ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 64

Document Title:

Energy Technology Engineering Center – 

Occupational Internal Dose 

Document Number: ORAUT-TKBS-0038-5 

Revision: 00

Effective Date: 02/22/2006

Type of Document: TBD

Supersedes: None

Subject Expert: Eugene W. Potter

Document Owner

Approval: Signature on File Approval Date: 02/08/2006

Melton H. Chew, TBD Team Leader

Approval: Signature on File Approval Date: 02/06/2006

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 02/06/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 02/22/2006

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 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 2 of 64

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

02/22/2006 00 Technical basis document for Energy Technology Engineering 

Center – Occupational Internal Dose. First approved issue. Training 

required: As determined by the Task Manager. Initiated by Melton 

H. Chew.
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TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

5.1 Introduction ..............................................................................................................................8

 5.1.1 Purpose..........................................................................................................................8

5.1.2 Scope.............................................................................................................................9

5.2 Internal Dosimetry Overview ....................................................................................................9

5.3 In Vitro Minimum Detectable Activities, Counting Methods, and Reporting Protocols ............. 13

5.3.1 In Vitro Urine Analysis ................................................................................................. 13

 5.3.1.1 1948 – 1957.................................................................................................. 13

5.3.1.2 1958 – 1966.................................................................................................. 13

5.3.1.3 1967 – 1974.................................................................................................. 15

5.3.1.4 1975 – 1988.................................................................................................. 17

5.3.1.5 1989 – 2005.................................................................................................. 19

5.3.2 In Vitro Methods for Individual Radionuclides .............................................................19

5.3.3 Fecal Sample Analysis ...............................................................................................19

5.4 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting Practices.............. 20

5.4.1 Whole-Body Counting ................................................................................................. 20

5.4.2 Chest Counting ...........................................................................................................20

5.5 Uncertainty.............................................................................................................................22

5.6 Detection Limits .....................................................................................................................22

5.7 Unit Codes and Description of Units....................................................................................... 23

5.8 Excreta Sample Kit Codes ..................................................................................................... 23

5.9 Solubility Type, Fraction Activity, and Particle Size by Facility................................................ 25

5.10 Facility-Specific Radionuclide Conversions ............................................................................ 26

5.11 Workplace Monitoring Data.................................................................................................... 26

5.12 Source Term Data..................................................................................................................30

5.13 Radon ....................................................................................................................................31

References ........................................................................................................................................32

Glossary ............................................................................................................................................37

ATTACHMENT A, SUMMARY OF URANIUM FUEL FABRICATION INTERNAL 

DOSIMETRY ISSUES..........................................................................................39

ATTACHMENT B, EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS..............................41
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ATTACHMENT C, ADJUSTMENT TO HELGESON NUCLEAR SERVICES 235U LUNG 

COUNTS, 1981-1983...........................................................................................61

ATTACHMENT D, ISOTOPIC COMPOSITION OF ATR-ETR FUEL USED IN THE 

POWDER ROOM, 1966-1967..............................................................................62

ATTACHMENT E, DETECTION LIMITS FOR GAMMA SCANS OF BIOASSAY 

SAMPLES, 1967-1975 .........................................................................................64

LIST OF TABLES

TABLE TITLE

5-1 Internal dose control program (in vitro) ...................................................................................

PAGE

20

5-2 Internal dose control program (in vivo) ................................................................................... 22

5-3 Detection limits 1958–1966.................................................................................................... 23

5-4 Detection limits 1967–1974.................................................................................................... 24

5-5 Detection limits 1975–1988.................................................................................................... 25

5-6 Detection limits 1989–2005.................................................................................................... 25

5-7 Unit codes and description of units ......................................................................................... 26

5-8 Excreta sample kit codes ....................................................................................................... 26

5-9 Solubility type, fraction activity, and particle size by facility..................................................... 27

5-10 Facility-specific radionuclide conversions............................................................................... 31

LIST OF FIGURES

FIGURE TITLE

B-1 Example of Request for Film Badge and Bioassay Services, 1963 ........................................

PAGE

41

B-2 Example Individual Personnel Keysort Card, 1966................................................................. 42

B-3 Example Bioassay Data Sheet, page 1 of 2 ........................................................................... 44

B-4 Example Bioassay Data Sheet, page 2 of 2 ........................................................................... 46

B-5 Example Bioassay Data Sheet, BioScience, Uranium Fluorometric ....................................... 48

B-6 Example Bioassay Data Sheet, BioScience, Uranium Radiometric ........................................ 50

B-7 Example Bioassay Data Sheet, BioScience, Uranium Radiometric, Showing Bottle Tags...... 52

B-8 Example Bioassay Data Sheet, UST, Uranium Fluorometric .................................................. 54

B-9 Example Bioassay Data Sheet, UST, Uranium Radiometric................................................... 56

B-10 Example Bioassay Data Sheet, UST, Mixed Fission Products ............................................... 58

B-11 Example Wound Monitoring Report........................................................................................ 60
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AERD Atomic Energy Research Department

AETR Advanced Epithermal Thorium Reactor

AI Atomics International

AMAD activity median aerodynamic diameter

ATR Advanced Test Reactor

Bq becquerel

CF critical or criticality facility

cm centimeter

CMD count median diameter

CTF critical or criticality test facility

D&D Decontamination and Decommissioning

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ETEC Energy Technology Engineering Center

ETR Engineering Test Reactor

EU enriched uranium

F fast (solubility rate)

FSCF (SNAP) flight system critical facility

g gram

GIF Gamma Irradiation Facility

HEU highly enriched uranium

HMRFSR Heavy Metal Reflected Fast Spectrum Reactor

hr hour

in. inch

IREP Interactive RadioEpidemiological Program

IVLC in-vivo lung count

JAERI Japan Atomic Energy Research Institute

keV kiloelectron-volt

KEWB Kinetics Experiment Water Boiler Reactor

kg kilogram

L liter

L-85 Alternate name for the WBNS after 1972

LLD lower limit of detection – the lowest value where radioactivity was considered to be 

present with reasonable certainty
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M moderate (solubility rate)

MDA minimum detectable amount

MFP mixed fission product

mg milligram

min minute

mL milliliter

MPBB Maximum Permissible Body Burden

MPC Maximum Permissible Concentration

MPLB Maximum Permissible Lung Burden

NAA North American Aviation, Inc.

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NMDF Nuclear Materials Development Facility

NRTS National Reactor Testing Station

NSEC Nuclear Science and Engineering Corporation

OCY Old Conservation Yard

OMR organic moderated reactor

OMRE Organic Moderated Reactor Experiment

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

oz ounce

pCi picocurie

PUA plutonium activity determined by autoradiography

PUB plutonium activity determined using a proportional counter to count alpha particles

R&D Research and Development

RMHF Radiation Materials Handling Facility

RSRMS Radiation Safety Records Management System

S slow (solubility rate)

S10FS SNAP 10 Flight Simulation Reactor

S2DR SNAP 2 Development Reactor

S8DR SNAP 8 Development Reactor

S8ER SNAP 8 Experimental Reactor

SER SNAP Experimental Reactor

SGR Sodium Graphite Reactor

SNAP Systems for Nuclear Auxiliary Power

SRE Sodium Reactor Experiment

SSFL Santa Susana Field Laboratory

STIR Shield Test and Irradiation Reactor (Modified STR)

STR Shield Test Reactor

TBC total body count

TBD technical basis document

TLD thermoluminescent dosimeter

TTA thenoyltrifluoroacetone

UAlx uranium aluminide
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UCLA University of California, Los Angeles

UF Uranium Fluorometric

UR Uranium Radiometric

U.S.C. United States Code

UST U. S. Testing

W watt

WBC whole-body count

WBNS Water Boiler Neutron Source

ZnS(Ag) zinc sulfide scintillation crystal activated with silver

μg microgram

μL microliter
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5.1 INTRODUCTION

Technical basis documents (TBDs) and site profile documents are general working documents that 

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

of sites. They will be revised in the event additional relevant information is obtained about the 

affected site(s). These documents may be used to assist the National Institute for Occupational 

Safety and Health (NIOSH) in the completion of the individual work required for each dose 

reconstruction.

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

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

employer facility” or a “Department of Energy [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 [probability of causation] 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) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

As noted above, the statute 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 

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

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, 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 radiation exposures 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 dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

5.1.1

The purpose of this TBD is to describe internal dosimetry systems and practices at the Energy 

Technology Engineering Center (ETEC). This document discusses historical and current practices in 

relation to the evaluation of internal exposure data for monitored and unmonitored workers.

Purpose
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5.1.2

ETEC, operated by The Boeing Company and its predecessors, has played an important role in the 

development of the U.S. nuclear program. ETEC operations have involved research and 

development (R&D) in the areas of development and testing of nuclear reactors, nuclear support 

operations, and non-nuclear energy R&D (Sapere and Boeing 2005, p. 2-2). This TBD is part of the 

overall ETEC Site Profile, which describes plant facilities and processes, historical information, and 

environmental data in relation to dose reconstruction for ETEC workers. It contains supporting 

documentation to assist in the evaluation of occupational internal doses from these processes in 

accordance with the Internal Dose Reconstruction Implementation Guideline (NIOSH 2002).

Scope

The methods and concepts of measuring occupational internal doses to workers have evolved since 

the beginning of ETEC operations. An objective of this document is to provide supporting technical 

data to evaluate internal ETEC occupational doses that can reasonably be associated with worker 

radiation exposures covered by the EEOICPA legislation.

In addition, this document presents the technical basis of methods used to prepare ETEC worker 

dose information for input to the NIOSH IREP (Interactive RadioEpidemiological Program) computer 

code. Information on measurement uncertainties is an integral component of the approach. This 

document describes the evaluation of uncertainty for ETEC exposure and dose records.

5.2 INTERNAL DOSIMETRY OVERVIEW

Section 2 of this Site Profile (ORAU 2005) contains some of the following discussion of historical 

activities. However, this TBD discusses the information in the context of internal dosimetry.

From 1948 to 1955, the Atomic Energy Research Department (AERD) of North American Aviation, 

Inc. (NAA), occupied a portion of Building 001 at the Downey Plant on Lakewood Boulevard in the 

City of Downey, California. On April 21, 1952, AERD constructed a small aqueous homogeneous 

reactor in an area that is now a loading dock. The reactor and its associated “exponential pile” were 

apparently used as a source of neutrons for reactor physics experiments. The Water Boiler Neutron 

Source (WBNS) used a 93%-enriched uranyl sulfate solution and operated at power levels up to 4 W. 

The WBNS operated at Downey until December 1955. In 1956, it was dismantled and moved to a 

facility at the Santa Susana Field Laboratory (SSFL). Little information is available about the specific 

uses of the area after the WBNS relocation. All available records indicate that the Plant was not left in 

a contaminated condition, and this was confirmed by a survey in 2000 (Liddy and Rutherford 2001). 

Other radioisotopes at the Downey Plant were apparently used for industrial radiography and were not 

a likely source of internal exposure.

In 1956, NAA formed Atomics International (AI) as one of its divisions, replacing AERD. Between 

1956 and 1960, AI performed nuclear R&D at a facility known as the Vanowen Building (Building 038) 

on Canoga Avenue in Canoga Park, California. AI designed, developed, and operated two small, 

aqueous, 93%-enriched uranyl sulfate research reactors, designated L-47 and L-77, at the Vanowen 

facility. The maximum power ratings were 5 W and 10 W, respectively. Other operations included 

reactor design, fuel development, and radiochemistry. L-47 operated between August 1957 and June 

1958, and L-77 operated between May 1958 and February 1960. The reactors were in the southeast 

corner of the building at what is now a loading dock. Historical information indicates that the areas 

that supported nuclear operations were routinely surveyed up to and following the removal of all 

radioactive material. No contamination above the current limits for unrestricted use was found. This 

was confirmed by a 1995 survey of the facility by the U.S. Nuclear Regulatory Commission
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(Rutherford 2002). Exposure of subsequent occupants to contamination levels of importance for 

internal dose reconstruction of the building would be unlikely.

In 1959, AI moved to a new facility on De Soto Avenue in Canoga Park. All radiological activities from 

the Vanowen Building, including the L-77 reactor, transferred to the new facility. Until 1983, nuclear 

operations were conducted at the De Soto facility in Buildings 101 and 104 (until 1984 these buildings 

were known as 001 and 004, respectively). Work continued at a “much-reduced level” in Building 104 

until the mid-1990s. The L-77 reactor, housed in Room 416-61 of Building 104, was decommissioned 

and decontaminated in the late 1970s. The facility was released for unrestricted use in February 

1982. Many nuclear fuel manufacturing operations were conducted in the 1960s and 1970s in 

Building 101 and Building 104. These operations used 2%- to 93%-enriched uranium metal and 

composites. Most notable of these was fuel for the Advanced Test Reactor (ATR), which utilized 

uranium-aluminum alloy plates. This operation was the source of internal exposures that were difficult 

to evaluate because of the unanticipated insolubility and small particle size (when airborne) of the 

alloy. Attachment A summarizes these problems.

Two other operations in Building 104 at the De Soto facility utilized radioactive material. A Gamma 

Irradiation Facility (GIF) used sealed 137Cs and 60Co sources for hardness testing and food irradiation 

research. Biannual leak tests of these sources confirmed that internal exposures from these devices 

were unlikely. Operations ended in the late 1980s and, following the removal of all equipment, the 

State of California released the GIF for unrestricted use in July 1999. The other Building 104 

operation, the Mass Spectroscopy Laboratory or “Helium Laboratory,” analyzed radioactive samples 

for helium content. Operations ended in 1995; the mass spectroscopy equipment was sent to 

Richland, Washington. In 1998, all remaining support equipment was removed and the facility was 

decontaminated. In October 1998, the State released the Helium Laboratory for unrestricted use. 

Thirteen separate radiation surveys in Buildings 101 and 104 demonstrated that no residual 

contamination exists that would be of interest for dose reconstruction (Boeing 2003).

SSFL consisted of four administrative areas and a buffer zone. NAA established Area IV in 1953 as a 

nuclear R&D facility. Starting in 1954, several nuclear reactors and critical assemblies were built and 

operated in Area IV. In 1959, a significant incident occurred when a sodium-cooled, graphitemoderated reactor in Building 4143 (the SRE or Sodium Reactor Experiment) had a loss of coolant, 

which resulted in damage to 13 fuel assemblies (Sapere and Boeing 2005, pp. 2-1, 2-3, 2-5). A 

government-owned, contractor-operated organization was formed to conduct non-nuclear research at 

the site, which was renamed ETEC in 1978. Most nuclear research programs and operations ended

in 1988. All research ended in 1998. ETEC was given the job of decontaminating and 

decommissioning the former nuclear facilities. Many of the facilities in Area IV have been 

decontaminated and decommissioned, but some of this work is still ongoing.

