Bold 1966c

December 6, 1966 B area of LINAC Report stated, “Near fatal radiation accident.” – 1 individual

involved. Recorded dose was 3 mrem. Individuals working 

in LINAC area did not necessarily require radiation safety 

training or escort at this time. 

Bold 1966d

December 28, 1966 Customer Particle Line 

Area

Fire in controlled area – 3 individuals, 150 pCi alpha Hughes 1967

March 6, 1967 LINAC Contamination incident – W-185 Friesenhahn 1967

April 1, 1967 LINAC Incident at the 16-m neutron detection station – 20 mR Bold 1967b

April 13, 1967 Hot suite machine room Fire - UZr fire, 20% enriched Smith 1967

April 27, 1967 Column contractor 

station

Column contractor accident – 26 g uranium Quimby 1967

July 5, 1967 L-441 Uranium fire – 1 individual, 127 grams U-235 Moore 1967a

August 9, 1967 Exposed shoe and home, Ce-144, Cs-134, Cs-137, Zr-95 Maschka 1967

November 1, 1967 Experimental Critical 

Facility

Skin contamination – 6,000 cpm alpha Bedsaul 1967

November 22, 1967 Fuel Fabrication Building Explosion – 60 g uranium Moore 1967b

April 25, 1968 L-236 Hand exposure, Sr-89, Sr-90 – 2.67 rem Jordan 1968

May 14, 1968 LINAC Overexposure of extremities, Finger ring – 43 rem total Cockle 1968b
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 138 of 143

ATTACHMENT I 

SUMMARY OF SOME GA ACCIDENTS AND INCIDENTS 

Page 2 of 2

Date Location Description Source

August 5, 1968 LINAC Radiation incident – radiation dose while machine is off Trimble 1968

September 11, 1968 Experimental Building Uranium fire – 325 g U-235, 8.5% U, 70% enriched Thompson 1968

January 17, 1969 L-218 Ta-182 contamination – 4 individuals Kesting 1969a

January 21, 1969 Clean High Bay in Fuel 

Fabrication Facility

Thorium container explosion – nose wipes showed no 

activity, average air concentration for week of Jan. 20, 1969, 

2.39 × 10-11 μCi/ml (reported as alpha beta)

Hughes 1969a

February 28, 1969 L-407 Contamination incident – Ce-144, Pr-144, 0.1 μCi Kesting 1969b

April 2, 1969 C-D Room adjoining the 

Fast Spectrum Cell in 

LINAC area

High personal radiation exposure – 1 individual, 1.8 rem Young and Jordan 

1969a

June 22, 1969 South Mezzanine Spill of U-235 solution – 80 g U-235 unrecovered Hughes 1969b

June 29, 1969 Fast Spectrum Cell Sub-Critical Time of Flight Spectrum Facility incident – 6 

individuals, I-133, 8 to 140 mrem 

Young and Jordan 

1969b

August 13, 1969 TRIGA Release of Kr-85 – 1.5 mCi Bedsaul 1969

August 29, 1969 TRIGA Fission product release in reactor room – 40 μCi released Trimble 1969
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 139 of 143

ATTACHMENT J 

HTGR AND RERTR IFM INFORMATION 

Table J-1. Radionuclide content of HTGR and RERTR IFM, September 30, 2003. 

Radionuclide

HTGR IFM

Activity (Ci)

RERTR IFM

Activity (Ci)

Total 

HTGR+RERTR

Activity (Ci)

