Page 1 of 78

DOE Review Release 09/26/2023

Site Profile for Ames Laboratory

ORAUT-TKBS-0055 Rev. 04

Effective Date: 09/18/2023

Supersedes: Revision 03

Subject Expert(s): Jackson R. Ellis

Document Owner 

Approval: Signature on File

Jackson R. Ellis, Document Owner

Approval Date: 09/15/2023

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 09/15/2023

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 09/15/2023

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 09/15/2023

Approval: Signature on File

Timothy D. Taulbee, Associate Director for Science

Approval Date: 09/18/2023

FOR DOCUMENTS MARKED AS A TOTAL REWRITE OR REVISION, 

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

New Total Rewrite Revision
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 2 of 78

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

06/22/2007 00 Approved new Site Profile for Ames Laboratory. Incorporates formal 

internal and NIOSH review comments. Adds Glossary and 

Attributions and Annotations section. There is no change to the 

assigned dose and no PER is required. Training required: As 

determined by the Task Manager. Initiated by Jerome B. Martin.

08/20/2008 00 PC-1 Approved page change revision made to include SEC-00075 on 

pages 10 and 11 in Section 1.3 where Sections 1.3.1, 1.3.2 and 

Table 1-2 were added. NIOSH required language was revised on 

pages 9 and 10 in Section 1.1. Added reference on page 74.

Incorporates formal internal review comments. As a result of formal 

internal review, editorial changes were made on the following pages:

8 in Acronym and Abbreviations Section; 11 (Table 1-2), 25 (Table

4-1), 30 (Table 4-2), 32 (Table 4-6), 34 (Table 5-1), 36 (Table 5-2 

and 5-3), 48 (Table 5-6), 51 (Table 6-1), in Sections 1.3, 4.2, 4.4, 4.7, 

5.0, 5.1, 5.4, and 6.2. SEC changes occurred on pages 11 (Table 1-

1), 32, 33, 37-39 (Table 5-4), 40, 41, 46 (Table 5-5), 49 (Table 5-7), 

50, 57, 60, 61 (Table 6-4), 63 (Figure 6-2) in Sections 1.3, 4.7, 5.0, 

5.1, 5.2, 5.3, 5.5, 6.1, and 6.3. Incorporates formal NIOSH review 

comments on page 2 (Publication Record), 30 (Table 4-2), 41, 48 

(Table 5-6), and 74 in Sections 4.4, 5.2, 5.4, and Reference Section.

References were updated on pages 10, 11, 19, 29, 30, 32, 33, 37, 

40, 41, 46, 47, 50-52, 57-62, 64, 65, 70, 74-76 in Sections 1.3, 2.3, 

4.4, 4.7, 5.0, 5.1, 5.2, 5.4, 6.1, 6.2, 6.3, 6.4, 7.0, and Reference 

Section. No sections were deleted. Training required: As determined 

by the Task Manager. Initiated by Karen S. Kent.

12/18/2009 01 Approved revision to add lateral chest doses for all periods, skin 

doses for various parts of the body, and an explanation of 

terminology. Sections 3.0 and 3.1 were modified as a result of these 

changes. Tables 3-1 through 3-6 were modified and Tables 3-7 and 

3-8 were added. The definition of occupational medical dose was 

revised, and three reference titles were changed. Wording in Section 

6.2.1 about external uranium dose was modified. Revision made to 

add unmonitored doses before 1953. Section 6.2.1 reworded to 

remove OTIB-0004 discussion and add new methodology. Section 

6.3.1 was modified to include two eras of unmonitored doses. Figure

6-2 was altered to reflect the new unmonitored dose methodology for 

doses before 1953. Table 5-1 was separated into three tables.

Changes were made to Table 5-7 (now 5-8) to clarify and correct 

intake values. Wording was added to Section 1.3 to caution the Dose 

Reconstructor on when to label the DR as a partial DR. Incorporates 

formal internal and NIOSH review comments. Constitutes a total 

rewrite or the document. Training required: As determined by the 

Objective Manager. Initiated by Jackson R. Ellis.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 3 of 78

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

01/14/2011 02 Revision initiated to include SEC-00166. Section 1.3.3 and Table 1-3 

added. Section 2.1.6 updated to note the end of hot cell operations in 

the Research Building. Paragraph added in Section 5.0 to discuss 

SEC-00166. Section 5.1.1.3 updated with internal intake information 

from SEC-00166. Table 5-4 and 5-8 updated with updated intake 

rates. Section 5.4.1 updated to remove statement that intakes from 

the Research Building hot cell would be negligible. Removed A&A 

point. Added SEC-00166 to the reference section as NIOSH 2010.

Incorporates formal internal and NIOSH review comments.

Constitutes a total rewrite of the document. Training required: As 

determined by the Objective Manager. Initiated by Jackson R. Ellis. 

01/03/2012 03 Revision initiated to include SEC-00185. Section 1.3.4 was added.

Section 1.3 edited to update partial dose reconstruction information 

and to introduce new SEC. Sections 1.3.1 and 1.3.2 were edited to 

show inclusion of all workers in those SECs. Expansion of class was 

inserted in sections of the document as necessary. Section 5.4.1 

edited to state environmental doses should be applied to certain 

Research Building workers. Battelle 2006a was updated to Battelle 

2011 and Battelle 2006b was cancelled and replaced with other 

references. Changes were made to several sections and tables in 

Sections 5.1 and 6.3. Section 6.3.1.2 was updated to reduce missed 

dose component of beta doses in Table 6-7. Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training required: As determined by the Objective 

Manager. Initiated by Jackson R. Ellis.

09/18/2023 04 Revision initiated to incorporate SEC-00245. Section 1.3 edited to 

update partial dose reconstruction information and to introduce SEC00245. Section 4.4 deleted per SEC. Section 5.1.1.4 deleted. Section 

5.2 edited to remove thorium intakes prior to 1990 per SEC. Section 

5.3.1 updated to present improved example tritium bioassay records. 

Section 5.3.2 edited to start tritium exposures in 1990 per SEC. 

Section 5.3.3 deleted to remove tritium unmonitored exposures per 

SEC. Sections 5.4.1 and 5.4.2 revised to remove fission product 

intakes per SEC. Summary table edited to incorporate revisions and 

deletions discussed above. Section 6.3.1 was edited to remove 

unmonitored doses prior to 1955 per SEC. Section 6.4 edited to 

extend exposure to organ dose factor to 1985, in line with NVLAP 

certification. Section 7 was updated to assign A&A citations to 

current project personnel. Brought into compliance with 

ORAUT-PROC-0031 as specified during discussions with NIOSH. 

Section 4.5 updated to include uranium intakes based on residual 

contamination and subsequent reduction per Battelle-TBD-6000 and 

ORAUT-OTIB-0070. Thorium ingestion intakes added based on 

NUREG/CR 6755. Incorporates additional formal NIOSH review 

comments. Sections 4.1 and 5.1 edited to incorporate environmental 

uranium intakes as documented in NIOSH Response to SC&A 

Review of the Ames Site Profile Document Regarding Uranium 

Internal Exposures white paper. Incorporates formal internal and 

NIOSH review comments. Constitutes a total rewrite of a focused 

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

Initiated by John M. Byrne and authored by Jackson R. Ellis.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 4 of 78

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 7

1.0 Introduction .............................................................................................................................. 9

1.1 Purpose...................................................................................................................... 10

1.2 Scope......................................................................................................................... 10

1.3 Special Exposure Cohort............................................................................................ 10

2.0 Site Description...................................................................................................................... 12

2.1 Facilities ..................................................................................................................... 12

2.1.1 Physical Chemistry Annex 1......................................................................... 14

2.1.2 Chemistry Building (Gilman Hall).................................................................. 14

2.1.3 Physics Building (Physics Hall) .................................................................... 15

2.1.4 Physical Chemistry Annex 2......................................................................... 15

2.1.5 Metallurgy Building (Wilhelm Hall)................................................................ 15

2.1.6 Research Building (Spedding Hall)............................................................... 15

2.1.7 Office and Laboratory Building ..................................................................... 16

2.1.8 Synchrotron Building (Spangler Geotechnical Laboratory) ........................... 16

2.1.9 Metals Development Building....................................................................... 16

2.1.10 Ames Laboratory Research Reactor (Applied Science Complex)................. 16

2.2 Operations.................................................................................................................. 17

2.2.1 Uranium Metal Operations ........................................................................... 17

2.2.2 Thorium Metal Operations............................................................................ 18

2.2.3 Other Operations.......................................................................................... 19

2.3 Source Terms and Personnel Monitoring.................................................................... 19

3.0 Occupational Medical Dose.................................................................................................... 20

4.0 Occupational Environmental Dose ......................................................................................... 20

4.1 Uranium and Thorium Production, 1942 to 1953 ........................................................ 20

4.2 Synchrotron Operations, 1949 to 1971 ....................................................................... 21

4.3 Ames Laboratory Research Reactor Operations, 1965 to 1977.................................. 23

4.4 Significant Environmental Event ................................................................................. 25

4.5 Environmental Contamination, 1953 to Present.......................................................... 26

4.5.1 Residual Radioactivity Period Uranium Exposure......................................... 26

4.5.2 Other Nuclides ............................................................................................. 28

4.6 Summary of Environmental External Doses and Intakes ............................................ 29

5.0 Occupational Internal Dose.................................................................................................... 30

5.1 Uranium Exposure...................................................................................................... 30

5.1.1 Analysis of Uranium Bioassay Data ............................................................. 30

5.1.2 Worker Exposure Categories ....................................................................... 33

5.1.3 Intake Calculations....................................................................................... 34

5.1.4 Assignment of Internal Exposures................................................................ 34

5.2 Thorium Exposure from Residual Contamination........................................................ 35

5.3 Tritium Exposure ........................................................................................................ 38

5.3.1 Tritium Exposure for Monitored Workers, 1965 to 1981................................ 39

5.3.2 Tritium Dose to Workers, 1990 to Present.................................................... 42

5.4 Fission Product Intakes .............................................................................................. 42

5.4.1 Fission Product Intakes from Early Fuel Research....................................... 42
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5.4.2 Fission Product Intakes from Research Reactor Operations and 

Decontamination and Decommissioning ...................................................... 43

5.5 Summary of Internal Dose Recommendations............................................................ 43

6.0 Occupational External Dose................................................................................................... 45

6.1 Introduction................................................................................................................. 45

6.2 External Radiation Dosimeters and Records .............................................................. 45

6.2.1 Historical Administrative Practices ............................................................... 46

6.2.2 Dosimetry Technology ................................................................................. 48

6.2.2.1 Beta/Photon Dosimeters............................................................ 49

6.2.2.2 Neutron Dosimeters .................................................................. 49

6.2.3 Dosimeter Calibration Procedures................................................................ 49

6.2.3.1 Beta/Photon Dosimeters............................................................ 50

6.2.3.2 Neutron Dosimeters .................................................................. 50

6.2.4 Workplace Radiation Fields.......................................................................... 50

6.2.4.1 Beta Radiation........................................................................... 51

6.2.4.2 Photon Radiation....................................................................... 51

6.2.4.3 Neutron Radiation ..................................................................... 51

6.2.5 Dosimeter Response to Radiation Fields ..................................................... 52

6.2.5.1 Beta/Photon Film Dosimeter Response..................................... 52

6.2.5.2 Beta/Photon Thermoluminescent Dosimeter Response............. 52

6.2.5.3 Neutron Dosimeter Response ................................................... 52

6.2.5.4 Neutron Dose Weighting Factor ................................................ 53

6.3 Recommendations for Worker External Dose Reconstruction..................................... 53

6.3.1 Unmonitored External Dose ......................................................................... 54

6.3.1.1 Co-Exposure Doses, 1955 to Present ....................................... 54

6.3.2 Missed External Dose for Monitored Workers .............................................. 56

6.3.3 Uncertainty in Photon Doses........................................................................ 57

6.3.4 Skin Dose .................................................................................................... 58

6.3.5 Extremity Dose............................................................................................. 58

6.3.6 Radiation Dose Fraction............................................................................... 58

6.4 Organ Dose................................................................................................................ 58

7.0 Attributions and Annotations .................................................................................................. 58

References ........................................................................................................................................ 62

Glossary ............................................................................................................................................ 73

ATTACHMENT A EXTENDED DESCRIPTION OF FIGURE 6-2 FLOWCHART.......................... 78
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 6 of 78

LIST OF TABLES

TABLE TITLE PAGE

1-1 Feasibility findings for SEC classes, August 13, 1942, through December 31, 1989 .............. 12

2-1 Timeline of activities, 1942 to 1954 ........................................................................................ 18

4-1 Survey of fence line gamma dose rates around the synchrotron facility with the beam 

directed west..........................................................................................................................22

4-2 Environmental dose measurements at the four facilities surveyed for thorium release 

effects ....................................................................................................................................26

4-3 Uranium inhalation intake rates from residual contamination.................................................. 27

4-4 Uranium associated ingestion intake rates from residual contamination................................. 28

4-5 Gamma analysis of soil samples............................................................................................ 29

4-6 Summary of 24 soil samples for beta and alpha activity ......................................................... 29

4-7 Environmental external doses ................................................................................................ 29

4-8 Summary of environmental intakes ........................................................................................ 30

5-1 Ames uranium worker bioassay data...................................................................................... 30

5-2 High exposure potential inhalation intake rates for 234U .......................................................... 35

5-3 Medium exposure potential inhalation intake rates for 234U ..................................................... 35

5-4 Low exposure potential inhalation intake rates for 234U ........................................................... 35

5-5 Daily intakes of 232Th at Wilhelm Hall, 1990 to 2010............................................................... 37

5-6 Dose calculation and IREP input parameters for default intakes ............................................ 44

6-1 Dosimeter type, period of use, exchange frequency, MRD, and MDL .................................... 46

6-2 Neutron dose energies, percentages, and associated ICRP [1991] correction factors............ 53

6-3 Assigned dose for unmonitored workers ................................................................................ 55

6-4 Potential missed dose for monitored workers......................................................................... 56

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 ISU campus map showing locations of laboratory buildings including former sites of 

Physical Chemistry Annex 1 and Annex 2.............................................................................. 13

2-2 ISU campus map showing locations of reactor facilities and Synchrotron Building................. 14

4-1 Wind rose for Des Moines, Iowa, 1950 to 1955...................................................................... 21

4-2 ALRR site...............................................................................................................................24

4-3 Environmental monitoring near the ALRR .............................................................................. 25

5-1 Bioassay record example 1.................................................................................................... 41

5-2 Bioassay record example 2.................................................................................................... 42

6-1 Comparison of Hp(10) for photons with energy responses for sensitive DuPont 502 

emulsion in two-element film badge and sensitive DuPont 555 emulsion in multielement 

film badge ..............................................................................................................................49

6-2 Flowchart for monitored and unmonitored workers................................................................. 57
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 7 of 78

ACRONYMS AND ABBREVIATIONS

A&A attribution and annotation

AEC U.S. Atomic Energy Commission

ALRR Ames Laboratory Research Reactor

AWE Atomic Weapons Employer

BNL Brookhaven National Laboratory

BRH Bureau of Radiological Health

C.F.R. Code of Federal Regulations

Ci curie

cm centimeter

d day

D&D decontamination and decommissioning

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DR dose reconstruction

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

F fast (absorption type)

ft foot

g gram

GM geometric mean

GSD geometric standard deviation

Hp(10) personal dose equivalent at 10 millimeters depth in tissue

HPS Health Physics Services

hr hour

ICN ICN Dosimetry Service

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

ISU Iowa State University

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

kV kilovolt

L liter

m meter

M moderate (absorption type)

MDA minimum detectable amount

MDL minimum detection level

MeV megaelectron-volt, 1 million electron-volts

mg milligram
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mi mile

mL milliliter

mo month

mR milliroentgen

mrad millirad

MRD minimum recordable dose

mrem millirem

mrep millirep

MTR Materials Test Reactor

MW megawatt

n neutron

NCA Nucleonic Corporation of America

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

OPOS one person–one statistic

ORAU Oak Ridge Associated Universities

ORAUT ORAU Team

pCi picocurie

PER program evaluation report

R&D research and development

s second

S slow (absorption type)

SEC Special Exposure Cohort

SLAC Stanford Linear Accelerator Center

SRDB Ref ID Site Research Database Reference Identification (number)

t ton

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

yr year

β beta

γ gamma

μCi microcurie

μg microgram

§ section or sections
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 9 of 78

1.0 INTRODUCTION

Technical basis documents (TBDs) and site profile documents are not official determinations made by 

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

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

dose reconstructions at particular U.S. Department of Energy (DOE) or Atomic Weapons Employer 

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

relevant information is obtained about the affected DOE or AWE facility(ies), such as changing 

scientific understanding of operations, processes, or procedures involving radioactive materials. 

These documents may be used to assist NIOSH staff in the evaluation of Special Exposure Cohort 

(SEC) petitions and the completion of individual dose reconstructions under Part B of the Energy 

Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA).