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

including a laboratory with bioassay capability. The program was under central direction and covered 

the facilities operated by AI (Lang undated, 1960). There is one possible exception. The Organic 

Moderated Reactor Experiment (OMRE) was an AI facility at the National Reactor Testing Station 

(NRTS) in Arco, Idaho. NRTS apparently provided dosimetry services to AI personnel (Lang 1960). It 

is uncertain, based on current information, if bioassay results, if any, were sent to AI.

AI established health and safety files on each employee that contained radiation exposure records, 

injury records, and other “pertinent” data (Lang undated, 1960). Today, personnel radiation exposure 

records are in the Radiation Safety Records Management System (RSRMS), which encompasses

about 170 file cabinets. In addition, the RSRMS includes records of routine surveys and 

environmental monitoring data (Sapere and Boeing 2005, p. 3-12).
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The environmental monitoring program at Area IV was established in May 1954 prior to construction 

of the first radiological facility. Stack air emissions were measured in all facilities with radiological 

work areas or where unencapsulated or unpackaged radioactive material was handled. Gross alpha 

and beta activity was monitored on a weekly basis (Sapere and Boeing 2005, p. 3-13). From 1959 to 

the present, ambient gross beta activity in air has been measured continuously in five locations. From 

1963 on, gross alpha activity was also measured. At present, ambient air samples are analyzed for 

isotopic content (Sapere and Boeing 2005, p. 3-15).

For simplicity, the remainder of this TBD uses the term ETEC to refer to all the organizations that 

operated at De Soto, Canoga Park, Downey, and SSFL, although technically ETEC did not exist for 

much of this time.

By 1959, routine urine samples were requested on Fridays, and each employee was required to 

submit the first voiding on Monday morning (following an absence from work of 48 hr or more). The 

time of the previous voiding was recorded to determine the excretion rate. If the Monday morning 

sample was verified as positive, a series of 24-hr samples was collected to determine the body 

burden. Employees were requested to fill these samples on Sunday (Kellehar 1959). Appropriate 

adjustments to this schedule were made for weekend work, etc. Uranium urinalyses were performed

for fuel fabrication workers and gross beta or mixed fission product urinalyses were performed for 

many workers at SSFL and Vanowen. Commercial laboratories were available for analysis of alpha 

emitters other than uranium and for soft beta emitters (Alexander 1959). By 1966, most routine 

samples were analyzed by a vendor. Special samples were requested when an intake was suspected 

or if a routine sample exceeded 10% of the urinary excretion expected from one Maximum 

Permissible Body Burden (MPBB). Special samples were analyzed by the onsite laboratory, the 

vendor, or both (Kellehar 1966a).

Entry into the bioassay program at ETEC was apparently based on job assignment. By the early 

1960s, the bioassay program “normally” consisted of urinalysis for personnel whose work 

assignments involved “potential exposure to radioactive materials.” The frequency of sampling varied 

from one to four per year, depending on the nature of the employee’s work, past exposure history, etc. 

(Lang undated). In 1963, a procedure called for completion of a “Request for Film Badge and 

Bioassay Services” form for new employees who required such services (Garcia 1963). Figure B-1 in

Attachment B is an example. The Health and Safety Operation Unit completed the bottom portion, 

which included a place to specify bioassay for the employee.

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

unencapsulated radioactive material was present required baseline and termination urine samples. A 

new baseline could be required for a change in job assignment. For new operations, a “pilot” 

bioassay program consisting of weekly urine samples could be required until a pattern was 

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

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

counts could also be required. Employees who performed work in these areas only periodically were 

subject to semiannual urine samples. Personnel such as project engineers, industrial engineers, etc., 

who frequently entered these areas but did not perform hands-on work provided annual routine urine 

samples.

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

changed to “personnel whose work assignments potentially expose them to respirable-sized 

radioactive aerosols” (Hart 1979). By the late 1980s, the criterion was “personnel whose work 

assignments potentially expose them to radioactive aerosols” (Tuttle 1989). Quarterly urine sampling 
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was the norm through the 1980s (Hart 1979; 1980a,b,c; Eggleston 1983, 1984; Tuttle 1985, 

1986a,b,c, 1988a,b, 1989).

Special bioassay consisting of more frequent urine sampling was in place very early (1960), and in the 

mid-1970s fecal sampling was also used but “only when gross internal contamination” was suspected 

(Hart 1979). Using the concept of an MPBB, an excretion rate was determined by radionuclide that 

would indicate that one MPBB had been received. For several years prior to 1968, the policy was to 

restrict employees from work in potential airborne areas until their body burden was less than 25% of 

the MPBB (Alexander 1968a). Starting in January 1968, ETEC imposed a restriction from work in 

areas with potential airborne exposure (or in some cases from all radiation areas) if the bioassay 

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

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

MPBB (Staszesky 1970).

It appears that the first in vivo counting occurred in 1967 at a U.S. Atomic Energy Commission (AEC)-

funded facility at the University of California, Los Angeles (UCLA). These counts were performed in 

conjunction with the insoluble enriched uranium (EU) intakes described in Attachment A. Although 

these counts were frequently referred to as whole-body counts (WBCs) or total body counts (TBCs) in 

site documents, they were really attempts to quantify the amount of 235U activity in the lung. At that 

time, chest (lung) counting was in its infancy and standard calibration methods using phantoms had 

not been developed. ETEC ultimately determined that the results of these counts were significant 

overestimates. From 1968 to 1983, in vivo counting was done on the site using portable counters 

operated by Helgeson Nuclear Services, Inc. Most of these counts were also chest counts for EU. 

Work with unirradiated highly enriched uranium (HEU) ended in 1983, and no more chest counts 

occurred (Tuttle 1986b). WBCs for fission and activation products apparently occurred sporadically 

during this period, probably in conjunction with suspected exposures or new projects.

Even though ETEC ended research activities with radioactive materials in 1988, the internal dosimetry 

program continued; most of the features of the program are listed above. The ETEC bioassay 

program is apparently not accredited by the DOE Laboratory Accreditation Program.

The bioassay records in the individual files generally consist of:

• Individual Personnel Keysort Cards (Figure B-2, Attachment B), which were used to track the 

type, frequency, and week of sample collection. In addition, the form summarized individual 

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

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

be the only place in vivo data are listed. Two versions were produced – 1963–1966 and 

1967–1970. The latter continued in use at least until 1983. This form was the successor to 

one that recorded only results prior to 1963 (example not included).

• Bioassay Data Sheets (Figures B-3 and B-4, Attachment B), which were two-page forms that 

contained all the information on analyses conducted by the onsite laboratory. Some, if not 

most, offsite analysis results were apparently transferred to this form.

• Individual Analysis Results Sheets (Figure B-5 to B-10, Attachment B), which contained all the 

information on analyses conducted by offsite laboratories. This form might show the 

calculated result (even if it was below the limit of detection), the sensitivity of the analysis, and 

the uncertainty.
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• Wound Monitoring Report (Figure B-11, Attachment B). The entries on this form are selfexplanatory.

• Individual hand-drawn or computer-generated plots of bioassay data, apparently done as 

follow-up to high results. Some can be difficult to read due to the quality of the copies.

• Incident workups following high bioassay results. These might contain data summaries in a 

more legible and understandable form than information in other locations because they were 

produced by health physicists familiar with the data.

5.3 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PROTOCOLS

5.3.1

5.3.1.1 1948 – 1957

In Vitro Urine Analysis

Apparently, no bioassay program existed before August 1958. Useful source-term data or facilityspecific ambient airborne concentrations are probably not available. This unmonitored period 

includes the operation of the WBNS from 1948 to 1955 at the Downey Plant, the operation of the L-47 

reactor at the Vanowen facility, and operations at SSFL from 1954 to 1957, including for example the 

operation of the WBNS in 1956 and 1957. Documents indicating environmental releases or incidents 

from these facilities have not been located. Bioassay data after 1957 in similar facilities, such as the 

WBNS or L-77 Reactor, might be the only data available to the dose reconstructor as an indicator of 

missed dose. Because this data would be from a different period, use it with caution.

5.3.1.2 1958 – 1966

According to Kellehar (1966) the bioassay program was initiated in August 1958. As mentioned

above, routine bioassay was normally by urinalysis at a frequency of 3 months. At first, gross alpha or 

gross beta measurements were made of the samples. Specific radionuclides could be determined 

“where required” (Lang 1960). Some detail has been found on early urinalysis methods. In addition 

to the in-house laboratory capability, bioassay services were contracted to the following vendors:

• Tracerlab, Richmond, California

• Nuclear Science and Engineering Corporation (NSEC), Pittsburgh, Pennsylvania

• U. S. Nuclear, Burbank, California

• Controls for Radiation, Cambridge, Massachusetts

• Biomedical Procedures, North Hollywood, California

• BioScience Laboratories, Santa Monica, California

• Atomic Corporation of America, Panorama City, California

• U. S. Testing (UST), Richland, Washington

Information on the periods during which ETEC used these laboratories was not found. The name of 

the laboratory responsible for the result appears in at least some early bioassay records. Due to 

various problems with the other labs (Fisher 1963), it appears that BioScience Laboratories analyzed 

the most samples early in this period. In 1964, ETEC initiated a contract with UST, which became 

International Technology Analytical Services, Quanterra Environmental Services, and finally Severn 

Trent Laboratories. From 1965 on, UST appears to have been the main laboratory vendor. Only 

information on uranium analyses was found for the in-house lab.
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AI Uranium Methods [A and B (fluorometric) and enriched uranium (radiometric)] 

The use of “A” and “B” to designate uranium bioassay methods apparently changed over the years. 

In general, it should be easy to distinguish fluorometric methods if the units were reported; 

fluorometric results were reported in micrograms or milligrams per unit volume rather than a unit of 

radioactivity per unit volume. The fluorometric methods fused uranium from raw urine with sodium 

fluoride and measured the fluorescence when the compound was exposed to ultraviolet light. The 

uranium fluorescence can be quenched by certain heavy metal ions, and the primary difference 

between the two methods was that Method A considered these quenching effects negligible. Method 

B took into account possible quenching effects (Mason and Burr 1958).

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

involved digesting the urine, removing some of the interfering cations, and electroplating the uranium 

onto nickel disks. The disks were counted in a proportional counter (Mason and Burr 1958).

No specific information on sensitivities for the in-house laboratory was obtained. The values for the 

contractor laboratories which follow should be used for dose reconstruction since they should be 

comparable to the in-house laboratory.

Vendor Uranium Methods [codes: 1A (fluorometric) and 1B (radiometric)] 

The fluorometric method was implemented during the first half of 1948 at AEC Sites (ORAU 2004, p. 

24). In May 1959, U. S. Nuclear quoted a price for fluorometric analysis with a minimum measurable 

concentration of 0.002 μg/mL based on a minimum volume of 1 L (Shepard 1959). Controls for 

Radiation, Inc., quoted 0.0001 μg/mL based on a minimum volume of 150 mL (O’Brien 1959). NSEC

gave its minimum measurable concentration as 0.0002 μg/mL (Edelmann 1959).

The early radiometric methods generally used separation chemistry followed by counting on a gasflow proportional counter or a ZnS(Ag) scintillation counter (ORAU 2004, p. 25). Shepard (1959) 

quoted a minimum measurable concentration of 7.5 dpm/L for radiometric determination of enriched 

uranium. NSEC gave its minimum measurable concentration for enriched uranium as 2 dpm/24-hr 

sample (NSEC 1957).

Gross Alpha

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

NSEC gave its minimum measurable concentration as 0.2 cpm/mL (NSEC 1957). It is assumed that 

this is a typographical error and 0.2 dpm/mL was intended.

Gross Beta

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

NSEC gave its minimum measurable concentration as 1.0 cpm/mL (NSEC 1957). It is assumed that 

this is a typographical error and 1.0 dpm/mL was intended.

Mixed Fission Products

The early procedure for mixed fission products was generally to add strontium carrier to the aluminum 

oxide solution for the plutonium procedure, then precipitate lanthanum hydroxide. This procedure 

extracted rare earths and strontium with yields ranging from 90% for cerium to 23% for strontium. The 

dried planchet was counted for beta activity with an approximate minimum detectable amount (MDA) 

of 60 dpm/sample (ORAU 2004, p. 27). This procedure separated and counted radionuclides of 

alkaline earths and rare earths such as strontium, yttrium, barium, lanthanum, cerium, europium, and 

promethium. It might not have accounted for radionuclides of ruthenium, cesium, zinc, cobalt, 

manganese, niobium, or zirconium (ORAU 2004, p. 27). NSEC gave its minimum measurable 

concentration as 2.0 cpm/mL (NSEC 1957). It is assumed that this is a typographical error and 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 15 of 64

2.0 dpm/mL was intended. This is the same value NSEC quoted for gross gamma measurements, so 

the technique might not have used chemical separations.

Polonium-210

BioScience used electrodeposition on a nickel disk and counting in a low-background proportional 

counter. The sensitivity quoted was 0.001 dpm/mL (Lee 1963).

Plutonium

One of the first plutonium bioassay analysis consisted of lanthanum fluoride precipitation and 

thenoyltrifluoroacetone (TTA) extraction and gross alpha counting. Electrodeposition on a stainlesssteel disk was followed by counting with nuclear track emulsion (autoradiography). Plutonium 

bioassay at ETEC apparently started around October 1966 (Kellehar 1966a). The urine maximum 

permissible concentration (MPC) that indicated 1 MPBB was listed as 4 dpm/24 hr. Therefore, in 

keeping with normal practice at the time, 0.4 dpm/24 hr would be the sensitivity required. BioScience 

quoted a sensitivity for plutonium bioassay of 0.00006 dpm/mL, but stated that it would have to 

subcontract the analyses (Lee 1963).

Strontium-90

BioScience coprecipitated 90Sr and 90Y with calcium as oxalates. Yttrium-90 was isolated and purified 

using a “milking” procedure with tributylphosphate, assayed, and reported as 90Sr equivalent. The 

method assumed that 90Sr and 90Y were in secular equilibrium. If not, counting could be delayed for 

ingrowth. The sensitivity quoted was 0.1 dpm/mL (Lee 1963).