Activity 

fraction

H-3 1.96E-01 1.61E+00 1.81E+00 6.78E-04

Mn-54 0.00E+00 2.14E-05 2.14E-05 8.03E-09

Fe-55 0.00E+00 4.27E+00 4.27E+00 1.60E-03

Co-60 0.00E+00 8.84E-01 8.84E-01 3.32E-04

Ni-59 0.00E+00 3.30E-01 3.30E-01 1.24E-04

Ni-63 0.00E+00 3.75E+01 3.75E+01 1.41E-02

Kr-85 5.57E+00 3.55E+01 4.11E+01 1.54E-02

Sr-90 1.27E+02 6.32E+02 7.59E+02 2.85E-01

Y-90 1.27E+02 6.32E+02 7.59E+02 2.85E-01

Tc-99 0.00E+00 1.40E-01 1.40E-01 5.26E-05

Ru-106 0.00E+00 3.43E-03 3.43E-03 1.29E-06

Sb-125 1.64E-02 5.97E-01 6.13E-01 2.30E-04

Cs-134 2.70E-02 1.71E+00 1.74E+00 6.52E-04

Cs-137 1.31E+02 6.91E+02 8.22E+02 3.09E-01

Pm-147 3.34E-01 1.22E+01 1.25E+01 4.70E-03

Sm-151 1.20E+00 3.16E+00 4.36E+00 1.64E-03

Eu-154 8.11E-01 1.28E+01 1.36E+01 5.11E-03

Eu-155 4.78E-02 2.15E+00 2.20E+00 8.25E-04

Th-232 2.10E-04 0.00E+00 2.10E-04 7.88E-08

U-233 2.92E-01 1.71E-07 2.92E-01 1.10E-04

U-234 3.13E-02 3.91E-04 3.17E-02 1.19E-05

U-235 2.27E-04 7.39E-04 9.66E-04 3.63E-07

U-236 1.04E-03 5.61E-03 6.65E-03 2.50E-06

U-238 3.84E-06 8.58E-04 8.62E-04 3.24E-07

Np-237 0.00E+00 2.48E-03 2.48E-03 9.31E-07

Pu-238 2.74E+00 0.00E+00 2.74E+00 1.03E-03

Pu-239 1.70E-02 1.30E+00 1.32E+00 4.94E-04

Pu-240 1.91E-02 1.35E+00 1.37E+00 5.14E-04

Pu-241 2.16E+00 1.95E+02 1.97E+02 7.40E-02

Pu-242 1.08E-04 3.35E-03 3.46E-03 1.30E-06

Total 3.98E+02 2.27E+03 2.66E+03 1.00E+00

Source: GA (2003). 
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 140 of 143

ATTACHMENT K 

HOT CELL FACILITY REMEDIATION INFORMATION 

Page 1 of 4

Table K-1. Sequence of D&D approach (GA 1998) 

Room Tasks

Controlled Machine Shop and 

Weld Area (108, 108A)

Decontamination of walls, floor, and ceiling. Removal of added floor. Removal of filler 

material between added floor and original floor. Removal of original floor.

MC (112) Removal of equipment inside cell. Decontamination of walls, floor, and ceiling.

Ladies’ and Men’s Restrooms, 

Men’s Change Room (116)

Removal of all equipment from rooms. Decontamination of walls, floor, and ceiling. 

Removal of internal nonsupporting walls. Removal of floor.

Met Cell (112) Complete removal of equipment inside cell. Decontamination of walls, floor, and ceiling.

Dark Room (105A)

Counting Room (106)

Removal of all equipment from room. Decontamination of walls, floor, and ceiling. 

Removal of internal nonsupporting walls. Removal of floor, showers, and drop ceiling. 

Removal of hot drains under floor. Capping of hot drain line going to Room 121.

MC (112) Begin removal of walls.

Counting Room (106)

Warm Metallography Room 

(107)

Complete Room 106. Decontamination of Room 107 walls, floor, and ceiling. Removal 

of floor. Removal of hot drain line from under floor. Capping of hot drain as it leaves 

Room 121 and enters Room 122.

Hot Cells (113, 115) Complete removal of south and west walls of Met Cell. Remove Koll-Morgan [periscope] 

from HLC and begin decontamination.