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

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

“AWE facility” or a “DOE facility.” The term “AWE facility” is defined in EEOICPA to mean “a facility, 

owned by an atomic weapons employer, that is or was used to process or produce, for use by the 

United States, material that emitted radiation and was used in the production of an atomic weapon, 

excluding uranium mining or milling.” 42 United States Code (U.S.C.) § 7384I(5). On the other hand, a 

DOE facility is defined as “any building, structure, or premise, including the grounds upon which such 

building, structure, or premise is located—(A) in which operations are, or have been, conducted by, or 

on behalf of, the [DOE] (except for buildings, structures, premises, grounds, or operations ... 

pertaining to the Naval Nuclear Propulsion Program); and (B) with regard to which the [DOE] has or 

had—(i) a proprietary interest; or (ii) entered into a contract with an entity to provide management and 

operation, management and integration, environmental remediation services, construction, or 

maintenance services.” 42 U.S.C. § 7384I(12). The DOE determines whether a site meets the 

statutory definition of an AWE facility and the U.S. Department of Labor (DOL) determines if a site is a 

DOE facility and, if it is, designates it as such.

Under EEOICPA, a Part B cancer claim for benefits must be based on an energy employee’s eligible 

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

designated time period and location (i.e., a “covered employee with cancer”). After DOL determines 

that a claim meets the eligibility requirements under Part B of EEOICPA, DOL transmits the claim to 

NIOSH for a dose reconstruction. EEOICPA provides, among other things, guidance on eligible 

employment and the types of radiation exposure to be included in an individual dose reconstruction. 

Under EEOICPA, eligible employment at a DOE facility includes individuals who are or were 

employed by DOE and its predecessor agencies, as well as their contractors and subcontractors at 

the facility. 42 U.S.C. § 7384I(11). Also under EEOICPA, the types of exposure to be included in dose 

reconstructions for DOE employees are those radiation exposures incurred in the performance of 

duty. As such, NIOSH includes all radiation exposures received as a condition of employment at DOE 

facilities in its dose reconstructions for covered employees, which may include radiation exposures 

related to the Naval Nuclear Propulsion Program at DOE facilities, if applicable. This is because 

NIOSH does not determine the fraction of total measured radiation exposure at a DOE facility that is 

contributed by the Naval Nuclear Propulsion Program at the DOE facility during a specified period of 

time for inclusion in dose reconstruction.

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

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

dose reconstructions for covered employees [NIOSH 2010a]:

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

conventional structures, and

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

reasons.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 10 of 78

1.1 PURPOSE

This site profile provides information about operations at Ames Laboratory that pertains to radiation 

exposures for monitored or unmonitored workers. The laboratory began as several buildings on the 

grounds of the Iowa State College, which later became Iowa State University (ISU). This document 

uses ISU for either.

1.2 SCOPE

Section 2.0 provides a description of the site and operations that pertain to possible radiation 

exposures and discusses radiation source terms. Section 3.0 provides guidance for the determination 

of occupational medical dose. Section 4.0 provides guidance for the determination of dose to workers 

outside radiological facilities due to releases of radioactive materials to the environment as well as 

radiation emitted by facilities, reaching the surrounding areas. Section 5.0 provides guidance for the 

determination of intakes of radionuclides inside facilities. Section 6.0 provides guidance for the 

determination of external doses from measured doses and for periods for which records of measured 

doses are missing. Attributions and annotations, indicated by bracketed callouts and used to identify 

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

1.3 SPECIAL EXPOSURE COHORT

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

employees at Ames Laboratory as additions to the SEC:

January 1, 1942 (later revised to August 13, 1942; see below) through December 31, 1954

Department of Energy (DOE) employees or DOE contractor or subcontractor employees 

who worked at the Ames Laboratory in one or more of the following facilities/locations: 

Chemistry Annex 1 (also known as “the old women's gymnasium” and “Little Ankeny”), 

Chemistry Annex 2,1 Chemistry Building (also known as “Gilman Hall”), Research 

Building, or the Metallurgical Building (also known as “Harley Wilhelm Hall”) from 

January 1, 1942 through December 31, 1954 for a number of work days aggregating at 

least 250 work days, or in combination with work days within the parameters (excluding 

aggregate work day requirements) established for one or more classes of employees in 

the SEC, and who were monitored or should have been monitored [Leavitt 2006, p.3].

1 This document refers to these facilities as Physical Chemistry Annex 1 and Physical Chemistry Annex 2.

January 1, 1955, through December 31, 1970

Sheet metal workers, physical plant maintenance and associated support staff (including 

all maintenance shop personnel), and supervisory staff who were monitored or should 

have been monitored for potential internal radiation exposures associated with the 

maintenance and renovation activities of the thorium production areas in Wilhelm Hall 

(a.k.a. the Metallurgy Building or “Old” Metallurgy Building) at the Ames Laboratory from 

January 1, 1955, through December 31, 1970, for a number of work days aggregating at 

least 250 work days or in combination with work days within the parameters established 

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

[Leavitt 2007, p. 3].

January 1, 1955, through December 31, 1960

All employees of the Department of Energy, its predecessor agencies, and its contractors 

and subcontractors who worked in any area of the Department of Energy facility at the 

Ames Laboratory from January 1, 1955 through December 31, 1960, for a number of 

work days aggregating at least 250 work days, occurring either solely under this 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 11 of 78

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

or more other classes of employees in the Special Exposure Cohort [Sebelius 2010, p.3].

August 13, 1942, through December 31, 1970

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

contractors and subcontractors who worked in any area of the Ames Laboratory at Iowa 

State University during the periods from August 13, 1942 through December 31, 1970, 

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

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

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

Cohort [Sebelius 2011, p. 3].

January 1, 1971, through December 31, 1989

All employees of the Department of Energy, its predecessor agencies, and their 

contractors or subcontractors who worked in any area of the Ames Laboratory in Ames, 

Iowa, during the period from January 1, 1971, through December 31, 1989, for a number 

of work days aggregating at least 250 work days, occurring either solely under this 

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

or more other classes of employees included in the Special Exposure Cohort [Azar 

2018, p. 3].

The first two classes covered two separate groups of employees based on work location and job 

description [NIOSH 2006, 2007a]. However, NIOSH determined that the information available about 

worker job description, work location, or movement about the site was insufficient to determine if an 

employee worked in the affected areas [NIOSH 2010b]. In 2011, the Secretary designated a fourth 

class that encompassed all previous periods and included all Ames employees (including predecessor 

agencies, contractors, and subcontractors). In addition, the class was revised to change the covered 

period’s start date from January 1, 1942, to August 13, 1942, the start of the Manhattan Engineer 

District, known later as the Manhattan Project [NIOSH 2011a]. The Secretary subsequently 

designated a fifth class that extended the covered period to include January 1, 1971, through 

December 31, 1989 [NIOSH 2017]. 

Although NIOSH found that it is not possible to reconstruct radiation doses completely for all the 

evaluated classes, it intends to use any internal and external monitoring data that might become 

available for an individual claim (and that can be interpreted using existing dose reconstruction 

processes or procedures). Therefore, partial dose reconstructions for individuals employed at Ames 

Laboratory from August 13, 1942 through December 31, 1989, but who do not qualify for inclusion in 

the SEC, can be performed using these data as appropriate.

This site profile provides internal and external exposures that might coincide with work periods that fall 

within the class periods. There are varying types of exposures that can be applicable to dose 

reconstructions during the class periods for employees who do not qualify for inclusion in the SEC.

The feasibility findings for all SEC class periods were evaluated, with consideration of how latter 

designations affect earlier designations, and an overall covered period approach to dose 

reconstruction was developed [Rutherford 2022a,b]. Table 1-1 provides sources of exposure and 

periods for which dose reconstruction is feasible.

Exposures that are listed as feasible in Table 1-1 are discussed in later sections of this site profile. 

Exposures from other sources or nuclides not included in Table 1-1 are infeasible. After 1990, there 

are no SEC restrictions, as the SEC period ends December 31, 1989. Exposures after 1989 are 

covered in later sections of this document.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 12 of 78

Table 1-1. Feasibility findings for SEC classes, August 13, 1942, through December 31, 1989.

Internal dosea

Source of exposure Dates reconstruction is feasible

Uranium 08/13/1942–12/31/1970

All other nuclides None

Environmental (except fission products) 08/13/1942–12/31/1989

External dosea

Source of exposure Dates reconstruction is feasible

Co-exposure dose 01/01/1955–12/31/1989

Environmental 08/13/1942–12/31/1989

Occupational medical X-ray Not applicable

a. Any internal and external monitoring data that might become available for an 

individual claim (and that can be interpreted using existing dose reconstruction 

processes or procedures) can be used for dose reconstruction as described in the 

internal and external sections below.

2.0 SITE DESCRIPTION

The Ames Laboratory site consists of several buildings at ISU in Ames, Iowa. The precursor to the 

Ames Laboratory was the Ames Project, which was established in 1942 in a contract between the 

Metallurgical Laboratory at the University of Chicago and the former Iowa State College (now ISU) 

[Fulmer 1947]. Ames Laboratory was established by the U.S. Atomic Energy Commission (AEC) in 

May 1947 [Karsjen 2003]. The site played a key role in the production of strategic nuclear materials 

for the Manhattan Project and the AEC.

Early in 1942, before the beginning of the Manhattan Project, the most pressing problem was the 

preparation of large amounts of pure uranium metal [Ames 1960a]. Faculty members in the Chemistry 

Department at ISU with expertise in rare earth metallurgy were called on to develop a method to purify 

uranium and reduce its cost of production [Ames 1960a]. By November 1942, successful methods 

had been developed and approximately one-third of the uranium in the Chicago pile was supplied by 

the Ames Project [Karsjen 2003]. The Ames Project was asked to turn its process over to industry 

and, in the meantime, to produce as much pure uranium as possible. Between mid-1942 and August 

1945, more than 1,000 t of pure uranium metal was supplied to the Manhattan Project [Ames 1960a]. 

Once the potential need for thorium metal was recognized, the Ames Project began to develop 

methods for purifying thorium in 1943. By late 1944, a large-scale process for thorium metal 

production was developed; between 1950 and April 1953, when thorium production was turned over 

to industry, Ames produced more than 65 t of pure thorium metal and thorium compounds [Ames 

1960a]. 

In addition to the early uranium and thorium metal production operations, personnel at Ames 

Laboratory handled several other radionuclides and operated an 80-MeV synchrotron, a 5-MW 

research reactor, and several radiation-generating machines. Each of these radiation sources is 

described in the following sections.

2.1 FACILITIES

The original buildings at ISU for the Ames Project included Physical Chemistry Annex 1, the 

Chemistry Building (now Gillman Hall), and the Physics Building (now Physics Hall) [Ames 1967]. In 

1944, a new building, Physical Chemistry Annex 2, was constructed to house operations to recover 

uranium from scrap material [Fulmer 1947]. 
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In November 1945, the buildings used by the Ames Project were designated as the Institute for 

Atomic Research. In May 1947, AEC established a major research facility at Ames known as Ames 

Laboratory [Ames 1962a], which operated as an integral part of the Institute for Atomic Research. 

Four additional buildings were constructed: the Metallurgy Building (now Wilhelm Hall), the Research 

Building (now Spedding Hall), the Office and Laboratory Building that connected the Chemistry and 

Physics Buildings, and the Synchrotron Building (now the Spangler Geotechnical Laboratory) [Ames 

1962a]. The Metals Development Building was added in 1960 [Ames 1960a]. Construction of a 5-MW 

heavy-water-moderated research reactor began in 1962, and operations began in 1965. The 

Technical and Administrative Services Facility was added to connect the Research Building 

(Spedding Hall) and the Chemistry Building (Gilman Hall); this facility includes only offices. The 

locations of Ames Laboratory buildings on the ISU campus are shown in Figure 2-1. The locations of 

the reactor facilities (now known as the Applied Science Complex) and the Synchrotron Building are 

shown in Figure 2-2.

Figure 2-1. ISU campus map showing locations of laboratory buildings including former 

sites of Physical Chemistry Annex 1 and Annex 2 [ISU 2006]. 

To avoid confusion between the historical building names and the current names as shown in 

Figures 2-1 and 2-2, the following section headings provide the historical name first and the current 

name in parentheses.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 14 of 78

Figure 2-2. ISU campus map showing locations of reactor facilities and 

Synchrotron Building [ISU 2006].

2.1.1 Physical Chemistry Annex 1

The production of uranium metal was conducted in Physical Chemistry Annex 1, an old wooden 

structure east of the Dairy Industries building and west of Wallace Road. Uranium operations began 

there in mid-1942 and ended on August 5, 1945, when the uranium purification process was 

transferred to industry. More than 1,000 t of pure uranium and more than 300 t of uranium scrap were 

produced during this period [Karsjen 2003]. In 1943, an open porch area was enclosed (to control 

dusty operations) and additions were constructed to accommodate increases in uranium production 

[Payne 1992]. Beginning in 1943, the building was also used to produce thorium metal until the 

processing equipment was transferred to the new Metallurgy Building in 1949 or 1950 [Ames 1960a]. 

Physical Chemistry Annex 1 was torn down in 1953 [Karsjen 2003]. The building site was 

decontaminated, surveyed in May 1976, and designated acceptable for future construction 

[Voss 1979a, p. 2]. 

2.1.2 Chemistry Building (Gilman Hall)

The initial Ames Project work was conducted in the Chemistry Building in early 1942. The process for 

purifying uranium metal and the methods and equipment to increase production were developed in 

this building. Uranium production operations were moved to Physical Chemistry Annex 1 in mid-1942, 

while other uranium research continued in the Chemistry Building such as determination of uranium 

properties, studies of uranium corrosion, development of protective coatings for uranium, and 

development of uranium alloys and compounds [Fulmer 1947]. Other research in the Chemistry 

Building involved development of pure thorium metal, thorium alloys and compounds, yttrium metal, 

cerium metal, and beryllium metal [Ames 1962a]. Analytical work centered on plutonium chemistry 

and the radiochemistry of the separation of fission products from uranium and plutonium, which was 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 15 of 78

conducted in the “hot laboratory” between 1942 and 1951 [Ames 1960a]. The Chemistry Building was 

decontaminated and surveyed in May 1976 [Voss 1979a, p. 2].

2.1.3 Physics Building (Physics Hall)

Research and development (R&D) to support other work at Ames Laboratory was conducted in the 

Physics Building. Analytical equipment was developed, including three beta-ray spectrometers, a bent 

crystal X-ray spectrometer, a kevatron, X-ray and neutron diffraction spectrometers, scintillation and 

conduction crystal spectrometers, X-ray and electron diffraction machines, and an electron 

microscope [Ames 1951, 1962a]. Nuclear fission was studied to identify the individual fission 

fragments, the energies involved, and the ionized state of the emitted particles [Ames 1951]. 

Research was conducted to determine the stopping power and shielding properties of various solid 

materials [Ames 1951]. Personnel of the Physics Department operated the 80-MeV synchrotron, 

which is described below [Ames 1962a]. In 1968, an addition to the Physics Building was completed.

The Physics Building Addition (also known as Physics Hall Addition) was a 44,000-square foot 

building constructed to house the graduate research program and the advanced undergraduate 

teaching laboratories of the ISU Physics Department [Ames 1968].

2.1.4 Physical Chemistry Annex 2

Physical Chemistry Annex 2 was a brick fireproof structure built east of Wallace Road in early 1944 to 

house uranium recovery from scrap uranium metal turnings from other Manhattan Project sites.

Operations in this building through December 1945 produced more than 300 t of recovered uranium 

metal [Fulmer 1947]. Operations ended in 1953 when the building was converted to a plumbing shop; 

it was razed in 1972 [Struss and Hawkins 1986]. The area where the building stood was covered with 

concrete and served as a parking lot and loading zone. The area was surveyed in May 1979 [Voss 

1979a]. Part of the General Services Building now covers part of the former building site (see 

Figure 2-1).

2.1.5 Metallurgy Building (Wilhelm Hall)

The Metallurgy Building, constructed by the AEC, was completed in October 1949 [Ames 1951]. The 

building housed research aimed at development of special metals and alloys used in nuclear energy 

projects. Zircaloy was initially developed at Ames Laboratory as part of a basic study of the zirconiumtin alloy phase diagram [Ames 1962a]. The subject of reactor coolants was studied along with the 

heat-transfer properties of various metals and alloys [Ames 1951]. Equipment available for research, 

development, and production in metallurgy included many types of furnaces, high-vacuum systems, 

pyrometric devices, fabricating and testing machines, metallographs, X-ray diffractometers, and 

ultrasonic, spectrographic, dilatometric, and other instrument types for examination and study of 

metals and alloys. A glovebox line in the Metallurgy Building was used to study the behavior of 

plutonium in molten metal systems [Ames 1962a]. 