Tritium

Tritium was not a significant personnel exposure hazard from research reactor operations, primarily 

due to the low power levels. However, 15 kCi of tritium in a triple-walled container was in Room 416-

51 in Building 004 at the De Soto Facility until November 1967 (Alexander 1967a). The earliest viable 

method for tritium analysis seems to have been liquid scintillation counting of raw urine in a 

scintillation cocktail. NSEC gave its minimum measurable concentration as 2,220 cpm/mL (NSEC 

1957). It is assumed that this is a typographical error and 2,220 dpm/mL was intended.

Thorium

No details of early thorium analyses were recovered. Tracerlab analyzed three urine samples for AI in 

1959. The detection limit was 0.2 μg/125 mL sample (Tracerlab 1959).

5.3.1.3 1967 – 1974

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

these data were mainly statements of work for bioassay services. These documents are believed to 

refer to services offered by UST (Kellehar 1966b, 1967, Spielman 1968, Bales 1969, Staszeski 1971). 

It is uncertain if these were the only bioassay services contracted during this period, but the 

information should be usable for dose reconstructions. Many of the methods listed in this TBD were 

undoubtedly in place prior to 1967.

Uranium Fluorometric (Procedure A)

The sample was acidified. A 100-μL aliquot was analyzed directly. Recovery was 93% ±18%. The 

detection limit was 0.5 μg/L. A fluorophotometer was used to measure the uranium.
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Uranium Fluorometric (Procedure B)

Uranium was extracted from the ashed residual salts with methyl isobutyl ketone using a salting 

solution of acid lammonium hydroxide. A fluorophotometer was used to measure the uranium 

present. Recovery was 83% ±8%. The detection limit was 0.05 μg/L. 

Uranium Radiometric

Uranium was isolated as in Procedure B above. It was measured by a gas-flow proportional counter 

or a ZnS(Ag) scintillation counter. Recovery was 83% ±8%. The detection limit was 0.5 dpm/sample. 

At ETEC the standard sample volume per day was 1,500 mL. The result was the total alpha activity.

Plutonium (Procedure A)

Plutonium is isolated by coprecipitation as the fluoride, extraction with TTA, and electrodeposition. 

Autoradiography was used to detect the plutonium. Recovery was 82% ±14%. The detection limit 

was 0.05 dpm/sample. At ETEC the standard sample volume per day was 1,500 mL. Plutonium 

results would have included activity from 238Pu, 239Pu, and 240Pu, but not 241Pu or 241Am.

Plutonium (Procedure B)

Plutonium was isolated as in Procedure A, and counted in a gas-flow proportional counter. Recovery 

was 85% ±10%. The detection limit was 0.5 dpm/sample. At ETEC the standard sample volume per 

day was 1,500 mL. Plutonium results would have included activity from 238Pu, 239Pu, and 240Pu, but 

not 241Pu or 241Am.

Tritium

Tritium was determined by liquid scintillation counting of an aliquot of the sample. Recovery was 

100% ±10%. The detection limit was 5,000 dpm/mL. 

Polonium

Polonium was spontaneously deposited on a silver disk from a dilute hydrochloric acid solution of the 

residual salts. The disk was alpha-counted. Recovery was 93% ±7%. The detection limit was 

0.5 dpm/sample.

Strontium

Strontium was isolated by precipitation as the oxalate, and then as the nitrate. Yttrium was removed 

by a nitric acid wash. Barium was removed as the chromate. Strontium was precipitated as the 

carbonate and counted with a gas-flow proportional counter. This provided a gross strontium result. 

Recovery was 85% ±9%. The detection limit was 4 dpm/sample. After the in-growth of 90Y (in 

2 weeks), the 90Y was isolated as the hydroxide, and then as the oxalate, which was burned to the 

oxide and counted in a gas-flow proportional counter. The 90Sr was computed from the 90Y, which 

was in secular equilibrium at separation time. Recovery was 78% ±11%. The detection limit was 

4 dpm/ sample.

Thorium

Thorium was isolated by a double fluoride precipitation and extraction with TTA, followed by 

spectrophotometric determination with morin. Recovery was 78% ±12%. The detection limit was 

1.0 μg for a 1,000-mL sample. Thorium-232 was determined by planchetting the thorium fraction and 

alpha-counting. The detection limit was 0.5 dpm/sample.

Phosphorus-32

Phosphorus-32 was separated first as the phosphomolybdate and then as the magnesium ammonium 

phosphate, which was planchetted and beta-counted in a gas-flow proportional counter. Recovery 

was 86% ±7%. The detection limit was 4 dpm/sample.
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Sulfur-35

The sample was ashed to drive off tritium and carbon, and counted by liquid scintillation techniques. 

Recovery was 95% ± 10%. The detection limit was 10 dpm per sample for a 1-mL sample aliquot.

Carbon-14

Carbon-14 was determined by liquid scintillation techniques. Recovery was 65% ±15%. The 

detection limit was 10 dpm for a 1-mL sample aliquot.

Gross Alpha

Gross alpha was determined by extracting most actinides from a 9N nitric acid solution into diethyl 

ether. This provided recoveries from 80% to 99% of most actinides. The detection limit was 1.0 dpm/ 

sample.

Gross Beta

Gross beta was determined from a beta count of the ashed residual salts. A 40K correction could be 

made. Recovery was 95% ±5%. The detection limit was 2 dpm for a 50-mL sample.

Promethium-147

Promethium-147 was chemically isolated by precipitation as the fluoride and extraction into TTA, and 

counted by liquid scintillation techniques. Recovery was 83% ±10%. The detection limit was 

5 dpm/sample. In 1967, for at least one case of suspected promethium exposure (147Pm-oxide), 

bioassay analyses were performed on urine and fecal samples (Alexander 1967a).

Americium and Curium

Americium and curium were isolated from contaminating actinides by HDEHP in toluene extraction out 

of 4N HNO3. Americium and curium were extracted into the HDEHP toluene solution at a pH of 4.5 

and back-extracted using 4N HNO3, electrodeposited, and counted using alpha spectroscopy. The 

recovery was 80% ±15% with a detection limit of 0.5 dpm/sample.

Gross Fission Products

Gross fission products were precipitated as the oxalate from a basic solution, planchetted, and betacounted in a gas-flow proportional counter. Recovery for all fission products averaged 82% ±5%. 

The detection limit was 5 dpm/sample based on the counting efficiency of the radiologically critical 

isotopes 90Sr and 90Y. A “gamma scan” was utilized for routine detection of certain gamma-emitting 

radionuclides. The gamma scan was performed by counting a sample in the well of a NaI(Tl) crystal 

and feeding the pulses to a multichannel pulse-height analyzer. The simultaneous measurement of 

more than one radionuclide using gamma energy analysis precluded making accurate statements 

about detection limits and precision. The laboratory offered a table of detection limits per sample 

based on experience; this information appears in Attachment E. Lower detection limits than those in 

the table were available at extra cost. Other gamma-emitting radionuclides could be determined.

5.3.1.4 1975 – 1988

The following analytical methods were taken from a series of annual reports (Hart 1979, 1980a,b,c; 

Eggleston 1983, 1984; Tuttle 1985). The measurement “type” in parentheses appears in many 

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

listing was not found. The reports list positive results for the year and, in some cases, follow-up 

results. The follow-up results were examined to see if the lowest value reported as non-zero could be 

determined. In general, there were not enough data to provide a satisfactory result. Therefore, dose 

reconstructors will have to use the detection limits in the previous section or, in some cases, on the 

individual results.
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Uranium Radiometric and Fluorometric (UR, UF)

Uranium was extracted from an acidic solution of ashed urine using aluminum nitrate, tetrapropyl 

ammonium hydroxide, and methyl isobutyl ketone. The uranium was recovered by back-extracting 

into water by evaporating to ketone. The water solution was planchetted for alpha counting for the UR 

analysis. The result was the total alpha activity. Fluorometric analysis required removal of an 

appropriate aliquot of the water solution prior to planchetting for pelletizing with NaF-LiF. The pellet 

was analyzed for uranium with a fluorometer. Most uranium samples were apparently analyzed by 

using both techniques.

Mixed Fission Products (FP1)

Mixed fission products were precipitated from a basic oxalate media. By adjustment of pH and 

oxalate concentrations, elements that are amphoteric or that form oxalate complexes in the form of 

excess oxalate were also precipitated. Alkali metals such as 137Cs did not precipitate. In addition, 

volatile fission products such as 131I were lost. The precipitate was washed with NAOH and water and 

planchetted for counting.

Mixed Fission Products (FP2)

FP2 uses the same extraction procedure as FP1; however, the soluble oxalate precipitates were 

gamma-counted for 137Cs and other gamma emitters. The results from the FP1 analysis and the FP2 

analysis were summed and reported as a single value.

Mixed Fission Products (FP3)

This analysis is the same as FP2 except oxalate insoluble results were reported separately as FP3a 

and oxalate soluble results were reported separately as FP3b. The FP3a analysis was assumed to 

indicate 90Sr, but other radionuclides, such as 60Co, might also be detected. Further analysis was 

used to quantify 90Sr specifically and identify interfering radionuclides if significant quantities occurred. 

The FP3b analysis was selective for 137Cs, using gamma-ray spectroscopy (Tuttle 1988a).

Plutonium (PUA, PUB)

After reduction to plutonium (III) and (IV) with hydroxylamine hydrochloride, plutonium was 

precipitated with lanthanum fluoride. This isolated the plutonium from most elements, including 

uranium, except thorium, rare earths, and actinides. After oxidation of plutonium with sodium nitrate 

in acid media, extraction of plutonium was performed with 0.5 M TTA in xylene. Following extraction, 

the aqueous solution containing plutonium was neutralized and concentrated by heating. After 

oxidation of the plutonium in a basic media, it was electrodeposited on a stainless-steel disc. The 

plutonium activity was determined by autoradiography (PUA) for greater sensitivity or counted for 

alpha radiation with a proportional counter (PUB). Plutonium results would have included activity from 

238Pu, 239Pu, and 240Pu, but not 241Pu. The PUA analysis was chemically selective for plutonium and 

excluded the 241Am that is generally present (Tuttle 1986b).

Gross Beta, High Level (GBH)

The gross sample was evaporated to dryness, followed by organic digestion by hydrogen peroxide 

and nitric acid. Natural potassium (40K) correction was determined by diluting the ashed salts to a 

known volume, and removing an aliquot for flame spectrophotometry. The remaining solution was 

evaporated to near dryness, planchetted, and counted for beta radiation with a proportional counter. 

The radioactivity in the urine sample due to 40K was subtracted from the gross count.

Gross Alpha (GA1a)

This analysis was specific for uranium and/or plutonium, which were extracted from ashed urine salts 

using aluminum nitrate, tetrapropylammonium hydroxide, and methyl isobutyl ketone. Transuranics 

did not extract to any appreciable extent. Uranium and/or plutonium were recovered by back-
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 19 of 64

extracting into water by evaporating the ketone. The uranium or plutonium was electrodeposited on a 

stainless-steel disc and autoradiographed.

Gross Alpha (GA1b)

This was the same analysis as GA1a, except the extraction solution was planchetted and counted for 

alpha radiation with a proportional counter.

Gross Alpha (GA2)

This analysis was specific for all alpha emitters. Metabolized actinides were converted to states 

suitable for coprecipitation with alkaline earth phosphates by digesting the gross urine sample in 10% 

nitric acid. The actinides were coprecipitated with calcium phosphate by neutralizing the acid solution 

with ammonia. The precipitate was washed, planchetted, and counted for alpha radiation with a 

proportional counter.

5.3.1.5 1989 – 2005

By 1989, all R&D activities had ended. All work with radioactive materials has been in conjunction 

with ongoing decontamination and decommissioning activities. UST was apparently the bioassay 

contractor between 1989 and 1992. Then ETEC switched to Teledyne (now Teledyne Brown 

Engineering). No analytical procedures were recovered, but they were probably based on the earlier 

methods presented above. The collection methods, etc., appear to be very similar to those used in 

previous years. The routine sampling strategy was to use generic screening urinalyses (gross alpha 

and gross beta minus 40K). If positive results were found, dose assessments were performed using 

default radionuclide mixtures or more specific analyses were performed (Barnes 1999). Periodic in

vivo measurements were performed to detect intakes from gamma emitters likely to be present at 

ETEC (137Cs, 60Co, 154Eu, etc.).

Particle size measurements were generally not performed in conjunction with routine workplace 

monitoring. Worker breathing zone air samples were collected (Barnes 1999), but it is unclear if this 

information was incorporated into an individual’s dosimetry records.

Since 1992, bioassay results returned from the vendor have been entered into a computer database 

(Boeing Canoga Park Internal Dosimetry Tracking System). Results before 1992 are available only in 

hard copy (Barnes 1999). In general, the entire sample was not used for the requested analysis; 

therefore, multiple results for one sample are possible. 

According to the health physicist responsible for internal dosimetry in 2005, no positive results have 

been reported since 1998. The sensitivities required for each analysis were published (Barnes 1999) 

and are listed in Table 5-6 in Section 5.6. 

5.3.2

The methods for individual radionuclides are covered for the specific periods under the sample types 

(i.e., urine and fecal).

In Vitro Methods for Individual Radionuclides

5.3.3

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

documents, little detail has been found about the analytical methods used. A statement in the 1967 

bioassay contract for UST indicated that all bioassay samples are wet ashed with nitric acid and 

Fecal Sample Analysis
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hydrogen peroxide. A salt fusion was added for fecal samples to ensure recovery of all radionuclides. 

An aliquot was apparently analyzed using methods similar to those discussed for urine sampling.

Detection limits for various isotopes for fecal samples were found for 1967. In the period from 1975 to 

1988 when the number of total and positive tests was well-documented, only a few fecal samples 

were mentioned, but this might have been because this series of reports did not always make the 

distinction between fecal and urine, which was by far the more common bioassay medium.

In 1967, bioassay analyses were performed on fecal samples for a case of suspected promethium 

exposure (Alexander 1967b). Fecal samples for uranium were apparently taken in conjunction with 

the investigation into the uranium aluminide exposures in 1967 (see Attachment A). Tables of total 

samples and positive results for 1975 and 1976 show analysis “type” codes for uranium fecal samples 

(F-UF and F-UR), but apparently no such analyses occurred in those years (Hart 1979, 1980a). In 

1982, positive PUA fecal results were reported for two individuals (one sample each) (Tuttle 1985).