Warm Metallography Room 

(107)

Manipulator Repair Room 

(122)

Complete work in Room 107, remove abandoned underground duct. Removal of hot sink 

from Room 122. Decontamination of walls, floor, and ceiling. Removal of steel liner from 

walls. Removal of floor sink. Removal of floor. Removal of nonsupporting wall. 

Removal of hot drain lines from under floor. Capping of drain line leading to Room 109.

Hot Cells (113, 115) Complete decontamination of HLC, begin decontamination of LLC.

Tritium Extraction Laboratory

(Room 109)

Corridor, Ladies Change Room

Removal of equipment from rooms. Decontamination of walls, floor, and ceiling. 

Removal of floor. Removal of hot drain lines from under the floor. Capping of drain line 

leading to Room 119. Removal of buried ventilation under Corridor and Dark Room.

LLC (Room 113) Complete decontamination of LLC. Remove Manipulators, PaR, and remaining Hot Cell 

windows.

Tritium Sample Prep. (120)

Tritium Effluent (119), and 

117A and 116A

Decontamination of walls, floor, and ceiling. Removal of nonsupporting walls. Removal 

of slab from Rooms 120 and 119 and abandoned ventilation duct under Room 120.

Hot Cells Removal of steel liner and decontamination of concrete walls.

X-ray laboratory (116) Tool 

Room (117)

Pump Room (114)

Decontamination Room (118)

Service Gallery (111)

116 & 117 – Removal of fume hood. Decontamination of walls, floor, and ceiling. 

Removal of ceiling.

114 – Removal of equipment. Decontamination of walls, floor, and ceiling.

118 – Decontamination of walls, floor, and ceiling. Removal of nonsupporting walls.

111 – Begin decontamination of walls, floor, and ceiling.

Yard Shed- Decontamination of walls, floor, and ceiling. Decontamination of exterior of shed. 

Dismantle shed. Removal of concrete slab.

Service yard – Old stack monitor room; decontamination of walls, floor, and ceiling. 

Decontamination and removal of concrete slabs.

Service Gallery (111)

Boiler Room (121)

Air Ducts

Building Roof

111 – Complete decontamination.

121 – Decontamination floor, walls, and ceiling. Remove floor. Remove hot drain lines.

Air Ducts – Remove air ducts from Facility.

Roof – Begin removal of roof.

Yard Remove hot drain line. Begin decontamination of liquid waste storage area.

Boiler Room (121)

Building Roof

Building Walls

121 – Complete demolition of room

Roof – Finish roof removal

Walls – Begin removal of building walls (except for old offices).

Hot Cells Begin removal of cell roof.

Yard

Outside HEPA System

Complete removal of hot drain lines and clearing of waste storage area.

Begin removal of HEPA system.

Hot Cells Complete removal of Hot Cell roof. Remove Hot Cell doors. Begin removal of cell walls.

Walls Complete removal of building walls.

HEPA System Finish removing HEPA System and surrounding concrete slabs. Decontamination and 

removal of Hot Cell HEPA pit.
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 141 of 143

ATTACHMENT K 

HOT CELL FACILITY REMEDIATION INFORMATION 

Page 2 of 4

Room Tasks

Building 23 Begin removal of remainder of slab.

Hot Cells Finish removing Hot Cell walls. Remove Hot Cell floor and hot drain beneath Hot Cells.

Hot Drain

Service yard

Remove exterior hot drains.

Begin soil remediation.

Building 23 Complete removal of building slab.

Hot Cells Begin removal of Hot Cell foundation, pits, and HEPA ducts.

Service Yard Continue with soil remediation.

Building Remove pit from under Room 111.

Hot Cells Continue removal of HEPA ducts. Begin removal of Hot Cell walls.

100, 102, 103, 104 Demolish remaining rooms.

Hot Cells Complete removal of HEPA ducts. Complete removal of Hot Cell wells.

Yard Begin soil remediation.

Yard Complete soil remediation.

Site Postdecontamination surveys, additional remediation.