Thorium production and research activities were moved from Physical Chemistry Annex 1 to the 

Metallurgy Building in 1949, and work on thorium continued until 1953. Poor contamination control 

practices and poor ventilation contributed to contamination of the building. However, contamination 

levels have been reduced by mitigation, decontamination, remodeling, and renovation projects.

Contamination still exists in many interspatial areas of the building and in some relatively inaccessible 

areas [Hokel et al. 1998]. 

2.1.6 Research Building (Spedding Hall)

The Research Building was constructed by the AEC and occupied in early 1951 [Ames 1951]. Many 

metals, including rare earths, were investigated for mechanical, chemical, electrical, and other 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 16 of 78

properties and were studied by experimental techniques that probed the inner structures and forces of 

the materials [Ames 1962a]. Research facilities in the building included a 150-kV accelerator that 

produced 14-MeV neutrons; a glovebox line for radiochemistry experiments; a hot canyon and hot cell 

with steel shielding, lead glass windows, and manipulators for work with highly radioactive materials; 

and an electron microprobe analyzer. The hot canyon was two stories high with the lower level in the 

basement adjacent to the hot cell. Research activities included electron beam welding, the study of 

the electronic structure of metals, and the separation, preparation, and measurement of properties of 

rare earth metals. The initial research on liquid metal coolants was done at Ames Laboratory in an 

engineering sodium test loop used in corrosion, fluid-flow, and heat transfer studies with liquid sodium 

[Ames 1967]. Work in the hot cell continued until 1982.

Ames Laboratory had a Texas Nuclear Model 9900 neutron generator located in the Research 

Building. The neutron generator used tritium targets with activities ranging from 5 to 10 Ci. [Voss 

1971].

2.1.7 Office and Laboratory Building

The Office and Laboratory Building connects the Chemistry and Physics Buildings and houses the 

administrative offices of the Ames Laboratory, the special research laboratories used jointly by 

chemists and physicists, and a large physical sciences reading room [Ames, no date]. 

2.1.8 Synchrotron Building (Spangler Geotechnical Laboratory)

The Synchrotron Building was constructed in 1949 on a 200-acre tract northwest of the campus that 

was set aside for special use by the Institute for Atomic Research [Ames 1962a]. The synchrotron 

room housed two electron accelerators that could project electrons up to 80 MeV onto a target, which 

produced high-energy gamma rays that interacted with nuclei to release neutrons, protons, and alpha 

particles [Ames 1967]. The accelerators were operated from a control room where there was a safety 

gate that prohibited access to the synchrotron room when the beam was on [Ames 1967]. In many 

cases, the products of these reactions were radioactive and were used in research in nuclear physics 

and radiochemistry [Ames 1967]. In addition, the synchrotron was used to probe nuclear structures 

and to provide radionuclides for nuclear spectroscopy [Ames 1962a]. Operations at the Synchrotron 

Building ended in June 1971; the equipment was decommissioned in the early 1990s.

2.1.9 Metals Development Building

Ames Laboratory facilities were expanded in 1960 to include the Metals Development Building. Its 

missions were to conduct process development research on larger-than-laboratory scale and to 

evaluate commercial feasibility of the developed processes. One major process was the production of 

very pure metals; small quantities of these metals were to be used elsewhere as standards. The 

building contained a complete pilot plant with facilities for each step of the metal production process 

from ore treatment to metal fabrication or analysis [Ames 1960a]. Equipment in the building included 

electron microscopes, an electron microprobe, metallography apparatus, liquid-liquid extraction 

apparatus, extrusion presses for producing rods and tubes, and rolling machines for making sheet 

metal [Ames 1967]. 

2.1.10 Ames Laboratory Research Reactor (Applied Science Complex)

Construction of the 5-MW, heavy-water-moderated Ames Laboratory Research Reactor (ALRR) 

began in 1961; the reactor was first operated in February 1965 [Ames 1967; Voigt 1981]. The reactor 

and its support facilities were about 1.5 mi northwest of the ISU campus on a 200-acre site used by 

the Institute for Atomic Research [Ames 1967]. The reactor fuel was 93%-enriched 235U contained in 

24 fuel assemblies in a hexagonal arrangement in a core 30 in. across and 25 in. high [Voigt 1981]. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 17 of 78

The reactor shielding was an irregular decahedral prism shape with a thermal column on one face and 

nine faces with beam tubes from which radiation beams (primarily neutrons) could be extracted and 

directed to experimental areas around the reactor [Ames 1967; Voigt 1981]. Other tubes and thimbles 

provided access to the reactor core for irradiation experiments. Research activities included radiation 

damage studies, determination of the crystalline structure of solids, determination of mechanical 

properties of reactor materials, and analysis of the decay products of nuclear fission. The 

experimental equipment featured an online isotope separator that received fission products directly 

from the operating reactor, separated them by weight, and analyzed them by isotope. Other research 

equipment included a neutron diffractometer used to determine the physical properties of solids and a 

hot cell for handling spent reactor fuel [Ames 1967]. Operation of the reactor resulted in airborne 

tritium concentrations in occupied spaces of the building [Voigt 1981]. A routine tritium bioassay 

program was part of the radiation safety program at the reactor [Voss 1971]. Operation of the reactor 

ended in December 1977 and decontamination and decommissioning (D&D) of the facilities was 

completed in 1981. At the time operations ended, the heavy-water coolant contained approximately 

1.7 Ci/L of tritium [Voigt 1981]. 

2.2 OPERATIONS

Two major operations at Ames Laboratory resulted in radiation exposure to the staff: the production of 

large quantities of pure uranium (1942 to 1945) and the production of thorium metal (1943 to 1953).

Several smaller operations contributed to staff exposure to radiation.

2.2.1 Uranium Metal Operations

The initial Ames process for production of uranium metal was based on the chemical reduction of 

uranium tetrafluoride (UF4) by calcium metal. Finely ground UF4 was mixed with granulated calcium 

metal and the mixture was poured into a refractory-lined container. A fuse wire buried in the charge 

was electrically heated to initiate the reaction, which continued until both uranium metal and calcium 

fluoride were in the molten state. The more dense uranium collected at the bottom of the container, 

where it was allowed to cool to room temperature, after which it was removed for casting. The 

uranium metal was cast by placing it in a graphite crucible, heating it in a vacuum, and allowing the 

liquid metal to flow into a graphite mold for specific shapes [Fulmer 1947]. Although more complex, 

the uranium production process was improved by replacing the calcium reagent with magnesium 

metal [Fulmer 1947]. 

Most of the uranium metal production operations were conducted in Physical Chemistry Annex 1, 

which had poor contamination control and poor ventilation [Friedell 1942]. Workers in this building 

were likely to have received intakes of uranium between 1942 and 1945 and thorium between 1943 

and 1950 as well as external exposures to beta and gamma radiation. These exposures are estimated 

in Sections 5.0 and 6.0. Effluents from the building were not monitored, and local environmental 

contamination by uranium compounds was likely. Uranium and thorium contamination of the ground 

surface around Annex 1 required removal and offsite disposal of the contaminated soil [Payne 1992]. 

No other records could be found that addressed the disposal of contaminated waste or the control and 

monitoring of air and liquid effluents.

A substantial amount of uranium scrap material was produced, which was processed in Physical 

Chemistry Annex 2 from early 1944 to December 1945 [Fulmer 1947]. Radiation exposures to 

workers in this building during this period were similar to the uranium exposures in Physical Chemistry 

Annex 1, and are estimated in Sections 5.0 and 6.0. Local environmental contamination around 

Annex 2 was assumed to be similar to that around Annex 1 (see Section 4.1). 
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2.2.2 Thorium Metal Operations

The uranium metal production method was adjusted between August 1943 and August 1944 to 

produce thorium metal [Fulmer 1947]. Thorium tetrafluoride (ThF4) was mixed with calcium metal with 

a zinc chloride booster to produce a thorium-zinc alloy with a 96% yield of thorium metal. The alloy 

was heated under vacuum in a graphite crucible to distill off the zinc. Casting of thorium metal was 

difficult because of its high melting point and its reactive properties. Beryllium oxide crucibles had to 

be used; melting the thorium in a crucible the size and shape of the desired ingot proved to be a more 

reliable method of casting, although it often did not separate well from the slag and oxide. Castings 

were improved in late 1946 by pouring molten thorium into graphite molds [Fulmer 1947]. Production 

of thorium metal and thorium compounds continued until April 1953 when thorium production 

operations were turned over to industry [Ames 1960a]. 

The purified feed material for the thorium production operation was prepared by dissolving thorium 

nitrate in nitric and oxalic acids, precipitating the thorium oxalate, drying the precipitate in trays, 

hydrofluorinating the precipitate to ThF4, and crushing the ThF4 to a fine powder [Fulmer 1947]. This 

process was used throughout thorium production operations.

Thorium production operations, which were conducted in Physical Chemistry Annex 1 after uranium 

operations ended in 1945, continued until 1949 or 1950 when the operation and equipment were 

moved to the new Metallurgy Building (Wilhelm Hall) [Ames 1960a]. Workers in Annex 1 were likely to 

have intakes of thorium and external exposures to beta and gamma radiation. These exposures are 

estimated in Sections 5.0 and 6.0. Effluents from the building were not monitored, and local 

environmental contamination by thorium compounds was likely. Thorium production operations in the 

Metallurgy Building improved with better ventilation, but personnel exposures and environmental 

contamination continued to be unquantified until 1953 because of a lack of monitoring and inadequate 

records. 

A summary of AEC activities at Ames Laboratory during the period from 1942 to 1954 is presented in 

Table 2-1.

Table 2-1. Timeline of activities, 1942 to 1954. 

Operation Perioda Building Activity

Uranium metal production 02/1942–08/1942 Chemistry (Gilman Hall) Process development

Uranium metal production 08/1942–12/1942 Chemistry Production

Uranium metal production 09/1942–08/1945 Annex 1 Production

Uranium scrap recovery Late 1943–early 1944 Chemistry Process development

Uranium scrap recovery Early 1944–12/1953 Annex 2 Production

Uranium metal casting 09/1942–08/1945 Annex 1 Production

Thorium metal production 08/1943–06/1946 Annex 1 Process development

Thorium metal production 06/1946–late 1949 Annex 1 Production

Thorium metal production Early 1950–04/1953 Metallurgy (Wilhelm Hall) Production

Studies of plutonium 

properties

06/1943–12/1947 Chemistry & Metallurgy Research

Plutonium/fission product 

separations

Summer 1943–12/1954 Chemistry & Research 

(Spedding Hall)

Research and hot cell 

work

Thorium metal casting 06/1946–late 1949 Annex 1 Production

Thorium metal casting Early 1950–04/1953 Metallurgy Production

Studies of uranium and 

thorium properties

Early 1942–12/1954 Chemistry Research

Development of analytical 

procedures

Early 1943–12/1954 Chemistry Research

Annex 1 demolition 1953 Annex 1 Demolition

a. Exact start and end dates vary depending on the reference. Listed dates are the most common or consensus.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 19 of 78

2.2.3 Other Operations

Methods for routine analysis of fission products were developed at Ames Laboratory. These activities 

resulted in the discovery of the previously unidentified isotopes 33P, 144Pm, 125

Sb, and five isotopes of 

ruthenium and rhodium. Research on the parent-progeny relationship of 90Sr/Y was conducted.

Pioneering research in applications of alpha and beta spectroscopy and mass spectroscopy to identify 

specific radionuclides was part of the development of laboratory methods. A process for separating 

233U from thorium was developed.

2.3 SOURCE TERMS AND PERSONNEL MONITORING

Ames Laboratory staff were exposed to radiation sources through several pathways including 

environmental effluents, radioactive intakes, and external radiation dose. 

Environmental effluents from the early Ames Laboratory buildings were unmonitored and 

uncontrolled. Radioactive effluents near these buildings could have exposed workers to unmonitored 

occupational environmental doses. An accidental release of thorium waste materials to the sanitary

sewer system occurred in 1951 [Voss 1979a]. This release, which could have contributed to the 

occupational environmental dose received by workers who were involved in the incident response, is 

addressed in Section 4.4. 

Uranium and thorium metal production involved several dusty operations that resulted in work area 

contamination and potential worker inhalation and ingestion. The principal sources of surface 

contamination and airborne dust were the processes of grinding uranium tetrafluoride into a fine 

powder, transferring the powder from the grinder, and mixing and loading the powder charge into the 

reduction crucibles. Because the uranium had been separated from radium and its decay products, 

222Rn was not a potential inhalation concern. The principal sources of thorium surface contamination 

and airborne dust were the processes of preparing and drying the fine powder and mixing and loading 

the powder charge into the reduction crucibles [Fulmer 1947]. 

There were frequent small explosions and fires associated with the uranium and thorium production 

operations. Payne [1992] cited as many as six small fires in a single day; these fires contributed to 

work area contamination and potential airborne radioactive material exposures. No records were 

found to indicate that air sampling or contamination control was associated with these fires.

Personnel protection for potentially dusty operations included the use of laboratory-provided clothing 

and gloves as well as restrictions on eating and smoking in areas where radioactive materials were 

handled. Although respiratory protection measures such as gas masks and dust masks were 

provided, their use was not enforced before about 1952. Showers were recommended for workers at 

the end of each day, but records indicated that not all workers complied with this recommendation 

[Klevin 1952; Payne 1992]. 

Many other radioactive materials were handled in Ames Laboratory buildings before routine 

monitoring for radiation exposures began in about 1952. Because the bioassay program was minimal 

and records are sparse, intakes that might have occurred before 1953 have been estimated. When 

reactor operations started in 1965, a routine tritium bioassay program was started. Tritium intakes can 

be more reliably estimated from that time forward. Occupational internal dose is addressed in 

Section 5.0. 

The uranium and thorium metal production operations resulted in beta and gamma radiation 

exposures to workers. Beta radiation was the dominant external source of radiation associated with 

unshielded sources of uranium such as uranium metals production, scrap recovery, and machining 

processes. The significance of beta emissions from thorium depends on the state of equilibrium with 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 20 of 78

the 232Th parent, which is a factor of the time elapsed since the thorium process feed material was 

separated [NIOSH 2006]. Photon exposure rates as high as 22 mR/hr were reported for a thorium 

storage area, which suggests that this raw material for the thorium production process was not newly 

separated [Klevin 1952]. 

Only two film badge results were identified for 1944 with results in units of “average roentgens/8-hour

day during week” [Tybout 1944, p. 7]. Before 1952, only pencil dosimeters were used in Ames 

Laboratory facilities and the records are sparse. The use of film badges began in late 1952; records 

from 1953 and 1954, along with workplace measurements, have been used to estimate earlier 

radiation doses. Records of external radiation exposures received from 1955 to the present are given 

in Section 6.0 for evaluation of occupational external dose.

3.0 OCCUPATIONAL MEDICAL DOSE

Occupational medical exposures are included in dose reconstruction only for medical examinations 

obtained at covered facilities [Oak Ridge Associated Universities (ORAU) Team (ORAUT) 2017]. The 

X-ray equipment for occupational medical examinations of Ames Laboratory staff was at the Iowa 

State Student Health Center/College Hospital [Voss 1957], which is not a covered facility under 

EEOICPA. Therefore, no occupational medical doses should be included in dose reconstruction. 

4.0 OCCUPATIONAL ENVIRONMENTAL DOSE

Occupational environmental dose refers to the dose received by workers on the site but outside 

facilities (i.e., buildings). These doses can be internal or external depending on the characteristics of 

the individual radionuclides. Radionuclides at Ames Laboratory included uranium, plutonium, thorium, 

and small amounts of others used in the R&D program [Fulmer 1947]. Tritium, argon, and krypton 

were released at the ALRR. These radionuclides are addressed in the following sections.

Occupational environmental dose was not measured (direct radiation dosimeters) until 1953, when 

workers were badged [ORAUT 2006a,b], and it was not calculated from environmental media 

concentrations until 1962. Sources of potential environmental exposures (releases to the 

environment) were not measured until 1962 [Voss 1963, p. 7]. 

Different activities were carried out during distinct periods of Ames Laboratory history. Occupational 

environmental doses are, therefore, addressed below for each of these periods and their activities. A 

significant release of radioactive materials to the environment from Ames Laboratory facilities is 

addressed in Section 4.4. 

4.1 URANIUM AND THORIUM PRODUCTION, 1942 TO 1953

Uranium production occurred in Physical Chemistry Annex 1 (see Figure 2-1) from mid-1942 to

August 5, 1945 [Karsjen 2003]. Uranium scrap recovery occurred in Physical Chemistry Annex 2 (see 

Figure 2-1) from early 1944 to December 1945, and some operations continued until 1954. Thorium 

production occurred first in Physical Chemistry Annex 1 from 1943 to 1949 when the operation moved 

to the new Metallurgical Building (Wilhelm Hall; see Figure 2-1). The operation in the Metallurgical 

Building continued until April 1953. Like Annex 1 and Annex 2, there was no on- or offsite designation 

for this facility. Workers, students, and college personnel moved freely by and around the building. No 

measurements were made of the particulate or gaseous effluents from the buildings or of the radiation 

levels outside the buildings [Ames 1962b, 1962–1964].