Due to the lack of information found, the urine MDAs for gross alpha or gross beta listed in Table 5-3 

in Section 5.6 should be converted to dpm/day and used for fecal analyses prior to 1967 if the 

sensitivity was not reported by the laboratory. From 1967 on, the values in Table 5-4 in Section 5.6

should provide claimant-favorable estimates because detection limits probably were lower with time.

Table 5-1 lists the frequencies for in vitro monitoring.

Table 5-1. Internal dose control program (in vitro). 

Routine monitoring type Period Frequency

Urine, single voida 1958–1988 Quarterly, semiannual, annual, based on job

Urine, single void, H-3 1958–1966 Weekly, based on job

a. Sample requested on Friday, first voiding on Monday morning requested. Positive samples were 

verified and followed with at least one 24-hr sample collected on Sunday.

5.4 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PRACTICES

5.4.1

Although Helgeson Nuclear Services regularly visited the site to conduct chest counts for 235U (see 

Section 5.3.2), whole-body counting for fission or activation products was apparently not part of the 

routine bioassay program at ETEC. Between 1975 and 1988, only 25 counts on 25 individuals were 

summarized in annual reports. In the same period, 385 chest counts were completed. All WBCs

were reported positive for 137Cs. Ten counts were performed in 1977 and 15 were performed in 1979. 

The counts were probably performed when intakes were suspected, but could have been projectending counts. Although only 137Cs was reported in the summaries, undoubtedly a wide spectrum 

was scanned that would have detected other high-energy gamma emitters, such as 60Co, with a 

reasonably low detection level. Barnes (1999) stated that WBCs occurred “periodically.” 

Whole-Body Counting

5.4.2

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

The 186-keV gamma ray from the decay of 235U was used to quantify the amount of EU in the lung by 

scintillation spectrometry (Tuttle 1986a). Calibration of this system was crude; it used a point source 

rather than a distributed source that would better simulate the radioactivity in a lung. The point source 

contained more 234U than the uranium fuel, so the activity-to-mass conversion was off by about a 

factor of 2. Moreover, no chest-wall thickness corrections were made. The first two corrections 

Chest Counting
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amounted to initial lung burdens that were about a factor of 4 too high. The correction for chest-wall 

thickness would have depended on each individual. This system was not used after 1967. It is 

uncertain to what extent corrections were made to individual employee records. Dose reconstructors

could use these data for claimant-favorable overestimates. For more realistic calculations, it would be 

best to rely on urine and fecal data that would hopefully be available.

Starting in 1968, Helgeson Nuclear Services provided lung counting services using equipment and 

techniques specifically developed to measure lung deposits of uranium. The first counts in 1968 were 

done with 0.5-in. by 8-in. NaI(Tl) detector in the shadow shield used for whole-body counting. A crude 

masonite phantom was used for calibration, and included a chest-wall thickness correction. The 

individual was raised on an air mattress until the chest was flush with the bottom of the detector. The 

counts took 40 min. The results were reported in milligrams of 235U ±2 sigma, with the uncertainty 

based on counting statistics alone (Helgeson 1968).

By 1977, two 5-in.-diameter, thin-window phoswich detectors were used, which provided a reduction 

in Compton scattered background by a factor of 4 in the 235U region. In addition, the counting 

chamber was totally shielded. These changes enabled reduction of the counting time to 20 min. The 

calibration was similar to that described for 1968, but was cross-checked with the REMAB phantom 

used at Oak Ridge National Laboratory (ORNL). The minimum sensitivity was reported to be 

30-60 μg of 235U and 1 to 3 mg of natural or depleted uranium (Helgeson 1978, 1983). Calibration 

was performed with a phantom containing a known amount of 235U and a known chest-wall thickness. 

The gross pulse-height spectrum obtained was adjusted for background by subtracting an assumed 

background spectrum, consisting of a straight line passing through the gross count values just below 

and just above the pulse-height spectrum region corresponding to 186 keV. The net count was 

adjusted for chest-wall thickness and converted to mass of 235U, based on the calibration (Tuttle 

1986a).

Until 1981, the results of lung count measurements were reported as zero if a result was below the 2-

sigma uncertainty based on counting statistics. The results were reported as the actual value if equal 

to or greater than the 2-sigma uncertainty. Starting in 1981, Helgeson Nuclear Services was asked to 

report all values, regardless of the assigned uncertainty. An analysis of noncensored data along with 

counts of three individuals with no history of uranium exposure was reported in 1986 (Tuttle 1986a). 

This analysis estimated that the Helgeson results were biased high by 32.5 μg of 235U when only 

natural levels were present. This suggested that the straight-line approach to subtracting background 

was not appropriate when measuring background subjects. At values approaching the maximum 

permissible lung burden (MPLB), the background subtraction method was less important. The MPLB 

for 235U was 245 μg. Dose reconstructors can consider the values reported by Helgeson a claimantfavorable overestimate. For a more realistic estimate, adjust the lung count results by assuming a 

linear relationship of the bias between 0 and 1 MPLB, as indicated in Attachment C. Above 1 MPLB, 

no adjustment in the results is recommended.

Helgeson apparently also did special counts as requested by the site. These included counts for 

241Am and may have included counts for plutonium and thorium as well. Where these counts appear 

in claimant records, dose reconstructors should consider a “less than” value as the reporting level 

(decision level) and the MDA should be taken as twice the “less than” value.

Table 5-2 lists the frequencies for in vivo monitoring.
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Table 5-2. Internal dose control program (in vivo). 

Routine monitoring type Period Frequency

Chest count U-235 (UCLA) 1967–1968a Monthly as long as count indicated > 

1⁄2 of lung burden (0.02 μCi)

Chest count U-235 (ORNL) 1967b One time

Chest count U-235 (Helgeson) 1968–1983 (enriched U 

work ended in 1983)

Three times/yr, 14-21 personnel 

selected for measurement each time

WBC Cs-137 (Helgeson) 1977, 1979 Uncertainc

WBC 1989–2005 Periodicallyd

a. Counts performed to follow up suspected enriched UAlx exposures in the “powder room.” The UCLA counter was not 

used after April 1968 (Tschaeche 1968a). 

b. Counts of one employee as a cross-check on the UCLA calibration (Alexander 1967a). 

c. It is unknown if these counts were routine or special. 

d. Frequency not specified in Barnes (1999). 

5.5 UNCERTAINTY

At ETEC the uncertainty for a single bioassay measurement was not reported consistently. Reviewed 

statements of work for bioassay services do not contain any specification for reporting uncertainty. 

UST results were apparently reported with a 2-sigma uncertainty if they were above the minimum 

detectable level. BioScience reported a 95% confidence interval (at least in some cases). The 

Individual Analysis Results Sheets (Figures B-5 to B-10, Attachment B) might show the uncertainty for 

individual measurements. Due to the calibration and other problems discussed above, the estimated 

error in the early UCLA lung counting results for 235U was ±200% at 1 sigma. The stated uncertainty 

for the Helgeson lung counts was about ±25% at 1 sigma (Tschaeche 1968b).

5.6 DETECTION LIMITS 

Urine results above the lower limit of detection (LLD), sensitivity, minimum detectable level, etc., were 

used to calculate the percentage of the MPBB received by the worker. The normalized result 

(1,500 mL/day) was divided by a standard excretion rate for one MPBB to produce this percentage. If

possible, the worker followed with additional urine samples until the results were less than the 

laboratory LLD, the lowest value at which radioactivity was considered to be present with reasonable 

certainty (Tuttle 1989). No exact definitions of these terms have been found. In effect, the detection 

limit value was used as the reporting level. Most results appear to have been reported with an 

uncertainty of 2-sigma error or a 95% confidence interval. If the uncertainty included zero, the result 

was considered to be background (i.e., no radioactivity detected in the sample). Assuming the 

uncertainties in the background and the sample were equal, the detection limits published in site 

documents were closer to a decision level. That is, there is only an approximate 5% chance that 

results at this level are really background results (false positives). The MDAs would be approximately 

twice these values to ensure that there is only an approximate 5% chance that results at the MDA 

would not be detected (false negatives). The detection limits recovered are listed in the tables as the 

reporting levels. Since the detection limits in the tables were mainly collected from contract 

documents, the reported sensitivity (“less than” values) in the claimant records should be used to 

determine MDA in lieu of the values listed in these tables. 

Tables 5-3 through 5-6 list MDAs and reporting levels for periods corresponding to the bioassay 

methods discussed in Sections 5.3 and 5.4. The reporting levels are listed in the units quoted in the 

references, which are generally the units of the results. However, various volumes were used to 

report the results. In general, the excretions assumed when reporting “per sample” or “per day” were

1,500 mL for urine samples and 135 g for fecal samples.
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Table 5-3. Detection limits 1958–1966a. 

Radionuclide

Method/

description MDAb Reporting levelc

Gross alpha Urine 15 dpm/L (U. S. Nuclear)

4 dpm/sample (NSEC)

7.5 dpm/L

2 dpm/sample

Gross beta Urine 150 dpm/L (U. S. Nuclear)

2 dpm/mL (NSEC)

75 dpm/L

1 dpm/mL

Gross beta (minus K-40) Urine 0.4 dpm/mL (BioScience)

0.04 dpm/mL (UST)

0.2 dpm/mL

0.02 dpm/mL

H-3 Urine 4,440 dpm/mL (NSEC)

10,000 dpm/mL (BioScience)

10,000 dpm/mL (UST)

2,220 dpm/mL

5,000 dpm/mL

5,000 dpm/mL

MFP (gross) Urine 4 dpm/mL (NSEC)

60 dpm/sample (Hanford)

0.2 dpm/mL (BioScience)

0.04 dpm/mL (UST)

2 dpm/mL

30 dpm/sample

0.1 dpm/mL

0.02 dpm/mL

Po-210 Urine 0.02 dpm/mL (BioScience)

0.02 dpm/mL (UST)

0.01 dpm/mL

0.01 dpm/mL

Plutonium Urine 0.0006 dpm/mL (UST) 0.0003 dpm/mL

Sr-90 Urine 0.2 dpm/mL (BioScience)

0.04 dpm/mL (UST)

0.1 dpm/mL

0.02 dpm/mL

Thorium Urine 0.4 μg/125 mL (Tracerlab)

0.0004 μg/mL (BioScience)

0.0004 μg/mL (UST)

0.2 μg/125 mL

0.0002 μg/mL

0.0002 μg/mL

Uranium UF-1A

Urine

0.004 μg/mL (U. S. Nuclear)

0.0002 μg/mL (Controls for Rad.)

0.0004 μg/mL (NSEC)

0.0004 μg/mL (BioScience)

0.0004 μg/mL (UST)

0.002 μg/mL

0.0001 μg/mL

0.0002 μg/mL

0.0002 μg/mL

0.0002 μg/mL

Uranium (enriched) UR-1B

Urine

15 dpm/L (U. S. Nuclear)

4 dpm/sample (NSEC)

0.012 dpm/mL (BioScience)

0.012 dpm/mL (UST)

7.5 dpm/L

2 dpm/sample

0.006 dpm/mL

0.006 dpm/mL

a. The date of the reference to the reporting levels, which are the best information available for the period 1958-1966. 

b. Assumed to be twice the sensitivity (see text). The reported sensitivity (“less than” values) in the claimant records 

should be used to determine MDA in lieu of the values listed in this table. These values should also be used for the inhouse laboratory.

c. U. S. Nuclear from Shepard (1959), Controls for Radiation, Inc., from O’Brien (1959), NSEC from Edelmann (1959) and 

NSEC (1957), BioScience from Lee (1963), and Tracerlab from Tracerlab (1959). UST results that were considered to 

be less than detectable generally have the reporting level included with the result.

5.7 UNIT CODES AND DESCRIPTION OF UNITS

The bioassay results from ETEC and its predecessor organizations are apparently not available in a 

computerized format. The units used to report the results are generally included in the hard-copy 

reports. Table 5-7 reflects this information.

5.8 EXCRETA SAMPLE KIT CODES

No codes have been found. Table 5-8 lists sample kit information summarized from various site 

documents.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 24 of 64

Table 5-4. Detection limits 1967–1974a. 

Radionuclide

Method/

description MDAb Reporting levelc

Ca-45 Urine 10 dpm/mL 5 dpm/mL

Ca-45 Urine 0.06 dpm/mL 0.03 dpm/mL

Ca-45 Feces 30 dpm/sample 15 dpm/sample

Cs-137 FP3b

Urine

60 dpm/sample 30 dpm/sample

Gross alpha Urine 0.02 dpm/mL 0.01 dpm/mL

Gross alpha Urine 0.010 dpm/mL 0.005 dpm/mL

Gross alpha Feces 4 dpm/sample 2 dpm/sample

Gross beta (less K-40) Urine 0.08 dpm/mL 0.04 dpm/mL

Gross beta Feces 4 dpm/sample 2 dpm/sample

H-3 Urine 10,000 dpm/mL 5,000 dpm/mL

I-131 Urine 0.12 dpm/mL 0.06 dpm/mL

I-131 Urine 0.4 dpm/mL 0.2 dpm/mL

MFP (gross) Urine 0.04 dpm/mL 0.02 dpm/mL

MFP (gross) Feces 16 dpm/sample 8 dpm/sample

Plutonium Urine 0.8 dpm/sample (estimated) 0.4 dpm/sample 

(estimated)

Plutonium Radiographic

Urine

0.0006 dpm/mL 0.0003 dpm/mL

Plutonium Procedure A

Feces

2 dpm/sample 1 dpm/sample

Plutonium Procedure B

Feces

4 dpm/sample 2 dpm/sample

Polonium Urine 0.02 dpm/mL 0.01 dpm/mL

Polonium Feces 6 dpm/sample 3 dpm/sample

Pm-147 Urine 0.10 dpm/sample 0.05 dpm/sample

Radium Urine 0.010 dpm/mL 0.005 dpm/mL

Radium Feces 20 dpm/sample 10 dpm/sample

Sr-90 Urine 0.04 dpm/mL 0.02 dpm/mL

Sr-90 Feces 16 dpm/sample 8 dpm/sample

Sr-90 FP3a 

Urine

60 dpm/day 30 dpm/day 

Thorium Urine 0.002 μg/mL 0.001 μg/mL

Thorium Feces 6 μg/sample 3 μg/sample

Uranium UF Urine 0.006 μg/mL 

(10-mL volume)

0.003 μg/mL 

(10-mL volume)

Uranium UF Urine 0.0004 μg/mL

(100-mL volume)

0.0002 μg/mL

(100-mL volume)

Uranium UF Feces 4 μg/sample 2 μg/sample

Uranium (enriched) UR Urine 0.012 dpm/mL 0.006 dpm/mL

Uranium (enriched) UR Feces 4 dpm/sample 2 dpm/sample

U-235 IVLC (UCLA) 0.8 mg 0.4 mg

U-235 IVLC 

(Helgeson)

60-120 μg (depends on 

chest-wall thickness)

30-60 μg

a. The date of the reference to the reporting levels, which are the best information available for the period 1967-1974. 

b. Assumed to be twice the sensitivity (see text). The reported sensitivity (“less than” values) in the claimant records 

should be used to determine MDA in lieu of the values listed in this table.

c. U-235 IVLC (UCLA) from Saxe (1967a). In vitro values are primarily from UST contract documents (Kellehar 1966b, 

1967, Spielman 1968, Bales 1969, Staszesky 1971).
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 25 of 64

5.9 SOLUBILITY TYPE, FRACTION ACTIVITY, AND PARTICLE SIZE BY FACILITY

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

classes and particle size values from the International Commission on Radiological Protection 

(NIOSH 2002, pp. 15, 16). With one exception, facility-specific solubility and particle size data for 

ETEC has not been found. Activity fractions were not available with the exception of those for limited 

fuel fabrication operations. For highly enriched research reactors, values from Shleien, Slayback, and 

Birky (1998) have been listed as an approximation. In some cases, default assumptions from Barnes 

(1999, Table D-3) have been used to estimate the activation products using a ratio to the fission 

products. Table 5-9 lists this information.