Table K-2. Concentrations of radionuclides found in soil, asphalt, and 

concrete before remediation of the HCF (GA 1998) 

Sample location Depth interval [cm (in.)]

Cs-137 

(pCi/g)

Cs-134 

(pCi/g)

Co-60 

(pCi/g)

A-4

a

0.0–5.0

5.0–30.0

60.0–87.5

(0.0–2.0)

(2.0–12.0)

(24.0–35.0)

1.8

<LLD

<LLD

<LLDb

<LLD

<LLD

<LLD

<LLD

<LLD

A-5

c

0.0–7.5

7.5–27.5

(0.0–3.0)

(3.0–11.0)

<LLD

4.6

<LLD

<LLD

<LLD

0.56

A-6

a

0.0–7.5

7.5–32.5

32.5–40.0

(0.0–3.0)

(3.0–13.0)

(13.0–16.0)

1.52

0.62

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

A-7

a

0.0–6.2

6.2–23.8

(0.0–2.5)

(2.5–9.5)

0.34

0.37

<LLD

<LLD

<LLD

0.25

B-1

a

0.0–15.0

15.0–42.5

(0.0–6.0)

(6.0–17.0)

<LLD

<LLD

<LLD

<LLD

<LLD

0.29

B-2

c

1.2–23.8

23.8–35.0

(0.5–9.5)

(9.5–14.0)

0.68

0.22

<LLD

<LLD

0.53

<LLD

B-3 0.0–7.5

7.5–23.8

(0.0–3.0)

(3.0–9.5)

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

B-4

c

0.0–15.0

15.0–30.0

(0.0–6.0)

(6.0–12.0)

0.22

<LLD

<LLD

<LLD

1.26

<LLD

B-6 0.0–5.0

5.0–22.5

(0.0–2.0)

(2.0–9.0)

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

B-7

c

0.0–3.8

5.0–25.0

(0.0–1.5)

(2.0–10.0)

2.97

0.77

<LLD

<LLD

1.44

0.57

B-8

d

0.0–12.5

12.5–30.0

(0.0–5.0)

(5.0–12.0)

38.3

0.43

0.24

<LLD

21

<LLD

B-9 0.0–12.5

12.5–30.0

(0.0–5.0)

(5.0–12.0)

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

C-1

c

2.5–13.8

13.8–30.0

(1.0–5.5)

(5.5–12.0)

0.81

0.07

<LLD

<LLD

1.75

<LLD

C-5

c

0.0–3.8

3.8–17.5

(0.0–1.5)

(1.5–7.0)

2.92

3.47

<LLD

<LLD

0.92

4.03

C-6 0.0–13.8

13.8–28.2

(0.0–5.5)

(5.5–11.5)

0.62

0.38

<LLD

<LLD

<LLD

<LLD

C-7 N/A–Not sampled

C-8

a

0.0–15.0

21.2–42.5

(0.0–6.0)

(8.5–17.0)

0.46

<LLD

<LLD

<LLD

<LLD

<LLD
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 142 of 143

ATTACHMENT K 

HOT CELL FACILITY REMEDIATION INFORMATION 

Page 3 of 4

Sample location Depth interval [cm (in)]

Cs-137 

(pCi/g)

Cs-134 

(pCi/g)

Co-60 

(pCi/g)

C-9 0.0–12.5

15.0–40.0

(0.0–5.0)

(6.0–16.0)

0.28

<LLD

<LLD

<LLD

<LLD

<LLD

C-10 0.0–20.0

20.0–40.0

(0.0–8.0)

(8.0–16.0)

0.17

<LLD

<LLD

<LLD

<LLD

<LLD

D-1

a

1.2–18.8

18.8–45.0

(0.5–7.5)

(7.5–18.0)

0.67

0.26

<LLD

<LLD

<LLD

<LLD

D-5 N/A–Not sampled

D-6

c

0.0–7.5

7.5–20.0

(0.0–3.0)

(3.0–8.0)