Workers conducted their assignments inside the facilities, and there were no specific assignments 

outside the facilities. That is, input materials arrived at the facilities and were processed, and products 
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were shipped from the facilities. Workers did not move between the facilities on campus and there 

was no transportation of materials among the facilities on campus [Ames 1962a]. 

Uranium exposures were developed from bioassay data during uranium operations as described in 

Section 5.1. The environmental exposure intakes during the production period (1942 to 1953) are 

from Table 5-4 in Section 5.1.4 and are repeated in Table 4-8 in Section 4.6. 

Thorium environmental doses were derived from contamination surveys of operating areas. No 

documents were found that stated the room and hood ventilation stacks on the facilities had filters.

Concentrations of dust were measured in the operation rooms [Voss 1979b]. To estimate a bounding 

dose outside the facilities, it was assumed that losses of 0.1% of the thorium as dust in a facility were 

emitted continuously and dispersed from ground level by local and regional meteorological conditions 

(see Figure 4-1) [Voss 1982] and a standard Gaussian atmospheric dispersion computer model 

[Napier et al. 2004]. When resuspension is included, the daily intake rate is 0.07 pCi/d for each of 

232Th, 228Th, and 228Ra. This applies to Annex 1 for June 1943 to 1949 and to the Metallurgy Building 

for 1950 to April 1953. These are upper bound intake rates, so the distribution is constant 

[Napier 2007].

Figure 4-1. Wind rose for Des Moines, Iowa, 1950 to 1955 [Voss 1982].

4.2 SYNCHROTRON OPERATIONS, 1949 TO 1971

Synchrotron operations occurred in what is now called the Spangler Geotechnical Laboratory from 

1949 to June 1971 (see Figure 2-2). Unlike the uranium and thorium facilities, this facility was fenced, 

which provided a defined exclusion area outside the building. Dosimeters were not provided to 

anyone at the facility until late 1952 [ORAUT 2006c]. There were small research amounts of 

radioactive materials that were generated during operations, and negligible particulate or gaseous 

effluents were released from the building. This information is consistent with the site profile for the 

Stanford Linear Accelerator Center (SLAC) [ORAUT 2007], which found that the environmental dose 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 22 of 78

from accelerator effluents was less than 1 mrem/yr and negligible. No routine measurements of direct 

gamma or neutron radiation were made outside the buildings, but a detailed survey of the facility, 

including fence line gamma dose rates, was made on May 16, 1961, during a special synchrotron 

operation. The results of that survey are listed in Table 4-1 [Ames 1961a, p. 8].

Table 4-1. Survey of fence line 

gamma dose rates around the 

synchrotron facility with the beam 

directed west.

a

Location

Dose rate

(mrem/hr)b 

1 0.75–1

2 3.5

3 7

4 6

5 5.5

6 7

7 4.75

8 3.5

9 2.6

10 1.5

11 2

12 2

13 1.8

14 1.8

15 1.8

16 1.8

17 1.6

18 1.5

19 1.5

20 1.75

21 1.6

22 1.6

23 1.5

24 0.75–1

25 4.5

a. Source: Ames [1961a, p. 8].

b. Background reading in the beam 

direction (see locations 3 to 6) before 

turning on the beam was 0.5 to

1.0 mrem/hr.

The fence line gamma dose rates measured in the radiation survey [Ames 1961a, p. 8] were worst 

case and were produced with the maximum beam current on a target and direction that would 

produce maximum dose rates at the fence line [1]. This condition was most unusual in relation to 

typical research studies because it produced radiation levels in the normally occupied parts of the 

Synchrotron Building that were clearly hazardous to staff (tens of milliroentgen per hour) [Ames 

1961a, p. 5]. Dosimeter results for synchrotron personnel verified that the machine was not operated 

in this condition for significant periods (during the May 16, 1961, survey or at any other time) [ORAUT 

2006c].

The synchrotron was operated part time by Physics Department faculty and graduate students. The 

typical research schedule would have been weekdays and evenings plus occasional weekends, which 

was estimated at a maximum of 3,000 hr/yr. Much of this time would have been occupied with 

experiment setup, maintenance, system startup, etc., so the maximum operating time would have 
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been about 2,000 hr/yr [2]. Most of this operating time would have been dedicated to various material 

property studies that involved electron beams, beam currents, and targets that produced gamma 

fields outside the building that were at least a factor of 10 (and more likely a factor of 100) less than 

the survey dose rates in Ames [1961a, p. 5] [3]. 

A topographical map that was part of the 1961 radiation survey [Ames 1961a, p. 13] indicates the 

Synchrotron Building was surrounded by hills that would have protected nearby buildings from a direct 

beam. Therefore, the primary source of environmental exposures from this facility would have been 

skyshine from the synchrotron. The Waste Chemical Handling Facility (see Figure 2-2) was not built 

until 1980, so the area was not affected by synchrotron operations between 1949 and 1971. The 

ALRR (now the Applied Science Complex) was and is about 750 ft from the nearest part of the 

Synchrotron Building, and construction or operations at the two facilities overlapped, at a maximum, 

from 1962 to 1971. Skyshine from both heavy particles and photons decreases at rates equal to or 

greater than the reciprocal of the square of the distance (1/r

2

) from accelerator facilities [National 

Council on Radiation Protection and Measurements (NCRP) 2003]. If it is conservatively assumed 

that all the radiation field measured in the 1961 survey was from skyshine, the dose rate at the ALRR 

would have been less than 0.13 mrem/hr during the worst-case operation and less than 

0.013 mrem/hr during routine operations. On the main campus of the University, the dose rate would 

have decreased to less than about 0.00025 mrem/hr during routine operations.

It is favorable to claimants to use the environmental external dose from synchrotron operations at the 

ALRR for full-time exposure (2,000 hr/yr) for all locations at 25 mrem/yr for the period from 1949 to 

1971.

4.3 AMES LABORATORY RESEARCH REACTOR OPERATIONS, 1965 TO 1977

The ALRR began operations in February 1965 and continued through December 1977 [Ames 1967;

Voigt 1981]. The facility was and is surrounded by a fence about 700 ft from the reactor building that 

designates what is on and off the site (see Figures 2-2 and 4-2). From review of dosimetry records, it 

seems evident that all workers inside the ALRR fence were provided dosimeters. However, not all 

dosimetry records list names. Therefore, not all workers at the ALRR have recorded doses. In 

addition, environmental doses from gaseous effluents from the operating reactor were not monitored.

However, environmental doses to the public from airborne releases were calculated and reported 

[Voss 1975, 1976, 1977]. The only air monitoring station near the reactor was on the roof of the 

reactor building as shown in Figure 4-3. 

Environmental Monitoring at Ames Laboratory: Calendar Year 1974 was the first annual report to 

provide gamma spectroscopy of environmental media samples [Voss 1975]; subsequent annual 

reports [Voss 1976, 1977] provided similar results. The average release estimates from these reports 

were used to determine the estimated environmental dose to offsite workers from gaseous releases 

from reactor operations. From the effluent data, it was shown that the contribution to radioactivity in air 

from ALRR operations consisted principally of 41Ar and tritium [Voss 1975, 1976, 1977]. An 

atmospheric dispersion model, which used annual average meteorological data for Ames and an 

exposure model, was used to determine external dose rates from the 41Ar and inhalation intake 

estimates for the tritium [Napier et al. 2004; Napier 2006]. 

At the fence line location with the highest dose from gaseous effluents, the average annual dose to a 

person for the entire year (8,760 hours) was estimated to be 4.2 mrem from 41Ar during the years of 

reactor operation. A worker at this location full time would not have been exposed for more than 

2,080 hr/yr, which would result in a submersion dose of about 1 mrem/yr.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 24 of 78

Figure 4-2. ALRR site [Voss 1979b, p. 38]. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 25 of 78

Figure 4-3. Environmental monitoring near the ALRR [Voss 1975].

At the fence line location with the highest concentration of tritium effluents, the average annual intake 

of tritium to a person for the entire year (8,760 hours) was estimated to be about 2.9 μCi/yr during the 

years of reactor operation. A worker at this location full time would not have been exposed for more 

than 2,080 hr/yr, which would result in an intake of tritium of about 0.7 μCi/yr or 2,700 pCi/d. This 

tritium intake would result in annual internal doses less than 0.001 rem and is considered negligible.

4.4 SIGNIFICANT ENVIRONMENTAL EVENT

The only significant environmental event in the history of the Ames Laboratory was the release to the 

environment from operations that occurred from July 1951 to August 1952 [Hayes 1956]. Metallic 

thorium was being produced from thorium nitrate tetrahydrate. During an early stage of the process, a 

filtrate with traces of thorium in the form of thorium nitrate and oxalate was released to the sewer that 

connected to the City of Ames sewer system. At the time disposal to the sewer was chosen, it was 

believed that this waste had a very low level of radioactivity. However, due to a change in feed 

material supplied to the laboratory, considerable quantities of mesothorium (228Ra) were being 

discharged in the filtrate. Mesothorium is one of the progeny of thorium decay; it decays by emitting 

gamma rays and beta and alpha particles. All sewage was processed in a complete-treatment 

sewage plant, which resulted in liquid effluent and dry sludge that was used for fertilizer. The liquid 

effluent contained negligible quantities of mesothorium. The dry product was spread on lawns at the 

sewer plant, airport, municipal parkway, and a cemetery. A thorough study of the incident disclosed 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 26 of 78

that there was little hazard to the public or the workers at the sewer plant [Hayes 1956]. Table 4-2 is a 

summary of environmental dose measurements at the four facilities in Ames, Iowa. 

An aerial survey showed doses to be lower than those measured on the ground. Occupancy factors 

would reduce the doses in Table 4-2 by a large fraction, making them below the current public dose 

limit of 100 mrem/yr as given in 10 Code of Federal Regulations (C.F.R.) Part 20. 10 C.F.R. 20, 2021.

Table 4-2. Environmental dose measurements at the four 

facilities surveyed for thorium release effects (mrem/yr).

a

Lawn site Maximumb Average

Sewer plant 780 400

Cemetery 440 350

Airport 530 350

Parkway 530 350

a. Source: Voss [1979a]. 

b. Background in City of Ames and Rural Story County, Iowa were 

measured at approximately 75 mrem/yr. Source: Voss [1979a, p. 36].

All the locations in Table 4-2 are outside the ISU campus and accessible by the public except portions 

of the sewer plant. Most Ames Laboratory workers were not exposed to the radioactive materials 

released during this event and were not involved in responding to the event. Offsite exposures are not 

included as occupational environmental doses, therefore, no additional dose due to this event is 

recommended for occupational environmental dose for any worker.

However, the Ames Laboratory health physics staff responded to the release event and were exposed 

while making radiation measurements and collecting environmental samples for analysis. Any

radiation exposures to the health physics staff being monitored during the incident response might

have been included in their recorded occupational doses.

4.5 ENVIRONMENTAL CONTAMINATION, 1953 TO PRESENT

4.5.1 Residual Radioactivity Period Uranium Exposure

After 1953, the internal exposure potential to uranium was reduced and continued to be reduced 

throughout site operation. The median (50th percentile) intake rates derived in Section 5.1.3 were 

used to estimate surface contamination levels based on a 30-day suspension as discussed in 

Section 3.4.2 of Battelle-TBD-6000, Site Profiles for Atomic Weapons Employers that Worked 

Uranium Metals [NIOSH 2011b]. The contamination levels were multiplied by a resuspension factor of 

1 × 10-5 m-1 and a breathing rate of 9.6 m3

/d to determine the following inhalation intake rates:

• Type F: 1.681 pCi/d

• Type M: 6.891 pCi/d

• Type S: 134.8 pCi/d

These intake rates are applicable for 1954. For subsequent years, a gradual reduction of the 

contaminants is presumed in accordance with depletion factors provided in ORAUT-OTIB-0070, Dose 

Reconstruction During Residual Radioactivity Periods at Atomic Weapons Employer Facilities

[ORAUT 2012]. A depletion rate of 0.00067 per day was used to determine annual depletion factors, 

per guidance in ORAUT-OTIB-0070.

Table 4-3 presents the inhalation intake rates from residual radioactivity. A comparison should be 

made of the absorption types (F, M, S) and the intake resulting in the highest dose assigned. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 27 of 78

Table 4-3. Uranium inhalation intake rates from residual 

contamination (pCi/d).a

Year Type F Type M Type S

1954 1.68 6.89 134.8

1955 1.32 5.40 105.5

1956 1.03 4.22 82.61

1957 0.81 3.31 64.68

1958 0.63 2.59 50.67

1959 0.49 2.03 39.62

1960 0.39 1.59 31.13

1961 0.30 1.25 24.39

1962 0.24 0.97 19.00

1963 0.19 0.76 14.96

1964 0.15 0.60 11.68

1965 0.11 0.47 9.15

1966 0.089 0.37 7.17

1967 0.070 0.29 5.61

1968 0.055 0.22 4.39

1969 0.043 0.18 3.44

1970 0.034 0.14 2.70

1971 0.026 0.11 2.10

1972 0.021 0.085 1.66

1973 0.016 0.066 1.29

1974 0.013 0.052 1.01

1975 0.010 0.041 0.79

1976 0.0077 0.032 0.62

1977 0.0061 0.025 0.49

1978 0.0048 0.020 0.38

1979 0.0037 0.015 0.30

1980 0.0029 0.012 0.23

1981 0.0023 0.0094 0.18

1982 0.0018 0.0073 0.14

1983-end 0.0014 0.0057 0.11

a. Intake rates are normalized to calendar days and assigned as a lognormal 

distribution with a GSD of 3.

Ingestion intakes during the residual period are estimated from the derived surface contamination 

levels using the methods specified in OCAS-TIB-009 [NIOSH 2004]. The annual ingestion rates were 

then determined by multiplying the initial intake rate by the annual depletion factors per guidance in 

ORAUT-OTIB-0070. Ingestion intakes are provided in Table 4-4. These intakes should be assigned 

along with their associated inhalation values from Table 4-3.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 28 of 78

Table 4-4. Uranium associated ingestion intake rates from

residual contamination (pCi/d).a

Year

Type F

(f1=0.02)

Type M

(f1=0.02)

Type S

(f1=0.002)

1954 3.5E−02 1.4E−01 2.8E+00

1955 2.7E−02 1.1E−01 2.2E+00

1956 2.1E−02 8.8E−02 1.7E+00

1957 1.7E−02 6.9E−02 1.3E+00

1958 1.3E−02 5.4E−02 1.1E+00

1959 1.0E−02 4.2E−02 8.3E−01

1960 8.1E−03 3.3E−02 6.5E−01

1961 6.3E−03 2.6E−02 5.1E−01

1962 4.9E−03 2.0E−02 4.0E−01

1963 3.9E−03 1.6E−02 3.1E−01

1964 3.0E−03 1.2E−02 2.4E−01

1965 2.4E−03 9.7E−03 1.9E−01

1966 1.9E−03 7.6E−03 1.5E−01

1967 1.5E−03 6.0E−03 1.2E−01

1968 1.1E−03 4.7E−03 9.2E−02

1969 8.9E−04 3.7E−03 7.2E−02

1970 7.0E−04 2.9E−03 5.6E−02

1971 5.5E−04 2.2E−03 4.4E−02

1972 4.3E-04 1.8E-03 3.5E-02

1973 3.4E−04 1.4E−03 2.7E−02

1974 2.6E−04 1.1E−03 2.1E−02

1975 2.1E−04 8.4E−04 1.7E−02

1976 1.6E−04 6.6E−04 1.3E−02

1977 1.3E−04 5.2E−04 1.0E−02

1978 9.9E−05 4.1E−04 7.9E−03

1979 7.7E−05 3.2E−04 6.2E−03

1980 6.1E−05 2.5E−04 4.9E−03

1981 4.8E−05 2.0E−04 3.8E−03

1982 3.7E−05 1.5E−04 3.0E−03

1983-end 2.9E−05 1.2E−04 2.3E−03

a. Intake rates are normalized to calendar days and assigned as a lognormal 

distribution with a GSD of 3.

4.5.2 Other Nuclides

Gamma spectral analysis of samples of environmental media did not begin until November 1974 

[Ames 1974]. By that time, uranium and thorium production and scrap recovery had ceased and 

Annex 1 had been demolished in 1953 [Karsjen 2003]. Annex 2 operations ceased in 1953, and the 

building was razed in 1972 [Struss and Hawkins 1986]. Synchrotron operations ceased in 1971, and 

the building was decommissioned in 1990. ALRR operations ceased in 1977, and D&D was 

completed in 1981 [Voigt 1981]. Only R&D activities continued after 1981.

Soil samples collected at five locations around the campus in 1974 showed only 137Cs in very low 

concentrations as listed in Table 4-5. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 29 of 78

Table 4-5. Gamma analysis of soil samples.a

Site

Sample weight

(g)

Cs-137

(pCi/g)b 

1S 336 0.839

2S 332 1.100

4S 500 1.100

6S 341 0.991

7S 426 0.286

Background 336 0.294

a. Source: Voss [1975].

b. Gamma detection limit for Cs-137 in soil samples was 0.01 pCi/g.