Table 5-5. Detection limits 1975–1988a

. 

Radionuclide

Method/

description MDAb Reporting levelc

Cs-137 FP3b

Urine

60 dpm/sample 30 dpm/sample

Cs-137 WBC 4 nCi 2 nCi

Plutonium PUA

Urine

0.0990 dpm/sample 0.0495 dpm/sample

Sr-90 FP3a 

Urine

60 dpm/day 30 dpm/day 

U-235 IVLC (Helgeson) 60-120 μg (depends on chestwall thickness)30-60 μg 

Uranium (total) UF 

Urine

0.60 μg/day 0.30 μg/day

Uranium (enriched) UR 

Urine

7.5 dpm/day 3.75 dpm/day

a. The date of the reference to the reporting levels, which are the best information available for the period 1975-1988. 

b. Assumed to be twice the sensitivity (see text). The reported sensitivity (“less than” values) in the claimant records 

should be used to determine MDA in lieu of the values listed in this table.

c. Sr-90, U (total), and EU (1987, 1988) from Tuttle (1988b, 1989). 

Table 5-6. Detection limits 1989–2005a. 

Radionuclide

Method/

description MDAb Reporting levelc

Gross alpha Urine 15 dpm/1,500 mL 7.5 dpm/1,500 mL

Gross beta Urine 30 dpm/1,500 mL 15 dpm/1,500 mL

Gross beta, corrected for K-40 Urine 45 dpm/1,500 mL 22.5 dpm/1,500 mL

Gamma scan Urine 60 dpm/1,500 mL 30 dpm/1,500 mL

Sr-90 Urine 6 dpm/1,500 mL 3 dpm/1,500 mL

Uranium fluorometric Urine 15 μg/1,500 mL 7.5 μg/1,500 mL

Uranium alpha spec. Urine 3 dpm/1,500 mL 1.5 dpm/1,500 mL

Thorium alpha spec. Urine 3 dpm/1,500 mL 1.5 dpm/1,500 mL

Plutonium alpha spec. Urine 3 dpm/1,500 mL 1.5 dpm/1,500 mL

Am-241 Urine 0.90 dpm/1,500 mL 0.45 dpm/1,500 mL

H-3, distillation Urine 3000 dpm/1,500 mL 1,500 dpm/1,500 

mL

H-3, electrolytic enrichment Urine 60 dpm/1,500 mL 30 dpm/1,500 mL

WBC (Co-60) WBC 4 nCi 2 nCi

a. The date of the reference to the reporting levels, which are the best information available for the period 1989-2005. 

b. Assumed to be twice the sensitivity (see text). The reported sensitivity (“less than” values) in the claimant records 

should be used to determine MDA in lieu of the values listed in this table.

c. Sensitivity reported in Barnes (1999, Table D-1). 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 26 of 64

Table 5-7. Unit codes and description of units. 

Computer code Description of units

NA NA

Several non-nuclear energy research programs were conducted in Area IV. The primary non-nuclear 

R&D activities were performed at the Liquid Metal Engineering Center, which became ETEC . The 

facilities that performed non-nuclear research are not included in Table 5-9 unless they were also 

used for nuclear research (Sapere and Boeing 2005, p. 2-12). Dose reconstructors should use

general environmental radioactivity levels to assess personnel working in non-nuclear areas 

exclusively.

Table 5-8. Excreta sample kit codes.

Kit codea Media Sample description

NA Urine Routine samples. Single voiding collected on Monday morning before returning to 

work. Collected in single bottle. Also referred to as “rate samples.” Positive 

samples were followed up with additional rate sample for verification. Time of 

sample collection and previous voiding were recorded. Normally collected in a 16-

oz container. By 1999, 1-L polyethylene containers were used. It appears that one 

full container (900-1000 mL) was collected and collection time noted.

NA Urine 24-hr samples. Used for follow-up to verified rate sample or for incidents. One or 

more samples could be requested. Single samples were collected at home on 

Sunday. Collected in 32-oz or 1-L polyethylene containers.

NA Urine Spot samples. For follow-up to incidents, spot sample could be collected as soon as 

possible. This sample was probably collected in 16-oz container.

NA Feces Could be requested in conjunction with urine samples as follow-up to incidents. No 

descriptions of historical fecal kits were located. By 1999, single voiding samples 

were collected in 83-oz polyethylene containers. The minimum mass that was 

considered adequate was 30 g.

a. NA means that no codes to identify sample kits were used at ETEC. For a time requests were color-coded but these do 

not appear in the individual records

In the 1960s, a test was conducted that involved dropping a 1-kg depleted uranium slug from a 

helicopter. Apparently the slug was never recovered. No contamination was found in the area 

(Sapere and Boeing 2005 [p. 2-13]). It is unlikely that this event would contribute significantly to the 

internal dose of occupationally exposed workers, and it has not been included.

5.10 FACILITY-SPECIFIC RADIONUCLIDE CONVERSIONS

AI fabricated fuel for the ATR and the Engineering Test Reactor (ETR) in 1966, 1967, and 1968 at the 

De Soto Facility in Room 1110-62, Building 101 (known at the time as Building 001). See 

Attachment A for details. A second fuel fabrication campaign for the ATR was completed in 1979 and 

1980. These are the only two operations for which facility-specific data were made available. These

data are listed in Attachment D and summarized in Table 5-10. 

5.11 WORKPLACE MONITORING DATA

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

use workplace monitoring data (NIOSH 2002). The following types of workplace data might be 

available for ETEC: breathing zone air samples, general area air samples, and surface contamination 

surveys. However, these data are not likely to be in individual exposure records. Data on respirator
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 27 of 64

Table 5-9. Solubility type, fraction activity, and particle size by facility. 

Facilitya Compoundb RNc

Solubility 

typed

Particle sized

(μ AMAD) Activity fractione

Downey Plant

Bldg. 001

WBNS reactor

1948–1955

Homogeneous water boiler-type reactors 

- 93%-enriched uranyl sulfate solution in 

H2O. Reactors operated at low power 

levels (4-10 W).

Sr-90 to be determined 5 4.89E-01

Cs-137 to be determined 5 5.10E-01

U-234 to be determined 5 5.36E-08

U-235 to be determined 5 1.24E-06

U-238 to be determined 5 1.93E-08

Vanowen Fac.

Bldg. 038

L-47, L-77 

reactors

1956–1960

Pu-238 to be determined 5 6.62E-04

Pu-239 to be determined 5 1.22E-05

Pu-240 to be determined 5 5.73E-06

Pu-241 to be determined 5 3.33E-04

Am-241 to be determined 5 1.16E-06

De Soto Fac.

Bldg. 104

L-77 reactor

1960–1979?

Bldg. 4073 

Area IV, SSFL

KEWB reactor

1956–1966

Bldg. 4093

L-85 reactor

1956–1980

De Soto Fac.

Bldg. 101

Powder room & 

new powder 

room 1966-1968

& 1979–1982

Uranium Aluminide Fuel Fabrication

Alloy of uranium and aluminum

U-234 S 1 See Table 5-10

U-235 S 1 See Table 5-10

U-236 S 1 See Table 5-10

U-238 S 1 See Table 5-10

Bldg. 4143

SRE

1957–1964

Decontaminated

1974–1983

Storage facility

1983–1999

Demolished

1999

High-temperature, sodium-cooled, 

graphite-moderated EU reactor (site of 

loss-of-coolant accident in 1959). 

Unalloyed uranium metal thermally 

bonded by NaK in stainless-steel tubes.

1957-July 1959, core was 2.78% EU

September 1960, 2nd core began 

operation. Core was 7.6 % (weight) Th232 with 93% EU (Atomics International, 

1959)

Sr-90 to be determined 5 to be determined

Cs-137 to be determined 5 to be determined

U-234 to be determined 5 to be determined

U-235 to be determined 5 to be determined

U-238 to be determined 5 to be determined

Pu-238 to be determined 5 to be determined

Pu-239 to be determined 5 to be determined

Pu-240 to be determined 5 to be determined

Pu-241 to be determined 5 to be determined

Am-241 to be determined 5 to be determined

Co-60 to be determined 5 to be determined

Eu-152 to be determined 5 to be determined

Eu-154 to be determined 5 to be determined

Th-232 to be determined 5 to be determined

H-3 to be determined

Bldg. 4010

SER

1959–1960

S8ER

1963–1965

SNAP reactors – “fully enriched” uranium 

dispersed in zirconium hydride fuel rods

Sr-90 to be determined 5 4.60E-01

Cs-137 to be determined 5 4.78E-01

U-234 to be determined 5 5.03E-08

U-235 to be determined 5 1.16E-06

U-238 to be determined 5 1.81E-08

Bldg. 4024

S2DR

1961–1962

S10FS

1965–1966

SNAP TTF

1971

Pu-238 to be determined 5 6.22E-04

Pu-239 to be determined 5 1.15E-05

Pu-240 to be determined 5 5.38E-06

Pu-241 to be determined 5 3.13E-04

Am-241 to be determined 5 1.09E-06

Co-60 to be determined 5 4.28E-03

Eu-152 to be determined 5 4.89E-02

Bldg. 4028

STR

1961–1964

STIR

1964–1972

Eu-154 to be determined 5 6.11E-03

H-3 1.95E-03
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 28 of 64

Table 5-9 (Continued). Solubility type, fraction activity, and particle size by facility.

Facilitya Compoundb RNc

Solubility 

typed

Particle sized

(μ AMAD) Activity fractione

Bldg. 4059

S8DR

1968–1969

Bldg. 4012

SNAP CTF

1962–1968

HMRFSR

1970–1972

Critical test facilities - SNAP development 

test facilities (fully enriched uranium 

dispersed in zirconium hydride fuel rods)

Sr-90 to be determined 5 4.89E-01

Cs-137 to be determined 5 5.10E-01

U-234 to be determined 5 5.36E-08

U-235 to be determined 5 1.24E-06

U-238 to be determined 5 1.93E-08

Bldg. 4373

SNAP CTF

1957–1963

Pu-238 to be determined 5 6.62E-04

Pu-239 to be determined 5 1.22E-05

Pu-240 to be determined 5 5.73E-06

Bldg. 4019

SNAP FSCF

1964–1965

Pu-241 to be determined 5 3.33E-04

Am-241 to be determined 5 1.16E-06

Bldg. 4009

OMR CF

1958–1967

Critical test facilities – Civilian nuclear 

power test facilities, EU low-power 

reactors. HEU and DU used and stored 

also.

Sr-90 to be determined 5 4.60E-01

Cs-137 to be determined 5 4.79E-01

U-234 to be determined 5 5.04E-08

U-235 to be determined 5 1.17E-06

SGR CF

1958–1967

U-238 to be determined 5 1.81E-08

Pu-238 to be determined 5 6.23E-04

Pu-239 to be determined 5 1.15E-05

High-energy rate 

forging

1980s

Pu-240 to be determined 5 5.39E-06

Pu-241 to be determined 5 3.13E-04

Am-241 to be determined 5 1.09E-06

Co-60 to be determined 5 4.29E-03

DU storage

Early 1990s

Eu-152 to be determined 5 4.90E-02

Eu-154 to be determined 5 6.12E-03

Bldg. 4100

AETR test facility

1960–1974

Critical test facilities – thorium- and 

uranium-fueled reactors. 20 different 

reactor core configurations studied. The 

AETR’s first 9 core configurations 

(through 1965) contained various 

amounts of U-233 and Th-232 and were 

driven by 93% enriched U fuel (Mountford 

and Morewitz, 1965). The activity 

fractions are the medians of the cores 

studied. Only very small amounts of the 

other radionuclides listed were likely to 

have been produced, but they were listed 

as ROCs by Boeing.

Sr-90 to be determined 5

Cs-137 to be determined 5

U-233 to be determined 5 9.81E-01

U-234 to be determined 5 1.83E-02

U-235 to be determined 5 5.71E-04

U-238 to be determined 5 6.63E-06

Radiation 

Instrument 

Calibration and 

Radiological 

Sample Counting 

Laboratory

1985–present

Pu-238 to be determined 5

Pu-239 to be determined 5

Pu-240 to be determined 5

Pu-241 to be determined 5 

Am-241 to be determined 5

Eu-152 to be determined 5

Eu-154 to be determined 5 

Th-232 to be determined 5 1.05E-03

H-3

Bldgs. 4003

4163, 4041, 

4654, 4689,

4653, 4606,

4773

SRE Support 

Complex

1954–1964

Nuclear support operations – SRE fuel 

assembly (uranium and thorium metal 

slugs), contained radioactive “hot cave,” 

tanks, hoods, and lines until 1975. SRE 

fuel loaded in 1960 was 7.6 % (weight) 

Th-232 with 93% EU (Aomics 

International, 1959)

Sr-90 to be determined 5 4.60E-01

Cs-137 to be determined 5 4.79E-01

U-234 to be determined 5 5.04E-08

U-235 to be determined 5 1.17E-06

U-238 to be determined 5 1.81E-08

Pu-238 to be determined 5 6.23E-04

Pu-239 to be determined 5 1.15E-05

Pu-240 to be determined 5 5.39E-06

Pu-241 to be determined 5 3.13E-04

Analysis of 

SNAP fuel burnup & irradiation 

experiments

1965–1973

Am-241 to be determined 5 1.09E-06

Co-60 to be determined 5 4.29E-03

Eu-152 to be determined 5 4.90E-02

Eu-154 to be determined 5 6.12E-03

Th-232 to be determined 5 to be determined

Bldg. 4005

Uranium Carbide 

Fuel Pilot Plant

1958-1993

Nuclear support operations – pilot plant 

for uranium-carbide fuel production in 

1966-1977, first using DU and then EU

U-234 to be determined 5 to be determined

U-235 to be determined 5 to be determined

U-238 to be determined 5 to be determined
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 29 of 64

Table 5-9 (Continued). Solubility type, fraction activity, and particle size by facility.