8.4

2.22

<LLD

<LLD

<LLD

<LLD

D-7

d

0.0–13.8

15.0–35.0

(0.0–5.5)

(6.0–14.0)

4.53

0.12

<LLD

<LLD

0.44

<LLD

D-8

a

0.0–5.0

5.0–21.2

(0.0–2.0)

(2.0–8.5)

2.71

0.11

<LLD

<LLD

0.52

0.27

D-9 0.0–5.0

5.0–17.5

(0.0–2.0)

(2.0–7.0)

7.11

0.23

0.65

<LLD

<LLD

<LLD

D-10 0.0–5.0

10.0–30.0

(0.0–2.0)

(4.0–12.0)

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

E-1

c

0.0–16.2

16.2–32.5

(0.0–6.5)

(6.5–13.0)

<LLD

1.89

<LLD

<LLD

<LLD

1.47

E-2

a

0.0–13.8

13.8–30.0

52.5–65.0

65.0–72.5

75.0–82.5

90.0–100.0

(0.0–5.5)

(5.5–12.0)

(21.0–26.0)

(26.0–29.0)

(30.0–33.0)

(36.0–40.0)

<LLD

0.79

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

0.22

<LLD

<LLD

<LLD

<LLD

E-3 0.0–13.8

13.8–25.0

(0.0–5.5)

(5.5–10.0)

<LLD

0.2

<LLD

<LLD

<LLD

<LLD

E-4 0.0–11.2

13.8–35.0

(0.0–4.5)

(5.5–14.0)

<LLD

<LLD

<LLD

<LLD

0.76

<LLD

E-5 0.0–7.5

11.2–37.5

(0.0–3.0)

(4.5–15.0)

2.28

<LLD

0.28

<LLD

<LLD

<LLD

E-6 0.0–15.0

0.0–30.0

30.0–55.0

55.0–87.5

(0.0–6.0)

(0.0–12.0)

(12.0–22.0)

(22.0–35.0)

0.54

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

E-7 0.0–7.5

7.5–27.5

(0.0–3.0)

(3.0–11.0)

1.58

0.21

<LLD

<LLD

<LLD

<LLD

E-8 N/A–Not sampled

E-9 0.0–5.0

22.5–37.5

(0.0–2.0)

(9.0–15.0)

0.19

<LLD

<LLD

<LLD

<LLD

<LLD

E-10a0.0–5.0

10.0–30.0

(0.0–2.0)

(4.0–12.0)

0.64

<LLD

<LLD

<LLD

<LLD

<LLD

F-2 0.0–11.2

11.2–30.0

(0.0–4.5)

(4.5–12.0)

0.65

<LLD

<LLD

<LLD

<LLD

<LLD

F-3 0.0–8.8

8.8–37.5

(0.0–3.5)

(3.5–15.0)

<LLD

<LLD

<LLD

<LLD

<LLD

<LLD

F-4

a

0.0–5.0

5.0–15.0

(0.0–2.0)

(2.0–6.0)

0.6

<LLD

<LLD

<LLD

0.19

<LLD

F-5

c

0.0–5.0

15.0–22.5

(0.0–2.0)

(6.0–9.0)

2.1

<LLD

<LLD

<LLD

1.5

<LLD

F-6

c

0.0–5.0

10.0–22.5

(0.0–2.0)

(4.0–9.0)

2.9

<LLD

<LLD

<LLD

2.4

<LLD

F-7

c

0.0–5.0

5.0–17.5

(0.0–2.0)

(2.0–7.0)

1.1

0.6

<LLD

<LLD

0.36

0.12
Document No. ORAUT-TKBS-0045 Revision No. 00 Effective Date: 09/26/2008 Page 143 of 143

ATTACHMENT K 

HOT CELL FACILITY REMEDIATION INFORMATION 

Page 4 of 4

Sample location Depth interval [cm (in)]

Cs-137 

(pCi/g)

Cs-134 

(pCi/g)