A summary of soil samples from 24 locations around the ISU campus in 1974 for alpha and beta 

particle activity is given in Table 4-6. 

Table 4-6. Summary of 24 soil samples 

for beta and alpha activity (pCi/g).a

Activity Betab Alphac

Average 10.56 0.61

High 13.01 0.96

Low 7.66 0.38

Background 11.9 0.76

a. Source: Voss [1975]. 

b. Beta detection limit was 0.25 pCi/g.

c. Alpha detection limit was 0.10 pCi/g.

These analytical results are consistent with the low doses that Ames Laboratory has reported over the 

years and with the policy of not monitoring workers in the environs of laboratory facilities. They are the 

bases for demonstrating that environmental intakes from contamination other than uranium and 

thorium discussed above were negligible.

4.6 SUMMARY OF ENVIRONMENTAL EXTERNAL DOSES AND INTAKES

Tables 4-7 and 4-8 summarize the conclusions of the preceding sections on environmental external 

doses and intakes.

Table 4-7. Environmental external doses (mrem/yr). 

Location Period Dosea

Skyshine from Synchrotron Building 1949–06/1971 25

Ar-41 from ALRR 1965–1977 1

a. The energy range for all environmental external dose is assumed to be 100% 30 to 250 keV 

[NIOSH 2007b]. Distribution is constant. To be applied only to unmonitored workers.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 30 of 78

Table 4-8. Summary of environmental intakes (pCi/d).

a

Period Radionuclide/absorption Intake rate

08/1942–1953 U (assume U-234); type F, M or S See Table 5-4

1954–present U (assume U-234); type F, M, or S See Tables 4-3 

and 4-4

06/1943–04/1953 Th-232, Th-228; type M or S, Ra-228;

type M 

0.07b

05/1953–presentc Th-232, Th-228; type M or S, Ra-228;

type M 

0.0007b,d

a. Apply the environmental intakes in this table if no occupational intakes are applied for the same 

radionuclide and the same period in accordance with the Tables 5-2 and 5-3, or use the 

instructions in Table 5-6. 

b. Distribution is constant.

c. This period was expanded to include the period after the end of thorium operations.

d. Environmental intakes during research activities and afterward were considered to be 0.01 of the 

intakes for the production period.

5.0 OCCUPATIONAL INTERNAL DOSE

The radionuclides of interest for internal dose at Ames Laboratory are uranium, thorium, tritium, and 

fission products. NIOSH has determined that internal doses from radionuclides other than uranium 

cannot be reconstructed for the period before 1990. Uranium can only be assessed for the period 

before 1971. Research using small quantities (micrograms) of plutonium in the 1940s involved 

methods of extraction of plutonium from up to 5 Ci of fission products in irradiated fuel, which would 

not have contained much plutonium [Fulmer 1947]. There is no indication of research that involved 

sufficient plutonium such that exposure to residual plutonium after 1989 would have been significant.

Similarly, internal exposures from mixed fission or activation products from the ALRR would not have 

been significant after operations ceased and the facilities were decommissioned. Tritium exposure 

from the ALRR operations is addressed in this section and can be assessed when bioassay results

are provided for a worker.

Residual thorium contamination is addressed to assess the residual exposures beginning in 1990. 

5.1 URANIUM EXPOSURE

Few data on uranium exposure were found; uranium bioassay data from 1944 and 1945 were 

collected for a study on the urinary excretion of uranium at some Manhattan Project sites [Ferretti et 

al., no date]. The evaluation report for SEC-00038 [NIOSH 2006] considered the bioassay data to be 

useful for estimating intakes. If uranium bioassay data are available for an individual worker, they 

should be used for dose reconstruction. For all other workers in all uranium facilities at Ames 

(Chemistry Building, Physical Chemistry Annex 1, and Physical Chemistry Annex 2) the data from 

Tables 7.1 to 7.4 in Ferretti et al. [no date] were used to estimate intakes from uranium production 

operations.

5.1.1 Analysis of Uranium Bioassay Data

The data in Ferretti et al. [no date] reported 67 sample results from 48 workers who were ranked by 

their supervisor into four exposure categories according to potential for exposure to uranium. The 

bioassay data are presented in Table 5-1.

Table 5-1. Ames uranium worker bioassay data (μg/L).a

Group Sample Case Concentration

1 1 1 40

1 2 1 96
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Group Sample Case Concentration

1 3 2 52

1 4 3 86

1 5 3 50

1 6 4 100

1 7 4 44

1 8 4 70

1 9 4 200

1 10 5 126

1 11 5 96

1 12 5 74

1 13 6 84

1 14 6 200

1 15 6 73

1 16 7 48

1 17 7 40

1 18 8 29

1 19 9 25

1 20 10 12

1 21 11 31

2 1 12 15

2 2 13 17

2 3 14 13

2 4 15 38

2 5 16 21

2 6 17 40

2 7 18 21

2 8 19 33

2 9 19 58

2 10 20 33

2 11 21 54

2 12 22 64

2 13 23 10

2 14 23 16

2 15 24 11

2 16 24 11

2 17 25 87

2 18 25 64

2 19 25 80

2 20 26 130

2 21 27 80

2 22 28 108

2 23 29 64

2 24 29 64

2 25 30 28

2 26 31 43

3 1 32 24

3 2 33 27

3 3 34 7

3 4 35 9

3 5 36 19

3 6 37 22

3 7 38 22

3 8 38 18

3 9 38 3

3 10 38 3
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Group Sample Case Concentration

3 11 39 5

3 12 40 18

3 13 41 15

3 14 42 33

4 1 43 <3

4 2 44 <3

4 3 45 7

4 4 46 <3

4 5 47 <3

4 6 48 9

a. Source: Ferretti et al. [no date].

Group 1 was expected to be the highest (greatest amount of uranium exposure); Group 2 (next 

highest amount of uranium exposure); Group 3 (very little, but continuous exposure); and Group 4, the 

lowest exposure (occasional incidental exposure). The workers were given strict instructions to avoid 

sample contamination. Samples were submitted between September 1944 and July 1945 [Ferretti et 

al., no date].

Bioassay sample dates were not provided; however, the data reflect that some individuals were 

sampled more than once at intervals of a few weeks or months. For this analysis, all unique 

combinations of individual (Case) and group (Group) are considered unique individuals. If an 

individual has more than one result, a mean was calculated, thereby having one result for each 

individual. A one person–one statistic (OPOS) approach was performed as described in ORAUTRPRT-0053, Analysis of Stratified Coworker Datasets [ORAUT 2014]. A regression on order statistics 

fit to the OPOS data was performed and provided a geometric mean (GM) of 24.69 μg/L and a GSD 

of 2.788. From these parameters:

The 50th-percentile bioassay result is estimated by:

50th percentile = GM × 1.4 L/d (5-1)

giving

24.69 μg/L × 1.4 L/d = 34.57 μg/d (5-2)

These data were evaluated with the Integrated Modules for Bioassay Analysis (IMBA) to obtain 

uranium intake rates for calculation of internal doses to workers. The following parameters were used: 

• Start of chronic intake is September 1, 1942 (substantial production of uranium metal began in 

September 1942 at Ames).

• Date of the urine sample is January 31, 1945, which is midway between September 1, 1944, 

and July 1, 1945 (date range in Ferretti et al., no date] for bioassay samples).

• The chronic intake continues to the effective sample date of January 31, 1945.

• The specific activity of natural uranium (0.68296 pCi/μg) was used to convert the mass intake 

rates to activity intake rates.

• Based on process information, it is presumed that workers could have been exposed to 

absorption types F, M, or S materials.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 33 of 78

The 50th-percentile intake rates are:

• Type F. 86.46 pCi/d 

• Type M. 354.5 pCi/d

• Type S. 6,932 pCi/d

5.1.2 Worker Exposure Categories

In accordance with ORAUT-OTIB-0060, Internal Dose Reconstruction [ORAUT 2018]: “Coworker 

dose is applied as a best estimate for individuals with a potential for intakes of radioactive material but 

who lack bioassay data or have unmonitored intervals. Data can be lacking because it was not 

available from the site or because monitoring was not performed. Typically, workers with a significant 

potential for intake should be assigned doses at the 95th percentile with a constant distribution, while 

those with less potential should be assigned the 50th percentile with a lognormal distribution.” 

ORAUT-OTIB-0014, Assignment of Environmental Internal Doses for Employees Not Exposed to 

Airborne Radionuclides in the Workplace, provides guidance on job categories and potential for 

exposure [ORAUT 2004]. 

Dose reconstructors should use the information from these references and the category information 

below to assign intakes to Ames workers: 

• High (assign 95th-percentile intakes). Individuals who operated process equipment or routinely 

handled radiological materials. This category includes operators, maintenance workers, 

laboratory workers, health physics monitors, etc. Doses should be applied as a constant.

• Medium (assign 50th-percentile intakes). Individuals who routinely worked in production areas 

and may have been periodically present when processing occurred. This includes supervisory 

staff, engineers, and individuals not normally in contact with the radiological materials. Doses 

should be applied as a lognormal distribution.

• Low. Individuals exposed to ambient air in the environment outside of uranium production 

areas as well as incidentally exposed to uranium work areas. This includes office workers or 

nonuranium workers documented to have been in a different location from the uranium work. 

These intakes allow for incidental exposures and should be applied as a constant. The 

rationale for this value is provided below.

As discussed above, uranium bioassay data are available for 48 workers involved in uranium 

production in 1944 and 1945. The supervisor listed six of these as Group 4 workers thought to have 

only incidental occasional exposure to uranium. The distribution of the bioassay results for the four 

groups are consistent with those rankings, with some overlapping of the range of results. Of the six 

workers in the Group 4 category, four had a bioassay result of <3 μg/L, one had a result of 7 μg/L, and 

one had a result of 9 μg/L.

The Group 4 workers presumably would have been exposed to both ambient air outside the 

production facilities and occasional incidental exposure to uranium work areas. Therefore, the highest 

result from that group of workers should provide a bounding estimate for workers in other 

(nonuranium) facilities or locations. Dose reconstructors should use an intake rate derived from this 

value for the Low intake category and apply it as a constant.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 34 of 78

5.1.3 Intake Calculations

Intakes are provided for the High, Medium, and Low exposure categories.

High Exposure Category

This category is the 95th percentile of the intake distribution and is applied as a constant value in the 

Interactive RadioEpidemiological Program (IREP). Using ORAUT-OTIB-0060, the GSD was rounded 

up to 3.0. The 95th-percentile intakes are derived from the 50th-percentile (GM) intakes given above 

with the GSD rounded up to 3.0, according to the equation:

95th percentile = GM × GSD1.645 (5-3) 

where

GSD = 3.0

The 95th-percentile intake rates are then:

• Type F. 526.9 pCi/d 

• Type M. 2160 pCi/d

• Type S. 42,240 pCi/d

Medium Exposure Category

This category consists of the 50th-percentile intake values in Section 5.1.1 applied as a lognormal 

distribution with the GSD rounded up to 3.0. The intake rates are: 

• Type F. 86.46 pCi/d 

• Type M. 354.5 pCi/d

• Type S. 6932 pCi/d

Low Exposure Category

This category is based on an excretion rate of 9 μg/L with the parameters discussed in Section 5.1.1. 

Those values were input into IMBA to determine the following intake rates:

• Type F. 31.52 pCi/d 

• Type M. 129.2 pCi/d

• Type S. 2525 pCi/d

5.1.4 Assignment of Internal Exposures

The intakes were derived from bioassay during the uranium production period. Uranium production 

activities in Annex 1 and Annex 2 ended in 1945. Production work in the Chemistry Building ended in 

1943. There were continuing activities in all three buildings after 1945.

Records of decontamination activities for the Chemistry Building are currently unavailable, but records 

for later years indicate the building was remodeled.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 35 of 78

Thorium production continued at Annex 1 for a few years after 1945. No decontamination information 

is available. The facility was demolished in 1953; no radiological monitoring data are available from 

the demolition work.

The use of Annex 2 after 1945 is not clear and it is not known if or when the facility was 

decontaminated. However, the AEC sold Annex 2 to ISU in 1953; it was used as a plumbing shop

until it was razed in 1972.

To allow for potential intakes of uranium after 1945, the production era intake rates should be applied 

through 1953 for the Chemistry Building, Annex 1, and Annex 2.

Workers at the uranium production facilities could have been exposed to uranium with absorption 

types of F, M, and S. Dose reconstructors should select the material type that provides the highest 

dose. These calculated intake rates presume that the period of the bioassay results included normal 

uranium production operations, including blowout and fire events.

Tables 5-2, 5-3, and 5-4 show the inhalation intake rates for the three absorption types. All intakes are 

per calendar day.

Table 5-2. High exposure potential inhalation intake rates for 234U (pCi/d).a

Nuclide

Absorption 

type Intake type Intake rate Distribution

U-234 F Inhalation 526.9 Constant

U-234 M Inhalation 2,160 Constant

U-234 S Inhalation 42,240 Constant

a. These are based on the 95th percentile and should be assigned as a constant distribution 

from August 1, 1942, to December 31, 1953.

Table 5-3. Medium exposure potential inhalation intake rates for 234U (pCi/d).

a

Nuclide

Absorption 

type Intake type Intake rate Distribution GSD

U-234 F Inhalation 86.46 Lognormal 3.0

U-234 M Inhalation 354.5 Lognormal 3.0

U-234 S Inhalation 6,932 Lognormal 3.0

a. Based on the 50th percentile and should be assigned as a lognormal distribution with GSD of 3.0 from 

August 1, 1942, to December 31, 1953. 

Table 5-4. Low exposure potential inhalation intake rates for 234U (pCi/d).

a,b

Nuclide

Absorption 

type Intake type Intake ratea Distribution

U-234 F Inhalation 31.52 Constant

U-234 M Inhalation 129.2 Constant

U-234 S Inhalation 2,525 Constant

a. This is environmental exposure; see Section 4.1. 

b. Based on the maximum Group 4 category bioassay result and should be assigned as a 

constant distribution from August 1, 1942, to December 31, 1953 (see Section 4.5.1 for 

explanation).

5.2 THORIUM EXPOSURE FROM RESIDUAL CONTAMINATION

According to the feasibility table in Section 1.3, dose reconstruction is not feasible for nuclides other 

than uranium before December 31, 1989. This section provides a method for estimating internal 

exposures to thorium and thoron after 1989.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 36 of 78

The Metallurgy Building (Wilhelm Hall), the location of the main activities with thorium, is still in use.

Therefore, a method to estimate internal exposures from thorium is provided for workers in this 

building after 1989. 

In March 1952, a study was performed in Wilhelm Hall to identify health and safety issues that were 

occurring during refining and thorium metal production. The results of the study are in Ames 

Research Laboratory Occupational Exposure to Thorium and Beryllium [Klevin 1952]. Starting in 1984 

and continuing to the early 1990s, surveys were conducted in Wilhelm Hall to determine locations of 

contamination left from the early production years [Hokel et al. 1998]. The following discussion using 

data from Hokel et al. and Klevin provides an estimate for intakes from 1990 to 2010. A summary of 

the air concentrations and daily inhalation and ingestion intakes is provided in Table 5-5. 

All the data in Hokel et al. [1998] were reviewed and considered. Much of the data was related to 

locations that were hard to access and considered not to be an inhalation issue. The accessible areas 

of the building, including rooms, air ducts, hallways, stairwells, transformer rooms, etc., were 

surveyed starting in 1984. One data set had floor surveys made in 1988, but the locations all had

fixed activity, which would not pose an airborne hazard (see Appendix 6 of Hokel et al. 1998). The 

data in Hokel et al. present an overview of the survey results. In 1996, some measurements were 

made using an alpha continuous air monitor in the subbasement pipe tunnels and in large vertical 

void spaces in the stairwells. All results were less than background for thorium, thoron, or radon 

[Hokel et al. 1998]. Therefore, the results used for this estimation are from a pipe tunnel survey that 

showed removable contamination on smears ranging from background (3 dpm) to 1,224 dpm. These 

numbers are considered high for generally accessible areas. Although other survey numbers in the 

report with removable contamination are higher, they are in inaccessible locations (in a drain line), 

hard to reach (inside a drawer), or small (edge of a sink), or the contamination was fixed.

Hokel et al. [1998] stated that the thorium in Wilhelm Hall is in equilibrium. For estimating dose to 

individuals working in Wilhelm Hall from 1990 to 2010, an intake estimate favorable to claimants 

would be as follows.

A removable surface concentration of 2,000 dpm/100 cm2 of thorium in equilibrium with its progeny 

was assumed [4]. This means the 232Th activity was approximately 200 dpm/100 cm2

. Applying a 

resuspension factor of 10-4 [5]: 

( )( )( ) (5-4) 2 -4 2 2 3 A = 200dpm/100cm 10 /m 100 × 100cm /m = 2 dpm/m

where

A = 232Th air concentration

The value of 2 dpm/m3 on November 15, 1995, when the survey of the pipe tunnel occurred, 

represents an upper bound. 