Facilitya Compoundb RNc

Solubility 

typed

Particle sized

(μ AMAD) Activity fractione

Bldg. 4011

Radiation 

Instrumentation 

Calibration 

Laboratory

1984–1996

Nuclear support operations – sealed 

sources. In 1960, a radioactive liquid spill 

from OMRE shipping cask occurred west 

of building. 

Sr-90 to be determined 5 4.60E-01

Cs-137 to be determined 5 4.79E-01

U-234 to be determined 5 5.04E-08

U-235 to be determined 5 1.17E-06

U-238 to be determined 5 1.81E-08

Pu-238 to be determined 5 6.23E-04

Pu-239 to be determined 5 1.15E-05

Pu-240 to be determined 5 5.39E-06

Pu-241 to be determined 5 3.13E-04

Am-241 to be determined 5 1.09E-06

Co-60 to be determined 5 4.29E-03

Eu-152 to be determined 5 4.90E-02

Eu-154 to be determined 5 6.12E-03

Bldg. 4030

Van de Graaff 

Accelerator

1960–1964

Nuclear support operations – neutron 

source, used H-3 target

H-3 1

Bldg. 4020

Hot Laboratory

1957–1988

D&D Completed

1999

Nuclear support operations – Used to 

examine fuel from SRE, SER, S2DR, 

S8DR, S10FS-3, and outside reactors. 

Fuel disassembled or separated from 

cladding.

Sr-90 to be determined 5 4.60E-01

Cs-137 to be determined 5 4.79E-01

U-234 to be determined 5 5.04E-08

U-235 to be determined 5 1.17E-06

U-238 to be determined 5 1.81E-08

Pu-238 to be determined 5 6.23E-04

Pu-239 to be determined 5 1.15E-05

Pu-240 to be determined 5 5.39E-06

Pu-241 to be determined 5 3.13E-04

Am-241 to be determined 5 1.09E-06

Co-60 to be determined 5 4.29E-03

Eu-152 to be determined 5 4.90E-02

Eu-154 to be determined 5 6.12E-03

Pm-147 to be determined 5 to be determined

Bldg. 4023

Liquid Metals

Component

Test Building

1962–1986

Nuclear support operations – tests with 

sodium loops containing radioactive 

contaminants 

Sr-90 to be determined 5 to be determined

Cs-137 to be determined 5 to be determined

Co-60 to be determined 5 to be determined

Eu-152 to be determined 5 to be determined

Eu-154 to be determined 5 to be determined

Bldg. 4029

Radioactive 

Measurement 

Facility

1959–1974

Nuclear support operations – Leaking 

calibration source contaminated source 

well

Ra-226 to be determined 5 1

Bldg. 4055

NMDF

1967–1979

Nuclear support operations –

Development work involving Pu

DU work 1967

U-234 to be determined 5 to be determined

U-235 to be determined 5 to be determined

U-238 to be determined 5 to be determined

Pu-238 to be determined 5 to be determined

Pu-239 to be determined 5 to be determined

Pu-240 to be determined 5 to be determined

Pu-241 to be determined 5 to be determined

Bldg. 4064

Fuel Storage 

Facility

1958–1993

D&D in 1997

Nuclear support operations – Secure 

storage of nonirradiated fuel (EU and Pu); 

following removal of fissionable material 

in mid-1980s, radioactive waste (soil) 

stored until 1993.

Cs-137 to be determined 5 to be determined

U-234 to be determined 5 to be determined

U-235 to be determined 5 to be determined

U-238 to be determined 5 to be determined

Pu-238 to be determined 5 to be determined

Pu-239 to be determined 5 to be determined

Pu-240 to be determined 5 to be determined

Pu-241 to be determined 5 to be determined

Am-241 to be determined 5 to be determined

Co-60 to be determined 5 to be determined

Eu-152 to be determined 5 to be determined

Eu-154 to be determined 5 to be determined
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 30 of 64

Table 5-9 (Continued). Solubility type, fraction activity, and particle size by facility.

Facilitya Compoundb RNc

Solubility 

typed

Particle sized

(μ AMAD) Activity fractione

Bldgs. 4021, 

4022

RMHF

1959–present

Nuclear support operations – Radioactive 

waste processing from onsite programs; 

2005 storage area from D&D activities

Sr-90 to be determined 5 4.60E-01

Cs-137 to be determined 5 4.78E-01

U-234 to be determined 5 5.03E-08

U-235 to be determined 5 1.16E-06

U-238 to be determined 5 1.81E-08

Pu-238 to be determined 5 6.22E-04

Pu-239 to be determined 5 1.15E-05

Pu-240 to be determined 5 5.38E-06

Pu-241 to be determined 5 3.13E-04

Am-241 to be determined 5 1.09E-06

Co-60 to be determined 5 4.28E-03

Eu-152 to be determined 5 4.89E-02

Eu-154 to be determined 5 6.11E-03

H-3 1.95E-03

Bldg. 4363

Mechanical 

Component 

Development & 

Counting

1956–1963

Nuclear support operations – In 1962, 

work done on contaminated component 

from SRE accident

Sr-90 to be determined 5 0.5

Cs-137 to be determined 5 0.5

17th Street 

Drainage

1959?–1999

Areas of known contamination – runoff 

from SNAP facilities

Cs-137 to be determined 5 1.0

OCY

Late 1960s–late 

1970s

Areas of known contamination – spill 

detected in 1976 (mixed fission products 

suspected)

Sr-90 to be determined 5 0.5

Cs-137 to be determined 5 0.5

Bldg. 4886

Former Sodium 

Disposal Facility

1956–1978

Areas of known contamination – storage 

drums contaminated with residual 

radioactivity. SRE fuel loaded in 1960 

was 7.6 % (weight) Th-232 with 93% EU 

(Atomics International, 1959)

Sr-90 to be determined 5 to be determined

Cs-137 to be determined 5 to be determined

U-234 to be determined 5 to be determined

U-235 to be determined 5 to be determined

U-238 to be determined 5 to be determined

Pu-238 to be determined 5 to be determined

Pu-239 to be determined 5 to be determined

Pu-240 to be determined 5 to be determined

Pu-241 to be determined 5 to be determined

Th-232 to be determined 5 to be determined

H-3 to be determined

a. Facilities were combined for this analysis if they were similar and had a common list of radionuclides of concern.

b. If chemical compounds were not available, the best description found is listed. 

c. Radionuclides are radionuclides of concern from Sapere and Boeing (2005, Tables 2-1, 2-2 and Section 2.0). Radionuclides were not included in the 

radionuclides of concern if they were an activation product limited to steel and rebar, had a short half-life, were a naturally occurring isotope consistent with 

natural background, or had not been observed in soil measurements.

d. Defaults used except for the powder room.

e. Activity fractions for fission products calculated from values published for a research reactor and activation products (Co-60 and Eu-152) calculated from 

Eu-154 ratios in Barnes (1999). A technical information bulletin on fission product nuclides to assign across the complex is to be published.

use are not likely to be available. Quantitative fit testing information was not located. In the case of 

surface contamination data, site/process-specific resuspension factors are not likely to be available.

5.12 SOURCE TERM DATA

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

source term evaluations (NIOSH 2002, p. 19). Data on the amount of dispersible material available 

does not appear to be available for ETEC.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 31 of 64

Table 5-10. Facility-specific radionuclide conversions. 

Process description

Activity per unit mass (Bq/g uranium)

U-234 U-235 U-236 U-238

Enriched uranium fuel fabrication, Room 1110-62 

(powder room), Building 101, De Soto Facility, 1966–

1967

1.35E+06 7.44E+04 7.31E+03 7.50E+02

ATR enriched uranium fuel fabrication, 1979–1982 

Building 101, De Soto Facility

2.37E+06 7.44E+04 1.14E+04 6.79E+02

5.13 RADON

For dose reconstruction under EEOICPA, occupational radon exposure is exposure to radon 

emanating from sources other than those naturally occurring in the area. Dose reconstructors must 

subtract the natural background level of radon exposure from any measured values when assessing 

occupational exposure (NIOSH 2002, p. 32). ETEC was not a processing or storage location for large 

quantities of 226Ra or 222Rn. Radon measurements made from December 1989 to February 1990 in 

several facilities indicated that none of the areas measured exceeded 2 pCi/L and all but one was less 

than 1 pCi/L. 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 32 of 64

REFERENCES

Alexander, R. E., 1959, Radiochemical Analysis for Radioisotopes in Biological Specimens, TDR No. 

3678, Atomics International, Canoga Park, California, March 25. 

Alexander, R. E., 1967a, “Radiation Safety Unit Weekly Newsletter for Period Ending November 18, 

1967,” Atomics International, Canoga Park, California, November 28. 

Alexander, R. E., 1967b, “Radiation Safety Unit Weekly Newsletter for Period Ending November 25, 

1967,” Atomics International, Canoga Park, California, November 30. 

Alexander, R. E., 1967c, “Radiation Safety Unit Weekly Newsletter for Period Ending July 29, 1967,” 

Atomics International, Canoga Park, California, November 30. 

Alexander, R. E., 1968a, “Restriction of Employees from Radiologically Controlled Areas,” Atomics 

International, Canoga Park, California, January 10. 

Alexander, R. E., 1968b, “Radiation Safety Unit Weekly Newsletter for Period Ending March 9, 1968,” 

Atomics International, Canoga Park, California, March 14. 

Atomics International, 1959, SRE Experimental Fuel Program Interim Report, NAA-SR-3456, Atomics 

International, Canoga Park, California, October 15.

Bales, T. J., 1969, “Notice of Award or Amendment to Agreement for Services,” internal letter, 

Atomics International, Canoga Park, California, April 16.

Barnes, J. G., 1999, Internal Dosimetry Technical Basis Document and Procedure, RS-00007, The 

Boeing Company, Canoga Park, California, October 28. 

Baurmash, L.,1967, “Particle Size of UO2 Aerosol in Processing Room in Bldg. 1,” internal letter to 

A. N. Tschaeche, Atomics International, Canoga Park, California, November 27. 

Boeing (The Boeing Company), 2003, Radiological Operations and Clean-up at the De Soto Facility, 

Canoga Park, California, June 12. 

Edelmann, A., 1959, telegram to M. E. Piedimonte, Atomics International, Nuclear Science and 

Engineering Corp., Pittsburgh, Pennsylvania. 

Eggleston, R. R., 1983, Annual Review of Radiological Controls – 1980, Rockwell International, 

Energy Systems Group, Canoga Park, California, October 12.

Eggleston, R. R., 1984, Annual Review of Radiological Controls – 1981, Rockwell International, 

Energy Systems Group, Canoga Park, California, May 29.

Fisher, W. L., 1963, “Analysis of Routine Bioassay Specimens,” internal letter to J. C. Lang, North 

American Aviation, Inc., Atomics International Division, Canoga Park, California, May 28.

Garcia, R. R., 1963, “Modification of the Personnel Monitoring Program Roles,” Procedure No. PM-1, 

North American Aviation, Inc., Atomics International Division, Canoga Park, California, July 1.

Hart, R. S., 1979, Annual Review of Radiological Controls – 1975, Rockwell International, Atomics 

International Division, Canoga Park, California, July 2.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 33 of 64

Hart, R. S., 1980a, Annual Review of Radiological Controls – 1976, Rockwell International, Energy 

Systems Group, Canoga Park, California, April 28. 

Hart, R. S., 1980b, Annual Review of Radiological Controls – 1977, Rockwell International, Energy 

Systems Group, Canoga Park, California, May 22.

Hart, R. S., 1980c, Annual Review of Radiological Controls – 1978, Rockwell International, Energy 

Systems Group, Canoga Park, California, August 6.

Hart, R. S., 1980d, Annual Review of Radiological Controls – 1979, Rockwell International, Energy 

Systems Group, Canoga Park, California, September 15.

Helgeson, G. L., 1968, Report of In Vivo Counting for Atomics International, Helgeson Nuclear 

Services, Inc., Pleasanton, California, May 29.

Helgeson, G. L., 1978, Report of In Vivo Counting for Atomics International Division Rockwell 

International, Helgeson Nuclear Services, Inc., Pleasanton, California, May 22.

Helgeson, G. L., 1983, Report of In Vivo Counting for Rockwell International Energy Systems Group, 

Helgeson Nuclear Services, Inc., Pleasanton, California, June 22. 

ICRP (International Commission on Radiological Protection), 1979, Limits for Intakes of Radionuclides 

by Workers, Publication 30, Annals of the ICRP, Vol. 2 (3-4), Pergamon Press, Oxford, 

England. 

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Annals of the ICRP, Vol. 24 (1-3), Pergamon 

Press, Oxford, England. 

Kellehar, R. G., 1959, “Internal Exposure and Bioassay,” AI-Memo 4645, Atomics International, 

Canoga Park, California, November 20.

Kellehar, R. G., 1966a, “Bioassay Program General Information,” internal letter to R. E. Alexander, 

Atomics International, Canoga Park, California, July 19.

Kellehar, R. G., 1966b, “Bioassay Services Evaluation,” internal letter to R. R. Kinsman, Atomics 

International, Canoga Park, California, December 12.

Kellehar, R. G., 1967, “Addition to Purchase Order #7110010,” internal letter to R. F. Kinsman, 

Atomics International, Canoga Park, California, June 7.

Lang, J. C., undated, Radiological Safety at Atomics International, AI-7850, Atomics International, 

Canoga Park, California. 