Co-60 

(pCi/g)

F-8

a

0.0–5.0

7.5–27.5

(0.0–2.0)

(3.0–11.0)

0.7

0.08

<LLD

<LLD

0.25

<LLD

F-9

a

0.0–5.0

22.5–30.0

(0.0–2.0)

(9.0–12.0)

0.45

<LLD

<LLD

<LLD

<LLD

<LLD

F-10c0.0–15.0

15.0–30.0

30.0–45.0

75.0–90.0

(0.0–6.0)

(6.0–12.0)

(12.0–18.0)

(30.0–36.0)

2.1

0.07

0.47

<LLD

<LLD

<LLD

<LLD

<LLD

0.58

<LLD

<LLD

<LLD

Judgment locations

General trestle area

R-11Ad

0.0–10.0

10.0–22.5

45.0–60.0

65.0–90.0

90.0–105.0

(0.0–4.0)

(4.0–9.0)

(18.0–24.0)

(26.0–36.0)

(36.0–42.0)

217

28.4

0.69

<LLD

<LLD

158

6.71

0.26

<LLD

<LLD

9.38

4.62

<LLD

<LLD

<LLD

General trestle area

R-11Bd

0.0–6.2

6.2–18.8

41.2–50.0

67.5–80.0

(0.0–2.5)

(2.5–7.5)

(16.5–20.0)

(27.0–32.0)

41.2

25.8

0.84

0.79

<LLD

0.66

<LLD

<LLD

5.38

9.9

0.19

0.2

General trestle area

R11Cc

0.0–5.0

6.2–22.5

60.0–67.5

75.0–82.5

(0.0–2.0)

(2.5–9.0)

(24.0–27.0)

(30.0–33.0)

163

5.88

0.23

0.18

6.56

0.39

<LLD

<LLD

23.2

1.18

<LLD

<LLD

Liquid waste tanks

Area R13d

0.0–17.5

35.0–50.0

55.0–97.5

(0.0–7.0)

(14.0–20.0)

(22.0–39.0)

84

7.81

0.58

2.49

0.88

<LLD

18.4

2.58

0.16

Liquid waste tanks

Area R16d

0.0–20.0

45.0–57.5

65.0–100.0

(0.0–8.0)

(18.0–23.0)

(26.0–40.0)

126.8

0.78

0.7

<LLD

<LLD

<LLD

107.3

1.16

0.92

Liquid waste tanks

Area R35d

0.0–17.5

35.0–60.0

62.5–100.0

(0.0–7.0)

(14.0–24.0)

(25.0–40.0)

110.4

2.36

0.31

1.32

<LLD

<LLD

10.9

0.6

<LLD

Drainage path from

liquid waste tanks

Area S06Ad

0.0–7.5

7.5–22.5

35.0–60.0

82.5–100.0

(0.0–3.0)

(3.0–9.0)

(14.0–24.0)

(33.0–40.0)

17.1

107.2

4.07

1.47

0.15

<LLD

<LLD

<LLD

4.3

14

0.92

0.45

Drainage path from

liquid waste tanks

Area S06Bd

0.0–18.8

30.0–50.0

50.0–90.0

(0.0–7.5)

(12.0–20.0)

(20.0–36.0)

73.5

5.6

1.59

<LLD

<LLD

<LLD

13.3

1.53

0.58

Drainage path from

liquid waste tanks 

Area S06Cd

0.0–25.0

25.0–50.0

57.5–97.5

(0.0–10.0)

(10.0–20.0)

(23.0–39.0)

34.5

5.59

0.6

6.29

0.69

<LLD

4.77

1.6

<LLD

a. The location soil was positive for licensed material in excess of background values. 

b. <LLD means that measured concentration was less than the lower limit of detection for the 

radionuclide. 

c. The location soil was positive for licensed material and exceeded 10% of proposed release 

criteria. 

d. The location soil was positive for licensed material and exceeded proposed release criteria.