To estimate residual air concentrations for other years, a single exponential decay of “available” 

activity over time in a facility was used. This applies to a facility in which the contamination is mostly 

undisturbed but undergoes some weathering or occasional cleaning such as minor moving of 

equipment or workbenches followed by cleaning.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 37 of 78

Table 5-5. Daily intakes of 232Th at Wilhelm Hall, 1990 

to 2010 (pCi/d).

a,b,c

Year

Inhalation

intake rate

Ingestion

intake rate

1990 12.2 11.4

1991 10.7 9.97

1992 9.34 8.72

1993 8.17 7.62

1994 7.14 6.67

1995 6.25 5.83

1996 5.47 5.10

1997 4.78 4.46

1998 4.18 3.90

1999 3.66 3.41

2000 3.20 2.99

2001 2.80 2.61

2002 2.45 2.28

2003 2.14 2.00

2004 1.87 1.75

2005 1.64 1.53

2006 1.43 1.34

2007 1.25 1.17

2008 1.10 1.02

2009 0.96 0.894

2010 0.84 0.782

a. Source: ORAUT [2023]. Distribution is constant.

b. Per the feasibility table in Section 1.3, exposures to nuclides 

other than uranium are not included before 12/31/1989. 

Therefore, thorium intakes are presented beginning in 1990.

c. Dose reconstructors should assign the 2010 value for 

employment years after 2010.

Klevin [1952] did a thorough survey during operations, including breathing-zone task-specific air 

concentrations, time-weighted air concentrations for various workers, and general room air 

concentrations. Because survey information in the building just after cessation of the thorium 

operations was not available, the Klevin data were used to represent air concentrations in 1955 with 

the caveats provided below:

• The Klevin air concentrations for specific tasks during production and the time-weighted 

averages are not relevant for Wilhelm Hall after 1954. The general room concentrations would 

probably be a bit high but reasonably representative when applied to long-term chronic intakes 

for people not actually performing tasks with thorium. These can be assumed to represent air 

concentrations in 1955. 

• The general air concentrations included one room that was 10 times higher than all the others 

and the lunchroom. The highest room was clearly related to operations so that datum was 

excluded, as was the lunchroom, which was assumed not to be contaminated (the result was 

zero). With these results removed, the geometric mean of the distribution was 7.9 dpm/m3 and 

the 95th percentile was 479 dpm/m3

. 

Because the 2 dpm/m3 from Hokel et al. [1998] represented an upper bound, this value is compared 

with the 479 dpm/m3 value. With 479 as A0 and 2 as A(t) and assuming t is from January 1, 1955, to 

November 15, 1995, for a total of 14,928 days: 

At A t ( ) exp = − 0 ( ) λ (5-5)
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 38 of 78

where

λ = 3.67 × 10-4/d or 0.134/yr

t = exposure time (d)

A(t) was solved for the midpoint of each year from 1995 to 2010. The air concentration was converted 

to intake using: 

( ) ( )( )( ) (5-6) 3 3 2,000 hr/yr pCi/d = dpm/m 1 pCi/2.22 dpm 1.2 m /hr 365 d/yr

I A      

where

I = the daily intake rate

A = the air concentration

The estimates in Table 5-5 for 232Th assume an occupancy factor of 1 (100% for a working year). This 

is an upper bound (constant) estimate for thorium intakes for individuals who worked in accessible 

areas in Wilhelm Hall from January 1990 to 2010. 

The dose reconstructor should include equal intakes of 228Ra and 228Th to account for the progeny 

radionuclides [6]. 

The associated ingestion intake for the start of the residual period would be:

I S F RO Ing = ÷ ×× ( ) (5-7)

where

S = 232Th surface contamination level on November 15, 1995, i.e. 200 dpm/100 cm2

F = fraction of the source term on November 15, 1995, i.e. Equation 5-5 above, with 

(t) = 14,928 d

R = ingestion rate [Biwer 2002], 1.12E-04 m2

/hour

O = occupancy, 2,000 hr/yr

This resulted in an initial ingestion rate at the start of the residual period, i.e. 01/01/1955, of 1,324 

pCi/calendar day. Since thorium exposures cannot be applied before December 31, 1989, Equation 5-

5 was used to calculate the residual ingestion intake rates, starting in 1990 (i.e. 35 years after 1955), 

and beyond. The daily ingestion intakes are listed in Table 5-5. 

Workers occasionally entered areas with higher concentrations under radiation work permits to 

perform maintenance or remodeling. However, the spotty contamination in these areas was about 10 

to 100 times greater than the value used in the calculation above [7]. Based on the rarity of exposure 

and the radiation protection measures used (including respiratory protection), the impacts from these 

possible acute intakes would have been adequately accounted for by the chronic intake scenario 

represented in Table 5-5 [8]. 

5.3 TRITIUM EXPOSURE

According to the feasibility table in Section 1.3, dose reconstruction is not feasible for nuclides other 

than uranium before December 31, 1989. This section provides a method to use, however, to 

estimate internal exposures due to tritium after 1989. Additionally, as stated in Section 1.3, monitoring 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 39 of 78

data can be used for individual claims to assess exposure. Therefore, this section provides a method 

to assess internal exposures due to tritium for monitored workers. The ALRR, a 5-MW, heavy water

moderated research reactor, operated from February 1965 to December 1977. Decommissioning was 

completed in 1981. See Section 2.1.10 for more information on reactor operations. Because of the 

operation of the reactor, tritium is the primary radionuclide that must be considered for internal dose.

No other radionuclides were considered internal dose hazards from the operation of the reactor. A 

routine bioassay program was established in 1965; bioassay data for workers were collected and 

recorded through at least 1981 while the reactor was being decontaminated and decommissioned 

[Voss 1971]. 

5.3.1 Tritium Exposure for Monitored Workers, 1965 to 1981

Several reports addressed tritium dose assessment. Highlights of those reports are given here and 

referenced for further study by the dose reconstructor.

Program for Monitoring Personnel for Tritium, Sources of Tritium at the Ames Laboratory describes 

potential sources of tritium exposure to laboratory personnel [Voss 1971]. 

In addition to the reactor, Ames Laboratory had a Texas Nuclear Model 9900 neutron generator, 

which was a potential source of exposure to tritium. The neutron generator used tritium targets with 

activities ranging from 5 to 10 Ci. The tritium, which was slowly released from the target, was 

documented as a potential airborne contaminant. Other possible sources of tritium were spent targets 

from the neutron generator, tritium gas in cylinders, and tritiated compounds that would have been 

present occasionally [Voss 1971]. 

Voss [1971] is summarized here to help understand dose records that might be in a worker’s files. It 

stated that workers and other personnel who were part of the tritium monitoring program included:

1. Reactor operators (monthly samples),

2. Reactor operators involved in specific operations that were likely to cause tritium exposures 

(sampled during and after the operation), 

3. Other reactor personnel and other personnel assigned to work at the reactor (provide samples 

semiannually for routine checks on tritium uptake), 

4. Personnel described in Item 3 who were involved in operations in which tritium exposure 

conditions existed (sampled during and after the operation), 

5. Other personnel under conditions of potential tritium exposure (monitored based on their work 

with sources of tritium), 

6. All personnel who showed tritium concentration greater than 1,500 dpm/mL of urine were 

resampled weekly until the detected activity level was below 1,500 dpm/mL, 

7. Samples were taken at the termination of workers assigned to areas where monitoring for 

tritium was required, and

8. As a check and control, termination samples were taken from other workers (it is not clear if it 

was all personnel or just some personnel).

Voss [1971] stated that liquid scintillation counting was used for tritium bioassay analysis.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 40 of 78

Assumptions described in Voss [1971] for radiation dose determination were:

1. The concentration of tritium was the same in all body water. Urine samples were used as 

representative of all body water in terms of tritium activity levels. The radiation dose was for 

the whole body.

2. The radiation dose was due to tritium as tritiated water; other compounds of tritium and other 

organ configurations are not included. Except in unusual circumstances, tritiated water was the 

expected source of exposure. Therefore, body water was the critical organ.

3. For calculation and derivation consistency, the mass of body water was taken as 43 kg 

(standard man value).

4. The effective half-life for dose determination could vary. The actual half-life from the data 

points was used up to a half-life of 15 days. If the half-life could not be determined from the 

data, a value of 12 days was used. Data points that showed a half-life greater than 15 days 

were treated as additional separate exposures.

5. The quality factor was 1, the effective absorbed energy per disintegration was 0.006 MeV, and 

the half-life of tritium was 12.26 years (4.48 × 103 days).

Other considerations for dose assignment that were discussed in Voss [1971] are:

1. The dose assignment system had two separate conditions of radiation dose delivery to the 

body: (1) the dose delivered by the uptake of tritium from time t = 0 until complete decayremoval (t = ∞) (additional uptakes might occur during the decay removal of the preceding 

uptakes) and (2) the dose that would be delivered by maintenance of a fixed level of tritium in 

the body.

2. For anyone with a bioassay sample greater than 1,500 dpm/mL in urine, sampling continued 

until the urine sample level was less than 1,500 dpm/mL.

3. For any person on the routine and incident monitoring program, a basic dose assignment was 

made. This basic dose was associated with the maintenance of a level of 750 dpm/mL of 

urine, which was equal to 38 mrem/yr whole body based on the methodology used at the time.

4. For single or multiple exposures, the dose assessment was made based on the effective halflife applicable to the individual’s data. The dose contribution from the incident exposure that 

would appear in the basic dose assignment was subtracted from the basic dose assignment.

5. The final reported dose was that due from an incident exposure plus the residual dose from 

the basic dose assignment. In summary, a person on the tritium monitoring program was 

assigned a basic radiation dose of 38 mrem/yr (prorated for a fraction of a year) to which was 

added the adjusted radiation dose contribution from all other exposures received during the 

year.

Several other documents discussed tritium dose assessment, but they are similar to the information 

provided above. Personnel Monitoring Program provides the tritium dose calculation procedures used 

at the time [Ames 1974].

Personnel in the tritium monitoring program should have bioassay and dose records. Tritium bioassay 

records were found for 1965 to 1981, although those for 1965 to 1968 appear to be incomplete.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 41 of 78

Figures 5-1 and 5-2 show examples of what a bioassay record in a worker’s file might look like. The 

data in this record include:

1. The year being reported.

2. The badge number of the individual. (Note: It appears that an individual would retain the same 

badge number during employment but, once they terminated, that badge number would be 

reassigned to a new worker.)

3. The name of the individual.

4. Dates during the reporting period for which monitoring for incidents was performed; the result 

was in “DPM.” It is assumed that incidents occurred during planned operations and that 

sampling was based on potential for intake.

5. Date that individual was employed. 

6. The individual’s dose before January 1 of the year for the record.

7. If the person has terminated, the termination date.

8. The four quarters for the year. The example in Figure 5-1 is for the third quarter and shows the 

dose for quarter 3. The example in Figure 5-2 is for the second quarter and shows the dose for 

quarter 2. The dose would be added and the total for the year calculated. It can be assumed 

that “0.00” means the worker was not monitored, and not necessarily that there was no dose.

Figure 5-1. Bioassay record example 1 [Ames 1977, p. 74].

Documents from the early 1970s indicate that a person in the tritium monitoring program would have 

been assigned a default radiation dose of 38 mrem/yr (or prorated for the time monitored during the 

year); however, a review of the available dose records indicated that the policy to assign a 38-

mrem/yr dose to everyone on the tritium program was not always followed [Voss 1971; Ames 1974]. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 42 of 78

Figure 5-2. Bioassay record example 2 [Ames 1977, p. 36].

No information was found that indicated the analysis method or the minimum detectable amount 

(MDA); therefore, it is recommended that 0.1 μCi/L (220 dpm/mL) be used as the MDA. This is a 

reasonable assumption in comparison with MDAs at other sites during this period and is consistent 

with the way the data were recorded [9]. 

5.3.2 Tritium Dose to Workers, 1990 to Present 

An interoffice memorandum dated May 8, 1987, addresses radiation dose at the Applied Science 

Center (inhalation exposure from tritium as tritiated water vapor in air resulting from deposition during 

reactor operations) [Voss 1987]. The memorandum addresses the methodology for estimation of dose 

to an individual in the reactor room assuming 100% occupancy without forced ventilation based on the 

standard man inhalation rates in Bureau of Radiological Health (BRH) [1970]. There is also an 

estimate for a maximum dose of 5.8 mrem/yr to an individual in the pump room. In a letter dated 

May 27, 1987, R.G. Struss, Associate Director for Operations, Ames Laboratory, discussed another 

estimate for airborne tritium concentrations that indicated an exposure potential of 8.6 mrem/yr in the 

reactor pump room for a 40-hr workweek [Struss 1987].

The dose reconstructor should assign a dose of 8.6 mrem/yr for personnel working in the Applied 

Science Center if no bioassay data are available. According to the feasibility table in Section 1.3, dose 

reconstruction is not feasible for nuclides other than uranium before December 31, 1989. Therefore, 

tritium exposures are assigned beginning in 1990.

5.4 FISSION PRODUCT INTAKES

5.4.1 Fission Product Intakes from Early Fuel Research

As described in Section 2.1.2, a hot laboratory was operated in the Chemistry Building; it was 

replaced in 1951 by a hot canyon/hot cell in the Research Building. In the 1940s, the hot laboratory 

was used to study extraction of plutonium from irradiated uranium by ion exchange columns

[Fulmer 1947]. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 43 of 78

According to the feasibility table in Section 1.3, dose reconstruction is not feasible for nuclides other 

than uranium before December 31, 1989. Therefore, fission product intakes cannot be estimated from 

these operations.

5.4.2 Fission Product Intakes from Research Reactor Operations and Decontamination 

and Decommissioning

No evidence was found during review of any of the Ames Laboratory documents that bioassay was 

performed or considered necessary for radionuclides other than tritium for the ALRR. Review of 

Decommissioning of the Ames Laboratory Research Reactor indicated that there was a negligible 

amount of fission products in the building after shutdown [Voigt 1981]. This indicates that there was 

probably a negligible amount of fission products released during reactor operations. Therefore, 

accounting for internal dose from reactor operations is not necessary.

During D&D operations, radiation protection appeared to be acceptable for the time and external 

exposure and internal exposure for tritium were tracked [Voigt 1981]. No evidence exists that 

bioassay for other radionuclides was performed. However, during D&D of the reactor, which included 

dismantlement, cutting, grinding, etc., some intakes from activation products probably occurred.

According to the feasibility table in Section 1.3, dose reconstruction is not feasible for nuclides other 

than uranium before December 31, 1989. Therefore, fission product intakes cannot be estimated from 

these activities.

5.5 SUMMARY OF INTERNAL DOSE RECOMMENDATIONS

Default occupational intakes for workers without the applicable bioassay data are summarized in 

Table 5-6. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 

09/18/2023 Page 44 of 78

Table 5-6. Dose calculation and IREP input parameters for default intakes (if no bioassay results are available).

Job category or task/building Period Material Mode

Absorption type or 

f1 Intake rate (pCi/d)

IREP 

Parameter 1

All workers in Chemistry Building, 

Annex 1, Annex 2

08/1942–12/1953 Natural uranium Chronic inhalation F, M, or S See Tables 5-2 and 5-3 Dose

Anyone routinely in Wilhelm Hall 

(Metallurgy Building)

1990–present Th-232 Chronic inhalation M or S See Table 5-5 Dose

Anyone routinely in Wilhelm Hall 1990–present Th-232 Chronic ingestion 5E-4 with M; 2E-4 

with S

See Table 5-5 Dose

Anyone routinely in Wilhelm Hall 1990–present Ra-228 Chronic inhalation M See Table 5-5 Dose

Anyone routinely in Wilhelm Hall 1990–present Ra-228 Chronic ingestion 0.2 See Table 5-5 Dose

Anyone routinely in Wilhelm Hall 1990–present Th-228 Chronic inhalation M or S See Table 5-5 Dose

Anyone routinely in Wilhelm Hall 1990–present Th-228 Chronic ingestion 5E–4 with M; 2E–4 

with S

See Table 5-5 Dose

Anyone routinely in Applied Science 

Center

1990–present Tritium Total of all modes Not applicable Not applicable 8.6 mrem/yr a

a. IREP distribution is constant.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 45 of 78

6.0 OCCUPATIONAL EXTERNAL DOSE

Workers at Ames Laboratory received external radiation doses between 1942 and 1952 that were

largely unmonitored [10]. Pocket chambers were available that might have been used to monitor 

external doses, but very few records could be found [11]. External exposures since 1953 have been 

monitored with film badges and thermoluminescent dosimeters (TLDs), and the records are 

essentially complete. However, there were extensive periods (1965 to 1981) when individual names 

were not recorded with dosimeter numbers, and many records cannot be clearly associated with 

individual workers [12]. For cases in which the individual’s name is clearly associated with dosimetry 

records and the records are essentially complete, these individuals are considered monitored 

workers. For all other cases in which the records are unidentified or partially complete, the individuals 

are considered unmonitored workers for the missing or incomplete periods. External doses received 

since 1982 have been reliably recorded for each individual. Details of the external dosimetry program 

are addressed in this section.