Lang, J. C., 1960, Radiological Safety at Atomics International, AI-Memo 5468, Atomics International, 

Canoga Park, California, July 13.

Lee, N. D., 1963, “Quotation on Urinalysis Services,” letter to Atomics International, BioScience 

Laboratories, Los Angeles, California, July 30.

Liddy, P., and P. Rutherford, 2001, Radiation Survey of the Downey Facility, RS-00019, The Boeing 

Company, Canoga Park, California, May 4. 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 34 of 64

Mason, M. G., and R. H. Burr, 1958, Recommended Bio Assay Procedures for Uranium, Atomics 

International, Canoga Park, California, October.

Moore, J. D., 1979, Revaluation of ATR Fuel Lung Burden Equivalent Ux Mass, Rockwell 

International, Rocketdyne Division, Canoga Park, California, October 19.

Moore, J. D., 1980a, Revaluation of ATR Fuel Lung Burden Equivalent Ux Mass, Rockwell 

International, Rocketdyne Division, Canoga Park, California, January 29.

Moore, J. D., 1980b, Revaluation of ATR Fuel Lung Burden Equivalent Ux Mass, Rockwell 

International, Rocketdyne Division, Canoga Park, California, May 12.

Moore, J. D., 1980c, Revaluation of ATR Fuel Lung Burden Equivalent Ux Mass, Rockwell 

International, Rocketdyne Division, Canoga Park, California, December 3.

Moore, J. D., 1981a, Revaluation of ATR Fuel Lung Burden Equivalent Ux Mass, Rockwell 

International, Rocketdyne Division, Canoga Park, California, July 28.

Moore, J. D., 1981b, Revaluation of ATR Fuel Lung Burden Equivalent Ux Mass, Rockwell 

International, Rocketdyne Division, Canoga Park, California, November 2.

Moore, J. D., 1982, Revaluation of ATR Fuel Lung Burden Equivalent Ux Mass, Rockwell 

International, Rocketdyne Division, Canoga Park, California, March 2.

Mountford, L. A. and H. A. Morewitz, 1965, “The Advanced Epithermal Thorium Reactor (AETR) 

Critical Experiments,” Nuclear Science and Engineering, Volume 21, pp. 421-428, American 

Nuclear Society, LaGrange Park, Illinois.

NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guideline, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio. 

NSEC (Nuclear Science and Engineering Corporation), 1957, Urinalysis Service Brochure, September 

1957, Pittsburgh, Pennsylvania. 

O’Brien, R. E., 1959, “Quote for Urinalysis for Uranium by Fluorometric Determination,” letter to 

Atomics International, Controls for Radiation, Inc., Cambridge, Massachusetts. 

ORAU (Oak Ridge Associated Universities), 2004, Technical Basis Document for the Hanford Site – 

Occupational Internal Dose, ORAUT-TKBS-0006-5, Rev. 01, Oak Ridge, Tennessee,

November 24. 

ORAU (Oak Ridge Associated Universities), 2005, Technical Basis Document for the Energy 

Technology Center – Site Description, ORAUT-TKBS-0038-2, Oak Ridge, Tennessee.

Remley, M. E., 1967, “Reportable Inhalation Exposures in Fuel Fabrication Facilities,” internal letter, 

Atomics International, Canoga Park, California, September 8.

Rutherford, P., 2002, Historical Radiological Activities at Building 038 (Vanowen Building), The Boeing 

Company, Canoga Park, California, January 15. 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 35 of 64

Sapere and Boeing (Sapere Consulting, Inc., and The Boeing Company), 2005, Historical Site 

Assessment of Area IV Santa Susana Field Laboratory, Ventura County, California, Volume 1 

– Methodology, Canoga Park, California. 

Saxe, D., 1967a, “Report of Inhalation Exposures at Atomics International, JAERI Program,” letter to 

Lawrence D. Low, Atomics International, Canoga Park, California, November 13.

Saxe, D. 1967b, “Report of Inhalation Exposures at Atomics International,” letter to Lawrence D. Low, 

Atomics International, Canoga Park, California, August 15.

Saxe, D. 1967c, Additional Information Regarding Report of Inhalation Exposures at Atomics 

International, Atomics International, Canoga Park, California, November 2.

Shepard, J. L., 1959, “Quotation on Urinalyses,” letter to Atomics International, U. S. Nuclear, 

Burbank, California, May 19.

Shleien, B., L. A. Slaback, and B. K. Birky, editors, 1998, Handbook of Health Physics and 

Radiological Health, Third Edition, Williams and Wilkins, Baltimore, Maryland. 

Spielman, K. K., 1968, “Notice of Award or Amendment to Agreement for Services No. N711-0010,” 

internal letter those listed, Atomics International, Canoga Park, California, September 25.

Staszesky, F. M., 1970, Radiation Engineering Analysis (To establish standards for bioassay 

sampling for radiation workers), Atomics International, Canoga Park, California, August 12.

Staszesky, F. M., 1971, “Request for Bioassay Services Quotations,” internal letter, Atomics 

International, Canoga Park, California, March 1971.

Tracerlab (Tracerlab, Inc.), 1959, Thorium Analysis of Urine Samples, Reactor Monitoring Center, 

Richmond, California, February 13.

Tschaeche, A. N., 1968a, Updated Whole Body Counting Data for Persons Exposed to Enriched 

Uranium, Atomics International, Canoga Park, California, August 15.

Tschaeche, A. N., 1968b, Errors in UCLA Whole Body Counting Data, Atomics International, Canoga 

Park, California, March 19.

Tschaeche, A. N., 1968c, Recalculation of Enriched Uranium Body Burdens Based on New Data for 

Corrections, Atomics International, Canoga Park, California, February 9.

Tuttle, R. J., 1985, Annual Review of Radiological Controls – 1982, Rockwell International, Energy 

Systems Group, Canoga Park, California, January 29.

Tuttle, R. J., 1986a, Annual Review of Radiological Controls – 1983, Rockwell International, 

Rocketdyne Division, Canoga Park, California, February 12.

Tuttle, R. J., 1986b, Annual Review of Radiological Controls – 1984, Rockwell International, 

Rocketdyne Division, Canoga Park, California, March 18.

Tuttle, R. J., 1986c, Annual Review of Radiological Controls – 1985, Rockwell International, 

Rocketdyne Division, Canoga Park, California, July 11.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 36 of 64

Tuttle, R. J., 1988a, Annual Review of Radiological Controls – 1986, Rockwell International, 

Rocketdyne Division, Canoga Park, California, July 5.

Tuttle, R. J., 1988b, Annual Review of Radiological Controls – 1987, Rockwell International, 

Rocketdyne Division, Canoga Park, California, July 28.

Tuttle, R. J., 1989, Annual Review of Radiological Controls – 1988, Rockwell International, 

Rocketdyne Division, Canoga Park, California, May 12. 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 37 of 64

GLOSSARY

Atomic Energy Commission (AEC)

An agency established by the U.S. Government for oversight of nuclear weapons and power 

production; a predecessor to the U.S. Department of Energy.

cladding

The outer layer of material encasing a reactor fuel element (e.g., aluminum or zirconium). 

Cladding promotes the transfer of heat from the fuel to the coolant and contains fission 

products and activation products that result from the fissioning of the fuel.

core

That part of the reactor consisting of the fuel and some of the control elements for reactor 

operation.

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation.

fission

A nuclear transformation characterized by the splitting of a nucleus into at least two other 

nuclei and the release of a relatively large amount of energy.

fission product

Radionuclides resulting from fission.

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

in vitro

Literally, in glass; outside the living body and in an artificial environment; internal bioassay 

sampling, such as fecal samples or urine samples.

in vivo

Literally, in the living; in the living body of a plant or animal; bioassay sampling by whole-body 

counting. 

isotope

Nuclides having the same number of protons in their nuclei (same atomic number), but a 

differing number of neutrons (different mass number).

natural uranium

Uranium that has not been through an enrichment process.

radiation

Energy transferred through air or some other media in the form of particles or waves (see 

ionizing radiation).
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 38 of 64

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma or X-rays, or 

neutrons from unstable atoms.

radionuclide

A radioactive species of an atom characterized by the constitution of its nucleus specified by 

atomic number (the number of protons) and the mass number (equal to the number of protons 

plus neutrons).

type

The rate of absorption from lung to blood of inhaled radioactive materials and includes types F 

(fast), M (moderate), and S (slow).

zirconium

A metallic element highly resistant to corrosion and often used to make cladding for nuclear 

fuel. It is sometimes alloyed in small amounts in the fuel itself.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 39 of 64

ATTACHMENT A

SUMMARY OF URANIUM FUEL FABRICATION INTERNAL DOSIMETRY ISSUES

In the fall of 1965, AI began fabricating reactor fuel elements. AI fabricated fuel for the ATR and the 

ETR in 1966, 1967, and 1968 at the De Soto Facility. In Room 1110-62, Building 101 (known at the 

time as Building 001), briquettes of an alloy of 93%-enriched uranium and aluminum, known as 

uranium aluminide (UAlx), were formed in an electric arc melting furnace. These briquettes were 

crushed to form a powder, which was cold-pressed into compacts that became the cores of the 

fabricated fuel plates. The room where these activities took place was known as the “powder room.”

The work was performed at six work stations: weighing station 1, a melting station, a crushing and 

sieving station, weighing station 2, a compact forming station, and a deburring station. The 

operations were performed with containment thought to be appropriate for the type of operation and 

the hazard of the material. From one to four air samplers operated in the room near the work stations.

In 1967, after 15 months of operation, urine bioassay data indicated that the material was probably 

insoluble and, therefore, that the air activity was not being compared to the appropriate MPC. 

Although the uranium was more than 93% 235U by weight, 234U accounted for more than 96% of its 

activity. The insoluble MPC for 234U was a factor of 6 lower than the soluble MPC. This led to the 

conclusions that the regulatory standard (weekly average MPC) had been exceeded on a number of 

occasions and that equipment and procedures for controlling the airborne uranium were insufficient.

An internal investigation determined that the primary reason for the ineffective confinement of the 

uranium was leakage from the crushing glovebox seal, from the fume hood of weighing station 2, and 

from the arc furnace. Temporary measures were put in place until more permanent fixes could be 

implemented. These changes were put in place and the project was completed. Along with the 

engineering changes implemented, workers were required to wear full-face respirators and lapel air 

samplers (Saxe 1967b). At some point, operations were relocated to a “new powder room.”

In 83 weeks of operation, the MPC for insoluble 234U (1 × 10-10 μCi/cm3

) was exceeded in 33 weeks. 

Twenty-one personnel who had worked in the powder room were monitored by chest counting. Three 

of these personnel were determined to have exceeded the MPBB and received work restrictions. The 

results for the other workers ranged from no detectable intake to less than one MPBB.

On August 15, 1967, a report of regulatory violations was submitted to the AEC. The report included 

the corrective measures taken (Sax 1967b). Inquiries by the AEC led to a review of an earlier fuel 

fabrication project, the Japan Atomic Energy Research Institute (JAERI), which took place in Room 

1110-061. This operation involved uranium metal plates enriched to 20% by weight. The activity of 

the material was still approximately 90% 234U. This operation did not involve crushing, but deburring 

the edges of the plates occurred during the process. This apparently took place outside ventilated 

enclosures until January 1967. An investigation into the JAERI fuel operation led to the conclusion 

that the average weekly concentration for insoluble 234U had been exceeded six times during the 

operation in November and December 1966. The soluble standard (6 × 10-10 μCi/cm3

) would not have 

been exceeded. The insoluble standard was used because of the experience with UAlx and the lack 

of evidence that uranium metal should be treated differently. Using this standard, five people had 

been exposed to concentrations greater than the AEC MPC. None of these people were involved in 

the ATR or ETR project (Remley 1967). Urinary excretion was measured on all five workers. The 

data were consistent with short-term exposures with small lung depositions. However, one of the 

workers initially showed a lung burden 1.5 times the MPBB. As stated in the main text, lung counting 

was its infancy during this time. The estimate for this worker was later reduced to 0.5 MPBB, based 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 40 of 64

on the difference in enrichments between the JAERI material and the calibration source for the lung 

counter at UCLA.

Air samples collected prior to 1966 were reviewed, and no concentrations exceeded 1 × 10-10 μCi/cm3

(Saxe 1967c).

A particle size study consisting of two general area air samples from the powder room indicated that 

the particles were less than 1 micron count median diameter (CMD) (Alexander 1967c). The actual 

median count diameter and geometric standard deviation were not provided in the reference; 

therefore, the activity median aerodynamic diameter (AMAD) could not be determined. In the only 

other particle size reference recovered, Baurmash (1967) measured the particle size distribution for a 

UO2 grinding operation in the “processing room in Bldg. 1.” The CMD reported for this operation was 

0.195 micron with a geometric standard deviation of 1.66. The mass median diameter was calculated 

as 0.42 micron. Using a density of 10.97 g/cm3 for UO2, the AMAD for this aerosol is 1.39 microns. 

While this information is extremely limited, it points out that small particle sizes were at least possible 

during uranium fuel fabrication. Therefore, dose reconstructors should consider a 1-micron AMAD 

particle size for these operations in addition to the default assumption.

A study of the solubility of UAlx in simulated lung fluid was completed by AI in 1968. While the original 

study has not been recovered, the reference to it indicates that UAlx was about 10 times more 

insoluble than the most insoluble uranium compound (U3O8) measured by ORNL (Alexander 1968b). 

ICRP (1979) lists U3O8 as a class Y compound. The quality of the AI study is uncertain. ICRP (1994) 

lists the most insoluble compounds of uranium as type S. Therefore, dose reconstructors should 

consider UAlx results as type S.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 41 of 64

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 1 of 20

Figure B-1. Example of Request for Film Badge and Bioassay Services, 1963. 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 42 of 64

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 2 of 20

Figure B-2. Example Individual Personnel Keysort Card, 1966. (Both sides are shown. 

See key on following page.) 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 43 of 64

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 3 of 20

Key to the Individual Keysort Card, Figure B-2. The following numbers correspond to the handwritten 

numbers on the form. The cards were set up for a 4-yr period (Kellehar 1966a).

1. Frequency

2. Week of collection

3. Date of each specimen collection, wound, and nasal smears 

4. Indication of analysis results; degree of positive results in any year

5. Type of analyses required

6. Changes in work affecting analyses and/or specimen collection (reverse side)

7. Special bioassays (reverse)

The following is an edited version of the description of the form provided by ETEC:

This is the "master" record for internal dosimetry results. This lists the type of analysis and the "per 

day" results of those analyses.