6.1 INTRODUCTION

Ames Laboratory responses to DOL requests for claimant records were limited to raw data from 

dosimetry files. Not all these files had been summarized for annual total doses for individuals and the 

files for some individuals were incomplete as noted above. To assist dose reconstructors in 

determining annual doses for claimants, the raw data were entered into spreadsheets and 

summarized to give annual totals for identified individuals; the spreadsheet data were analyzed to 

give 50th- and 95th-percentile values that could be applied to unidentified (unmonitored) individuals 

[ORAUT 2006c].

Dose reconstructors should have access to dosimetry records from Ames Laboratory for each 

claimant, but the records might be incomplete and difficult to use. The spreadsheets assembled by 

ORAUT [2006c] should be used if there is any doubt about completeness.

According to the feasibility table in Section 1.3, external dose can be adequately estimated for 

unmonitored workers beginning January 1, 1955. 

6.2 EXTERNAL RADIATION DOSIMETERS AND RECORDS

An AEC survey conducted at Ames Laboratory from March 18 to 21, 1952, found the personnel 

monitoring program inadequate. Several recommendations were made in the survey report to 

promote improvements in radiation protection including film badge service and regular radiation 

monitoring [Hokel et al. 1998]. Before this time, only pocket chambers (pencil dosimeters) were used 

to monitor radiation exposure and records were incomplete [Voss 1949].

Regular film badge service began in February 1953; beta/gamma film badges were exchanged on a 

biweekly frequency with the results tabulated monthly and summarized annually [Ames 1952–1954, 

1953–1954a,b]. Dosimetry services evolved and improved over the following 50 years; the 

characteristics of the dosimeters are summarized in Table 6-1.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 46 of 78

Table 6-1. Dosimeter type, period of use, exchange frequency, minimum recordable dose (MRD), and 

minimum detection level (MDL).

a

Dosimeter type, 

provider Period

Exchange

frequencyb

MRD 

skin 

(mrem)c 

MRD

β/γ

deep

(mrem)c

MRD 

neutron

(mrem)c 

MDL

skin 

(mrem)d

MDL

deep 

(mrem)d

MDL 

neutron 

(mrem)d

Pocket chambers, 

Ames Laboratory

1942–1952 Daily N/A 5 N/A N/A 5 N/A

β/γ film, Ames 

Laboratory in-house 

system

02/1953–05/1953

06/1953–12/1953

01/1954–02/1957

03/1957–12/1961

Biweekly N/A 10 N/A

N/A 25 N/A

25 25 N/A

10 10 N/A

40 40 N/A

40 40 N/A

40 40 N/A

40 40 N/A

β/γ/NTA film, BNL 04/1954–06/1957 Biweekly 15 15 10 40 40 50

β/γ/NTA film, NCA 07/1957–06/1963 Biweekly 10 10 10 40 40 50

β/γ/NTA film, 

Atomic Film Badge 

Corporation

07/1963–03/1965 Monthly 10 10 10 40 40 50

Pocket chambers, 

Ames Laboratory

04/1965–06/1965 Daily N/A 5 N/A N/A 5 N/A

β/γ/NTA film, HPS 04/1965–06/1979 Monthly 10 5 28 40 40 50

β/γ film, HPS 11/1979–09/1981 Monthly 10 5 N/A 40 40 N/A

β/γ TLDs, Landauer 06/1980–12/1981 Monthly and 

quarterly

40 10 N/A 30 30 N/A

β/γ TLDs, Landauer 01/1982–12/1994 Quarterly 40 10 N/A 30 30 N/A

β/γ TLDs, Siemens 01/1995–06/1996 Quarterly 10 10 N/A 30 30 N/A

β/γ TLDs, ICN 

Dosimetry Service

01/1996–09/1998 Quarterly 10 10 N/A 30 30 N/A

β/γ TLDs, Landauer 10/1998–12/2004 Quarterly 40 10 N/A 30 30 N/A

β/γ TLDs, Global 

Dosimetry

01/2005–present Quarterly 10 10 N/A 30 30 N/A

a. BNL = Brookhaven National Laboratory; HPS = Health Physics Services; NCA = Nucleonic Corporation of America; 

NTA = nuclear track emulsion, type A; N/A = not applicable. 

b. The exchange frequency was established from dosimetry reports.

c. Based on minimum doses recorded on dosimetry reports.

d. Estimated MDL typical of film dosimeter capabilities [Wilson 1960, 1987; Lalos 1989; Cardarelli et al. 1993; Wilson et al. 

1990]. 

6.2.1 Historical Administrative Practices

Some of the early administrative practices at Ames Laboratory related to dosimetry recordkeeping 

were unusual and inconsistent with current practice. The recordkeeping practices that correspond to 

each step in the evolution of dosimetry services (Table 6-1) are addressed in this section. 

Pocket chambers (pencil dosimeters) were used at Ames Laboratory at various times, possibly as 

early as 1942, until 1952 when film badge service began. However, the records of pocket chamber 

results that were found apply only to a brief period in February 1949 and only for a few individuals 

[Voss 1949]. 

An in-house film badge system was established at Ames in fall 1952 in response to recommendations 

from the AEC. Weekly film badges were provided to a few individuals who worked at the synchrotron 

from September 1952 to March 1953, but records are incomplete [Ames 1952–1954, 1953–1954a,b]. 

Regular film badge service began in February 1953 for all laboratory staff. Beta/gamma film badges 

were exchanged every 2 weeks and results were summarized monthly. However, between late 

September 1953 and early January 1954, there were three 4-week and one 3-week exchange periods 

[Ames 1952–1954, 1953–1954a,b]. The in-house film badge system had variable exchange periods 

during 1954, 1957, 1958, 1959, and 1961. There was a 4-week exchange period in January, a 3-week 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 47 of 78

period in February, and a 4-week period in December 1954; otherwise, the exchange periods were 2 

weeks each [Ames 1954, 1954–1955a,b]. The exchange frequency during 1955 and 1956 was every 

2 weeks with no exceptions [Ames 1954–1955a,b, 1955, 1955–1956, 1956]. There was a 4-week 

exchange period in May and June 1957; all other exchange periods in 1957 were 2 weeks [Ames 

1957a,b,c]. There was a 3-week exchange period in December 1958 and January 1959; all other 

exchange periods in 1958 and 1959 were 2 weeks [Ames 1957–1959, 1958–1959, 1959–1960a,b]. 

The exchange frequency during 1960 was every 2 weeks without exception [Ames 1960b]. During 

1961, there were 9 3-week exchange periods and 12 2-week periods [Ames 1961b].

Initially, the MRD was 10 mR, and no distinction was made between beta and gamma doses 

(recorded nonzero doses were assumed to be gamma doses). As of June 1953, the MRD was 

reported as 25 mR. In 1954, the MRD was specified as 25 mrep and 25 mR for beta and gamma, 

respectively. In March 1957, the MRD decreased to 10 mR for both beta and gamma doses; this MRD 

was unchanged through 1961 [Ames 1957a,b,c, 1957–1959, 1958–1959, 1959–1960a,b, 1960b, 

1961b].

Between October 24 and November 21, 1953, unusually high beta dose readings (about 700 mR) 

were “caused by the film being left unprotected near X-ray radiation” [Ames 1952–1954, 1953–

1954a,b]. Corrections that are favorable to claimants for this exposure that was not received by 

personnel are recommended in ORAUT [2006c]. 

In the fourth quarter of 1953 and continuing through 1961, beta and gamma doses were reported 

separately and were added to give a total dose [Ames 1952–1954, 1953–1954a,b]. Between 1954 

and 1961, neutron doses were added to the beta and gamma doses to give a total dose [Ames 1952–

1954, 1953–1954a,b, 1954, 1954–1955a,b, 1955, 1955–1956, 1956, 1957a,b,c, 1957–1959, 1958–

1959, 1959–1960a,b, 1960b, 1961b]. 

In the annual summaries for 1953 to 1956 (and for January and February 1957), an assumed dose of 

25-mR gamma was assigned for each month in which the dosimeter reading was zero or less than the 

MRD [Ames 1952–1954, 1953–1954a,b, 1954–1955a,b, 1955, 1955–1956, 1956]. From March 1957 

to 1961, when the dosimeter readings were zero or less than the MRD, an assumed dose of 10-mR 

gamma was assigned for each dosimeter exchange period [Ames 1957a,b,c, 1957–1959, 1958–1959, 

1959–1960a,b, 1960b, 1961b]. This practice of assigning doses equal to the MRD is roughly 

equivalent to the standard method for correction of missed dose for monitored workers [NIOSH 

2007b]. Therefore, no additional correction to the gamma dose is required for monitored workers for 

the period from 1953 to 1961.

Beginning in April 1954 and continuing through June 1957, a beta/gamma/neutron film badge service 

was provided by Brookhaven National Laboratory (BNL) for the synchrotron staff [Ames 1954, 1954–

1955a,b]. The film badges were exchanged every 2 weeks and results were included in the annual 

dose summaries. The MRD was 15 mrep and 15 mR for beta and gamma, respectively. Neutron 

exposure was reported as the number of recoil proton tracks in the open window, and the number of 

tracks was multiplied by 10 to give the neutron dose in millirem [Ames 1954, 1954–1955a,b, 1961b]. 

Between July 1957 and June 1963, beta/gamma/neutron film badge service was provided by the 

Nucleonic Corporation of America (NCA) for the synchrotron staff [Ames 1957a,b,c]. The film badges 

were exchanged every 2 weeks and results were included in the annual dose summaries. The MRDs 

were 10 mrad and 10 mrem for beta and gamma, respectively. Neutron exposure was reported as the 

number of recoil proton tracks in the open window, and the number of tracks was multiplied by 10 to 

give the neutron dose in millirem [Ames 1957a,b,c]. 

The data discussed above for 1953 to 1961 were compiled on a spreadsheet, and the inconsistencies 

were corrected or minimized [ORAUT 2006c]. For example, if beta, gamma, and neutron doses were 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 48 of 78

reported separately but added together for a total dose, the separated beta, gamma, and neutron 

doses were entered in the spreadsheet to facilitate analysis. For cases in which data were 

questionable, assumptions were made that are favorable to claimants [13]. 

Dosimetry recordkeeping from January 1962 to June 1963 was less than adequate. The only records 

found for 1962 were some calibration data from April and the annual summary data [Ames 1962b, 

1962–1964]. It was assumed that the in-house film badge system continued through December 1962 

with badges exchanged on a biweekly frequency, but no records could be found to confirm this 

assumption. Monthly film badge service was provided for Ames Laboratory personnel with the 

potential for radiation exposure by the Atomic Film Badge Corporation from July 1963 to March 1965.

No records were found for the first half of 1963. The annual summary for 1963 reflects results from 

July to December 1963 only. Therefore, all Ames Laboratory workers are assumed to have been 

unmonitored for the first half of 1963 [Ames 1963]. The annual summary for 1964 accurately reflected 

the sum of the monthly dosimeter readings during 1964 [Ames 1964]. 

Dosimetry records for January to March 1965 appeared to be reliable; however, the Atomic Film 

Badge Corporation defaulted on its contract and went out of business in April 1965 [Mattmueller 

1965]. Ames Laboratory staff were monitored with pocket chambers on a daily basis and records were

compiled in house from April to June 1965 [Ames 1965]. Film badge service was provided by Health 

Physics Services (HPS) from July 1965 to September 1981. The monthly service included beta, 

gamma, and neutron film badges for reactor and synchrotron staff and beta/gamma film badges for 

the remaining laboratory personnel [Ames 1965]. The MRDs were 10 mrad for beta, 5 mR for gamma, 

and 28 mrem for neutrons [Ames 1965]. 

TLD services for beta and gamma dosimetry were provided by Landauer from June 1980 to June 

1995. Similar services have been provided in subsequent years by Siemens, ICN Dosimetry Service

(ICN), and Landauer. The MRD for TLD services was 10 mrem for beta and gamma, except Landauer 

specified an MRD of 40 mrem for hard beta (greater than 1.5 MeV) [Landauer 1982–1983; Siemens 

1995; ICN 1996]. Dosimetry records compiled since 1981 are considered complete and reliable.

Records for monitored workers include an identification number, name, Social Security Number, and 

recorded doses for the current period, calendar quarter, and calendar year. The TLD services used by 

Ames Laboratory have been accredited by the National Voluntary Laboratory Accreditation Program 

(NVLAP) since 1985 [Landauer 1985–1986]. 

6.2.2 Dosimetry Technology

Ames Laboratory used film badge dosimeters from 1952 to 1981; TLDs have been used since 1980.

The initial beta/gamma film badge system was operated in house from 1952 to 1962. The film was 

Kodak type K and the developer was Kodak D-19. However, no records could be found that described 

the film holder, although it seems likely that one of several commercially available holders would have 

been procured and that such a holder would have had standard design features as described in AEC 

[1955]. 

The initial neutron dosimetry service at Ames Laboratory, which began in 1954 and included beta and 

gamma dosimetry, was provided by BNL. The film badge holder was the basic Oak Ridge National 

Laboratory multielement dosimeter; the neutron film was nuclear track emulsion, type A (NTA) 

[ORAUT 2006d]. Similar film badge holders and film were provided by NCA, Atomic Film Badge 

Corporation, and HPS for the periods listed in Table 6-1.

With the termination of operations at the synchrotron in 1971 and the removal of fuel from the ALRR 

in October 1979, neutron dosimetry services were no longer needed after 1979. From 1980 to the 

present, beta/gamma TLD service has been provided by Siemens, ICN, and Landauer for the periods 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 49 of 78

listed in Table 6-1. These services have been accredited by NVLAP since 1985 [Landauer 1985–

1986]. 

6.2.2.1 Beta/Photon Dosimeters

Figure 6-1 shows the response of a film badge to photon radiation of different energies; it also shows 

the Hp(10) response. The figure shows two responses for film badges: one for a sensitive DuPont 502 

emulsion in a two-element badge [Pardue et al. 1944] and one for a sensitive DuPont 555 emulsion in 

the multielement badge [Thornton et al. 1961]. The response of the sensitive Eastman Type 2 film in a 

multielement film badge is similar to that of the sensitive DuPont 555 emulsion. The film badges show 

an overresponse at photon energies around 100 keV, due primarily to relatively (compared with 

tissue) high atomic numbers (silver [47] and bromine [35]) in the film emulsions. The two-element film 

badges under respond to lower energy photons; the multielement film badge typically over responds

to photons between 50 and 150 keV.

Figure 6-1. Comparison of Hp(10) for photons with energy 

responses for sensitive DuPont 502 emulsion in two-element film 

badge [Pardue et al. 1944] and sensitive DuPont 555 emulsion in 

multielement film badge [Thornton et al. 1961].

6.2.2.2 Neutron Dosimeters

NTA film was used by the dosimetry services provided to Ames Laboratory from 1954 to 1979. In 

general, the response of the NTA film decreases with decreasing neutron energies greater than a 

minimum threshold energy estimated to be between 500 and 800 keV [Griffith et al. 1990; ORAUT 

2006d]. An unknown fraction of the total neutron dose was received from neutrons with energies less 

than about 800 keV. 

6.2.3 Dosimeter Calibration Procedures

Calibration procedures for beta/photon dosimeters were followed consistently from 1953 to 1981 with 

the exception of 1962 and the first half of 1963 [14]. The suppliers of neutron dosimeters were 

generally relied on for neutron calibrations.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 50 of 78

6.2.3.1 Beta/Photon Dosimeters

A set of calibration control film badges was exposed and processed with each biweekly set of films

used in the Ames Laboratory in-house film badge system from 1953 to 1961 [Ames 1974]. The beta 

calibration was accomplished by exposing films for varying times to a milled uranium plate to produce 

beta doses between 10 and 1,000 mR. The gamma calibration was done by exposing films to doses 

of gamma radiation from a 10-mg 226Ra source between 10 and 1,000 mR [Ames 1954, 1954–

1955a,b]. The interpretation of film density into exposure units was made from curves drawn for the 

calibration films [Ames 1974]. 

Between January 1962 and June 1963, there was evidence of calibrations done only during April 

1962 [Ames 1962–1964]. From June 1963 to March 1965, when the Atomic Film Badge Corporation 

provided film badge services, monthly gamma calibrations were done by exposing films to doses of 

gamma radiation from a 10-mg 226Ra source from 100 to 10,000 mR [Ames 1963, 1965]. There was 

no evidence of beta or neutron calibrations during this period. 