Date is the date the sample was obtained from the individual.

Type lists the type of sample obtained. Method is a code to identify the procedure utilized. An 

incomplete record exists of the specific laboratory techniques associated with these methods. When 

results are listed as "< XXX," XXX is used as the reporting level rather than the value listed in the 

table.

Results are listed. In general, unless otherwise listed, they are in "dpm/day" (for radiological data) or 

"μg/day" (for fluorometric or other chemical analyses). Results were standardized to a 1,500-mL/day 

urine excretion rate or a 135-g/day fecal excretion rate. The results line up with the methods. 

Statistical errors were generally NOT provided, but can be obtained by manual search of the source 

data.

This (back) side of the Bioassay Card sometimes has additional bioassay data. No results are listed 

on this side of the card, but it can be useful for determining the work location of the individual and 

possibly the type of work being done at the time.

Special bioassays may also be listed on this side of the card. If listed on this side, it generally 

suggests that the bioassay was taken in response to a specific incident. However, absence of 

bioassay data on this side of the card should not be interpreted as an indication that the individual 

was never involved in an incident.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 44 of 64

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 4 of 20

Figure B-3. Example Bioassay Data Sheet, page 1 of 2 (used by onsite laboratory). 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 45 of 64

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 5 of 20

Bioassay Data Sheet, page 1 (front).

The following is an edited version of the description of the form provided by ETEC:

This sheet records bioassay results when a determination was done in the AI laboratory (it also 

appears to have been used to log in and document vendor results that were sent in tabular form). It 

was designed to be a multipurpose form for use with all types of analyses.

The general concept of the sheet was that all steps of a hand calculation could be performed by 

entering the appropriate data and performing the indicated mathematical operations. The calculations 

on this side of the page were preliminary; the sample results are listed on the second page of this 

form.

One will occasionally find results penciled in on the top right corner of the page.

The calculations are relatively straightforward. They are typical for a radiochemical/radiological 

laboratory.

Multiple count blocks were provided; it appears this was used to log in sequential counts to determine 

half-lives of the isotopes.

Information blocks are self-explanatory.

Note that only those blocks on the form pertinent to the type of measurement were filled in. One can 

use this to determine the type of sample being analyzed.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 46 of 64

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 6 of 20

Figure B-4. Example Bioassay Data Sheet, page 2 of 2 (used by onsite laboratory). 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 47 of 64

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 7 of 20

Bioassay Data Sheet, page 2 (back).

The following is an edited version of the description of the form provided by ETEC:

This sheet records bioassay results when a determination was done in the AI laboratory (it also 

appears to have been used to log in and document vendor results that were sent in tabular form). It 

was designed to be a multipurpose form to be used with all types of analyses.

This is the "back" side of the form.

The general concept of the sheet was that all steps of a hand calculation could be performed by 

entering the appropriate data and performing the indicated mathematical operations. The calculations 

on the front side of the page were preliminary; the sample results are listed on this second page of the 

form.

One will occasionally find results penciled in on the top right corner of the page.

The calculations are relatively straightforward. They are typical for a radiochemical/radiological 

laboratory.

Multiple count blocks were provided; it appears this was used to log in sequential counts to determine 

half-lives of the isotopes.

Information blocks are self-explanatory.

Note that only those blocks on the form pertinent to the type of measurement were filled in. One can 

use this to determine the type of sample being analyzed.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 48 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 8 of 20

Figure B-5. Example Bioassay Data Sheet, BioScience, Uranium Fluorometric. 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 49 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 9 of 20

The following is an edited version of the description of the form provided by ETEC:

Report for Uranium fluorometric results from BioScience Laboratories. These are the "record of entry" 

reports from the vendor laboratory that processed urine and fecal bioassay results.

These records should be correlated to the bioassay summary record (Individual Keysort Card [Figure 

B-2]). The individual's name is usually written in pencil in the upper left margin.

The Source Number probably refers to an identification code for Atomics International.

The Report Date is the date the results were available.

The number in the Patient block is the Health Physics ID Number used to identify the person.

The meaning of the No. Spec. Type entry is not known at this time. It is sometimes blank.

Entry Date refers to the date that the sample was received for processing. In general it is one to two 

days after the date the sample was obtained by Rocketdyne from the individual. This date is obtained 

from the Individual Keysort Card. 

Results are generally reported in micrograms U (μg) per 100 milliters of samples (ug/100 ml). On 

positive samples, one will usually find a value pro-rated into a "per day" rate using 1500 ml/day as a 

urine excretion rate, and 135 g/day for a fecal excretion rate.

By dividing the uranium radiometric values by the uranium fluorometric values, one can derive a 

specific activity for the sample. This specific activity can then be used to estimate the general 

enrichment level of the uranium the individual was exposed to. For simplicity, for low intakes, the

ETEC practice was to model the intake as pure U-234 (which results in overestimation of "true" dose).
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 50 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 10 of 20

Figure B-6. Example Bioassay Data Sheet, BioScience, Uranium Radiometric.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 51 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 11 of 20

The following is an edited version of the description of the form provided by ETEC:

Report for uranium radiometric results from BioSicence Laboratories. These are the "record of entry" 

reports from the vendor laboratory that processed urine and fecal bioassay results.

These records should be correlated to the bioassay summary record (Individual Keysort Card [Figure 

B-2]). Two versions were used (see Figure B-7 for a side by side comparison).

The individual's name is usually written in pencil in the upper left margin. The Source No. generally 

lists a sample number. The Report Date is the date the results were available.

The number in the Patient block is the sample number to identify the person. 

The meaning of the No. Spec. Type entry is not known at this time.

Entry Date refers to the date that the sample was received for processing. In general it is one to two 

days after the date the sample was obtained by Rocketdyne from the individual. This date is obtained 

from the Individual Keysort Card. 

Background refers to the background on the alpha counter (generally counts/hour).

Counter efficiency is the efficiency of the counter used.

Recovery is the efficiency of the chemical recovery during the chemical preparation of the sample.

Urine Volume (ml) is the amount of sample actually analyzed.

Gross is the counter results UNCORRECTED for background.

Net - 0.95 error is the NET count rate (gross - bkgd) with a 95% statistical error indicated. If the error 

exceeds the net count, it is considered a non-detect. If the net exceeds the error, then it is considered 

a detect. The basis for this calculation of error (i.e., count times, normal or Poisson, etc.) is not 

available.

Disintegrations/min is the net activity determined for the sample.

In some cases, one will find a value of "XX dpm/day" written in. This is the sample result pro-rated to 

a daily excretion rate using a 1500 ml assumption for daily urine volume and a 135 g/day rate for 

feces.

Disintegrations are calculated by (Net CPM) / (Counter Efficiency/100) * (Recovery/100) [Assuming 

Efficiencies are written in percentage; fractional "percentages" are more likely true efficiency fractions 

and would not need conversion from percentage.]

By dividing the uranium radiometric values by the uranium fluorometric values, one can derive a 

specific activity for the sample. This specific activity can then be used to estimate the general 

enrichment level of the uranium the individual was exposed to. For simplicity, for low intakes, ETEC

practice was to model the intake as pure U-234 (which results in overestimation of "true" dose).
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 52 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 12 of 20

Figure B-7. Example Bioassay Data Sheets, BioScience, Uranium Radiometric, Showing Bottle Tags.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 53 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 13 of 20

The following is an edited version of the description of the form provided by ETEC:

These are two more examples of the BioScience UR records described elsewhere.

Note that these records include bottle tags listing the exact times of sample collection.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 54 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS 

Page 14 of 20

Figure B-8. Example Bioassay Data Sheet, UST, Uranium Fluorometric.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 55 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 15 of 20

The following is an edited version of the description of the form provided by ETEC:

This is the vendor record for results for uranium fluorometric sampling from U. S. Testing.

The calculations are self-explanatory.

See previous discussion at BioScience UR record description for additional data.

Specimen Number contains the individual’s Health Physics Identification Number.

A sample number was generally written in the center box.

Date Received is the date the sample arrived at the laboratory. This was generally one to two days 

after the collection date. The actual date the sample was obtained is on the Individual Keysort Card. 

These sheets should be correlated to that card.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 56 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 16 of 20

Figure B-9. Example Bioassay Data Sheet, UST, Uranium Radiometric.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 57 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 17 of 20

The following is an edited version of the description of the form provided by ETEC:

This is the vendor record for results for uranium radiometric sampling from U. S. Testing.

The calculations are self-explanatory.

See previous discussion at BioScience UR record description for additional data.

Specimen Number contains the individual’s Health Physics Identification Number.

A sample number was generally written in the center box.

Date Received is the date the sample arrived at the laboratory. This was generally one to two days 

after the collection date. The actual date the sample was obtained is on the Individual Keysort Card. 

These sheets should be correlated to that card.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 58 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 18 of 20

Figure B-10, Example Bioassay Data Sheet, UST, Mixed Fission Products 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 59 of 64

ATTACHMENT B 

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 19 of 20

The following is an edited version of the description of the form provided by ETEC:

This is the vendor record for results for mixed fission product sampling from U. S. Testing.

The calculations are self-explanatory.

Specimen Number contains the individual’s Health Physics Identification Number.

A sample number was generally written in the center box.

Date Received is the date the sample arrived at the laboratory. This was generally one to two days 

after the collection date. The actual date the sample was obtained is on the Individual Keysort Card. 

These sheets should be correlated to that card.

These measurements were generally defaulted to Cs-137. In addition, one should be aware that Cs137 was generally assumed to be accompanied by Sr-90 in a 50%/50% mix. Analyses of MFP results 

should therefore account both for Cs-137 exposures and Sr-90 exposures as being co-incident.

Note: Based upon laboratory analyses directly comparing Cs-137 and Sr-90 values, the mixture ratio 

for Cs-137 and Sr-90 during decontamination and decommissioning operations after 1991 were in a 

ratio of approximately 85% Cs-137 to 15% Sr-90. In 1999, the ratio published in the site TBD was 

90% Cs-137 to 10% Sr-90 for SSFL locations other than Bldg. 4059.
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 60 of 64

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 20 of 20

Figure B-11. Example Wound Monitoring Report. 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 61 of 64

ATTACHMENT C

ADJUSTMENT TO HELGESON NUCLEAR SERVICES U-235 LUNG COUNTS, 1981–1983

Measured

(μg U-235)

Actual

(μg U-235)

32.5 0

40 8.6

50 20.2

60 31.7

70 43.2

80 54.8

90 66.3

100 77.8

110 89.4

120 101

130 112

140 124

150 135

160 147

170 159

180 170

190 182

200 193

210 205

220 216

230 228

240 239

245 245
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 62 of 64

ATTACHMENT D

ISOTOPIC COMPOSITION OF ATR-ETR FUEL USED IN THE POWDER ROOM, 1966–1967

Date received

Isotopic composition (weight percent)

U-234 U-235 U-236 U-238

6/27/1966 0.579 93.2 0.304 5.917

6/27/1966 0.571 93.2 0.301 5.928

10/12/1966 0.516 93.15 0.310 6.010

10/13/1966 0.656 93.16 0.295 5.900

11/11/1966 0.563 93.14 0.289 6.000

12/8/1966 0.591 93.11 0.326 5.970

12/8/1966 0.627 93.13 0.286 5.960

1/20/1967 0.563 93.09 0.305 6.020

2/6/1967 0.570 93.15 0.320 5.956

2/8/1967 0.580 93.17 0.290 5.961

2/9/1967 0.550 93.14 0.310 6.001

2/10/1967 0.570 93.19 0.270 5.966

2/11/1963 0.630 93.14 0.300 5.929

3/11/1967 0.660 93.17 0.300 5.872

3/16/1967 0.610 93.13 0.350 5.912

4/16/1967 0.500 93.13 0.340 6.030

6/19/1967 0.557 93.23 0.324 5.860

6/22/1967 0.582 93.16 0.304 5.950

6/23/1967 0.600 93.14 0.290 5.972

6/23/1967 0.580 93.12 0.294 6.010

6/23/1967 0.570 93.14 0.301 6.000

Average 0.582 93.152 0.305 5.958

Standard deviation 0.039 0.033 0.019 0.048

Specific activity (μCi/g) 6,263.14 2.16 64.70 0.340

Fractional specific activity 

(μCi/g-U) 36.46 2.01 0.20 0.020

Fractional specific activity 

(Bq/g-U) 1.35E+06 7.44E+04 7.31E+03 7.50E+02

Data from Tschaeche (1968c)
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 63 of 64

ATTACHMENT D (CONTINUED)

ISOTOPIC COMPOSITION OF URANIUM USED FOR ATR FUEL FABRICATION, 1979-1982

Date

Isotopic composition (weight percent)

U-234 U-235 U-236 U-238

10/18/1979 1.006 93.145 0.525 5.234

1/29/1980 1.145 93.142 0.498 5.348

5/12/1980 1.007 93.141 0.472 5.379

12/3/1980 1.007 93.142 0.476 5.374

7/28/1981 1.004 93.14 0.455 5.648

11/2/1981 1.002 93.136 0.455 5.407

3/2/1982 0.9977 93.141 0.448 5.414

Average 1.024 93.141 0.476 5.401

Standard deviation 0.0534 0.0027 0.0276 0.1246

Specific activity 

(μCi/g) 6,263.14 2.16 64.70 0.340

Fractional specific 

activity (μCi/g-U) 64.141 2.012 0.308 0.018

Fractional specific 

activity (Bq/g-U) 2.37E+06 7.44E+04 1.14E+04 6.79E+02

Data from Moore (1979, 1980a,b,c, 1981a,b, 1982). 
Document No. ORAUT-TKBS 0038-5 Revision No. 00 Effective Date: 02/22/2006 Page 64 of 64

ATTACHMENT E

DETECTION LIMITS FOR GAMMA SCANS OF BIOASSAY SAMPLES, 1967–1975

Isotope(s)

Detection limit

(dpm/sample)

Na-24 80

Co-60 75

Sc-46 60

Cs-134 175

Ba-La-140 250

Co-58 50

K-40 500

Ru-l06 500

Zn-65 100

Cu-64 300

Mn-56 70

Mn-54 60

Ga-72 140

Zr-Nb-95 20

Cs-137 50

As-76 120

Ru-103 50

Zn-69m 50

I-131 60

Cr-51 400

Np-239 80

Ce-141 60

Ce-Pr-144 500