From July 1965 to September 1981, when film badge services were provided by HPS, a set of 

calibration control film badges were exposed to beta and gamma sources with each biweekly set of 

personnel film badges [Ames 1965; Voss 1971, 1975, 1979a,b, 1982]. The calibration procedures 

were the same as those used from 1953 to 1961.

Between 1980 and 1984, when TLD services were provided by Landauer, there is no evidence of any 

dosimeter calibrations by Ames Laboratory. After 1985, NVLAP accreditation was relied on for 

dosimeter calibrations.

6.2.3.2 Neutron Dosimeters

From April 1954 to June 1957, when film badge services were provided by BNL, a set of four 

calibration badges was included in each monthly set of films [Ames 1954, 1954–1955a,b, 1955, 

1955–1956, 1956]. The calibration badges were exposed to known doses of beta, gamma, and 

neutron radiation. The beta exposures were to a milled uranium plate, and the gamma exposures 

were to a 10-mg 226Ra source, but records of the source of neutron exposures were not found [Ames 

1954, 1954–1955a,b, 1955, 1955–1956, 1956]. 

From July 1957 to June 1963, when NCA provided film badge services, a set of area monitor film 

badges was included with each month’s set of personnel dosimeters [Ames 1957a,b,c, 1957–1959, 

1958–1959, 1959–1960a,b, 1960b, 1961b, 1962–1964, 1963]. The monthly readings on area monitor 

film badges were used to check on the validity of personnel dosimeter readings, but actual 

calibrations were not performed by Ames Laboratory. NCA calibrated its NTA film badges by exposing 

films for 40 hours to a fast neutron flux of 18 n/cm2

/s (2.6 × 106 n/cm2/40 hr), which produced 

approximately 10 recoil proton tracks in 25 fields or an exposure equivalent to 100 mrem [Ames 

1960b]. From June 1963 to March 1965, when the Atomic Film Badge Corporation provided film 

badge services, there was no evidence of neutron calibrations. 

Between July 1965 and September 1981, when HPS provided film badge services, a set of calibration 

control film badges were exposed to beta and gamma sources with each biweekly set of films [Ames 

1974]. However, no evidence could be found that neutron exposures were included in this calibration 

procedure.

6.2.4 Workplace Radiation Fields

Workplace radiation fields at Ames Laboratory arose from the production of uranium and thorium 

metal, the operation of the synchrotron and ALRR, and research activities in laboratories. With few 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 51 of 78

exceptions, the following sections show that, for external dose reconstruction, all beta radiation fields 

were greater than 15 keV, all photon radiation fields were between 30 and 250 keV, and all neutron 

fields were between 0.1 and 2 MeV [ORAUT 2021]. Assuming that 100% of the radiation fields were 

within these ranges is a simplifying conservative assumption that is favorable to the claimant. 

6.2.4.1 Beta Radiation

The major sources of beta radiation at Ames Laboratory between 1942 and 1954 were the large 

quantities of uranium and thorium metal produced. The uranium handled is assumed to have been in 

equilibrium with its 234Th (24.1-day half-life) and 234mPa decay products [DOE 2001, p. 2–17]. From an 

external dose standpoint, the most significant beta radiation emitted from uranium metal is that from 

234mPa, with a maximum energy of 2.29 MeV [Shleien et al. 1998]. The beta dose rate (with 7 mg/cm2 

filtration) is 233 and 179 mrad/hr at the surface of a uranium slab and UF4, respectively [BRH 1970]. 

Thorium, if recently separated from its 228Ra and subsequent decay products, would have very little 

beta emission. However, if ThF4 was stored for several years before being processed to thorium 

metal, beta radiation from 228Ra and 228Ac would be measurable [Shleien et al. 1998]. 

A large number of beta-emitting radionuclides were handled in research work at Ames Laboratory 

throughout its history [15]. However, the activities of these sources were relatively small; the sources

had a much smaller potential for beta radiation exposure to workers in comparison with the beta dose 

from uranium metal. With the exception of tritium, the energy of all beta sources was assumed to be 

greater than 15 keV; beta doses would have been readily measured in the open-window portion of the 

film badges used at the Laboratory. Tritium exposures at the ALRR were monitored by urine bioassay, 

which is addressed in Section 5.0 of this site profile.

6.2.4.2 Photon Radiation

Bremsstrahlung radiation doses from uranium metal ingots were calculated by Anderson and Hertel 

[2005]. Contact photon dose rates of a few milliroentgens per hour were calculated for typical plates, 

billets, and cylindrical ingots of the type produced at Ames Laboratory. Another source of photon 

radiation in the early years was stored thorium in the form of ThF4. A gamma dose rate of 22 mR/hr in 

a thorium storage area was reported by Klevin [1952]. Exposures to photon radiation were possible in 

the experimental areas at the synchrotron and ALRR, and near the many gamma-emitting 

radionuclides that were handled in research work. A hot laboratory was located in the Chemistry 

Building, and a hot canyon/hot cell was located in the Research Building for work with more intense 

gamma sources. Photon sources had energies in the 30-to-250-keV and greater-than-250-keV 

ranges. X-ray diffraction machines could have operated at energies less than 30 kV, and measured 

exposures probably would have been reported as beta dose. Limited data do not permit an accurate 

estimate of the fraction of photon exposures expected in these energy ranges. Assuming that 100% of 

the photons were in the 30-to-250-keV range is a simplifying conservative assumption that is 

favorable to claimants [ORAUT 2021]. Photon radiation in the workplace could have been readily 

measured at Ames Laboratory with available dosimeter technology since 1953. 

6.2.4.3 Neutron Radiation

The work involving plutonium and the research associated with thorium and uranium in the 

laboratories had the highest potential for exposure to neutrons.

The primary sources of neutrons at Ames Laboratory were the synchrotron, which operated from 1949 

to 1971, and the ALRR, which operated from 1965 to 1977. In both facilities, fast, intermediate, and 

thermal neutrons would have been present in the workplace [ORAUT 2021]. Unfortunately, no records 

could be found that characterized the neutron spectrum in either facility, and there were no records of 

neutron surveys around experimental areas or penetrations in the shielding. The only documented 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 52 of 78

evidence of neutrons in the workplace is the many nonzero fast neutron doses measured with NTA 

film worn by personnel at each facility [ORAUT 2006c]. Assuming that 100% of the neutrons were in 

the 0.1-to-2-MeV energy range is a simplifying conservative assumption that is favorable to claimants.

6.2.5 Dosimeter Response to Radiation Fields

The personnel dosimeters used at Ames Laboratory were properly selected for the radiation fields 

encountered in the workplace. The response of each of the dosimeters used is described in the 

following sections.

6.2.5.1 Beta/Photon Film Dosimeter Response

Ames Laboratory used film badges for beta/photon dosimetry from 1952 to 1981. An in-house film 

badge system was used from 1952 to 1962; film badges were provided by four commercial services 

from 1954 to 1981 [16]. The dosimeters provided an open window with little filtration, a lower energy 

window that allowed beta particles and lower energy photons to enter a film area with a plastic filter, 

and a film area with a metal (usually aluminum) filter. The open window enabled measurement of beta 

particles and lower energy photons. The plastic filter enabled measurement of intermediate energy 

photons, and the metal filter enabled measurement of higher energy photons (1-cm depth).

The beta/photon dosimeters were calibrated regularly by methods consistent with accepted practice, 

but there was no evidence of any formal dosimeter response testing by an independent entity. The 

recorded beta and gamma doses appear to be consistent with those expected from the source terms 

and personnel monitoring described in Section 2.3. 

6.2.5.2 Beta/Photon Thermoluminescent Dosimeter Response

Ames Laboratory has used TLDs for beta and photon dosimetry since 1980. The TLDs, their 

analyses, and recordkeeping have been provided by reliable commercial services. The dosimetry 

services have been accredited by NVLAP since 1985 [Landauer 1985–1986]. 

6.2.5.3 Neutron Dosimeter Response

The neutron doses received by synchrotron staff measured by NTA film appeared to be consistent 

with similar accelerator operations at other sites. The neutron doses received by ALRR personnel 

were generally higher than those experienced at other sites (e.g., the High Flux Beam Reactor at 

BNL) [ORAUT 2013]. Fast, intermediate, and thermal neutrons were present in both Ames Laboratory 

facilities, but no records were found that characterized the neutron energy spectrum at either facility.

Thus, there is an unknown fraction of the total neutron dose (due to neutrons with energies below 

about 800 keV) that was not measured by NTA film.

To correct the measured neutron doses for the unmeasured fraction of neutrons with energies below 

800 keV, data from similar facilities were examined. The 200-MeV electron synchrotron at SLAC was 

determined to be similar to the 80-MeV synchrotron at Ames Laboratory in relation to neutron 

exposures [ORAUT 2007]. The Materials Test Reactor (MTR) at the Idaho National Laboratory was 

determined to be similar to the ALRR with respect to neutron exposures [ORAUT 2010]. Neutron 

dosimeters with NTA film were used at both facilities and neutron spectral data are available to

estimate the correction factor for missed neutron dose. 

Accelerator operations at SLAC started in 1966; a multielement dosimeter was used until 1971 for 

beta/gamma and neutron (NTA film) dosimetry. ORAUT [2007] concluded that all SLAC neutron 

doses measured with NTA film should be multiplied by 1.53 ± 0.14 to account for the unmeasured 

neutrons with energies below 800 keV. The SLAC and Ames Laboratory synchrotrons were similar, 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 53 of 78

but their neutron spectra were not identical [17]. To allow for differences in the neutron spectra and 

other variables, a neutron dose correction factor of 2 is recommended for Ames synchrotron workers 

[18]. 

The MTR at Idaho National Laboratory, which operated from 1952 to 1970, was similar to the ALRR in 

that it was fueled with enriched uranium and had a power level of 30 MW [ORAUT 2010]. The MTR 

was water-cooled and light-water moderated, whereas the ALRR was heavy-water moderated. Both 

reactors had ports where fast neutron beams could be extracted from the core and directed to 

accessible experimental areas. Both were used to study reactor fuels and structural materials.

Neutron doses at both reactors were monitored with multielement dosimeters including NTA film. The 

neutron spectrum in experimental areas at the MTR was measured with Bonner spheres in 1961 

[Hankins 1961]. The data from these measurements were reanalyzed in ORAUT [2013]. 

Measurements made at 22 locations at the MTR were analyzed in relation to the response of NTA 

film. It was determined, on average for the 22 locations, that 52% of the total neutron dose would 

have been detected by NTA film and 48% would have been below the 800-keV threshold and 

undetected [ORAUT 2007]. Because the MTR and the ALRR had very similar neutron spectra, a 

neutron dose correction factor of 2 is recommended for ALRR workers.

6.2.5.4 Neutron Dose Weighting Factor

At Ames Laboratory, neutron dosimeter measurements were based on fluence-to-dose conversion 

factors and quality factors similar to those from International Commission on Radiological Protection

(ICRP) Publication 21 [ICRP 1973] and NCRP Report 38 [NCRP 1971]. It is necessary to adjust the 

neutron dose to account for the change in neutron quality factors between historical and current 

scientific guidance. ICRP Publication 60 corrections factors [ICRP 1991] should be applied to the 

neutron doses in accordance with ORAUT-OTIB-0055, Technical Basis for Conversion from NCRP 

Report 38 Neutron Quality Factors to ICRP Publication 60 Radiation Weighting Factors for Respective 

IREP Input Neutron Energy Ranges [ORAUT 2006e]. Table 6-2 shows the correction factor to use.

Table 6-2. Neutron dose energies, percentages, and associated ICRP [1991] correction factors. 

Process description

Neutron energy

(MeV)

Default dose 

fractiona 

(%)

ICRP [1991]/NCRP [1971] 

correction factor

Neutron exposures associated with 

synchrotron and research reactor activities

0.1–2 100 1.91

a. The assumption that all neutron energies are between 0.1 and 2 MeV is favorable to the claimant.

6.3 RECOMMENDATIONS FOR WORKER EXTERNAL DOSE RECONSTRUCTION

Dose reconstruction for Ames Laboratory workers is based on the above information, which requires 

that assessment of additional dose be added to the measured photon dose from three primary 

causes:

• Adjustments to measured photon dose for dosimeter uncertainty, 

• Adjustments to measured neutron dose using a correction factor to account for neutrons with 

energies less than 800 keV that were not measured by NTA film, and

• Multiplication of the adjusted neutron dose by an ICRP [1991] neutron weighting factor 

adjustment of 1.91 for neutron energies between 0.1 and 2 MeV. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 54 of 78

6.3.1 Unmonitored External Dose

At Ames Laboratory, “unmonitored workers” includes workers who were monitored but for whom 

records have not been found [19]. According to the feasibility table in Section 1.3, NIOSH is not able 

to calculate unmonitored doses before 1955. Co-exposure dose, based on a review of Ames 

Laboratory dosimetry data, was used to estimate doses beginning in 1955. 

6.3.1.1 Co-Exposure Doses, 1955 to Present

Extensive dosimetry records have been found for Ames Laboratory workers; however, many of the 

records for 1965 to 1981 do not identify the person receiving the radiation dose. If a worker was 

monitored but cannot be identified in the dosimetry records, that individual must be considered 

unmonitored and assigned a dose in each year for which no clearly identified records exist. 

A co-exposure data study was used for this site profile to permit dose reconstructors to complete 

certain cases for which external monitoring data were unavailable or incomplete. Co-exposure data 

are data from workers at a site (potentially grouped by work location, job description, or other 

appropriate category) whose measured doses are considered representative of those received by one 

or more workers with no individual monitoring data [ORAUT 2011]. 

The general approach to applying co-exposure data for cases with little or no individual external 

monitoring data is to assign the 50th- or 95th-percentile dose with the intent that the assigned doses 

represent, but do not underestimate, the doses that would have been assigned if the worker had been 

monitored (50th percentile) or if the monitored worker was clearly identified in the dosimetry records 

(95th percentile). Dose reconstructors should assign the GM (50th percentile) or the 95th-percentile 

values as constants [ORAUT 2011]. 

Some workers might have never been monitored during their employment at Ames Laboratory.

Workers with job titles such as security patrolman, craftsman, janitor, secretary, or clerk who did not 

work routinely in radiological areas were probably not monitored. External environmental doses 

should be assigned to those workers who were unlikely to be exposed to radiation. The 

50th-percentile dose can be used to estimate doses for those workers when professional judgment 

indicates the worker was likely to have been exposed to intermittent low levels of external radiation 

[ORAUT 2011]. 

Some workers with job titles such as scientist, chemist, metallurgist, engineer, technician, or machinist 

were probably monitored, but some or all of their dosimetry records might be missing because of the 

lack of clearly identified dosimetry records between 1965 and 1981. If any part of a worker’s dosimetry 

record is missing (unidentified), co-exposure dose should be assigned for the years for which records 

are missing [ORAUT 2011]. The 95th-percentile dose should be assigned to those workers who could 

have been regularly exposed.

Other options are available through the guidance in ORAUT-OTIB-0020, Use of Coworker Dosimetry 

Data for External Dose Assignment [ORAUT 2011] and DCAS-IG-006, Criteria for the Evaluation and 

Use of Coworker Datasets [NIOSH 2020]. For instance, for cases in which routine monitoring data 

exist and co-exposure dose is used to supplement missing quarters or years, assign the percentile 

dose that is consistent with the recorded dose unless there is reason to believe that the worker’s job 

or location differs significantly from that held during the year in which the dose was recorded. Note 

that the period in which dosimetry records might be missing due to the lack of clearly identified 

dosimetry records (between 1965 and 1981) is also the period of the most varying co-exposure 

doses. Use judgment and other information in the claim files to determine whether 50th- or 95thpercentile doses are more appropriate [NIOSH 2020]. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 55 of 78

The co-exposure study for Ames Laboratory included all available dosimetry records from 1952 to

1981 [ORAUT 2006c]. All dose results were analyzed, including zeros and blank values, to determine

the 50th- and 95th-percentile doses for each year for beta, gamma, and neutron exposures 

[McCartney 2006]. According to the feasibility table in Section 1.3, the co-exposure doses are 

assigned beginning in 1955. The results of the analysis are summarized in Table 6-3. The missed 

dose recommended for monitored workers in Table 6-4 was added to the 50th- and 95th-percentile 

values in Table 6-3. Specifically, half of the maximum annual missed doses were added to the 

reported annual doses except the reported positive doses, in which case the maximum missed dose 

was reduced by the dose corresponding to one badge exchange (because it is not possible that all 

individual badge results were zero if a positive annual dose was reported) [ORAUT 2011]. For beta 

doses, the maximum missed dose was not added to ensure missed dose is not assigned twice when 

evaluating dose to shallow dose organs [ORAUT 2005]. 

Table 6-3. Assigned dose for unmonitored workers (mrem/yr).a

Year

Beta

50th percentileb

Beta

95th percentileb

Gamma

50th percentile

Gamma

95th percentile

Neutron

50th percentilec

Neutron

95th percentilec

1955 0 0 520 520d 650 650

1956 0 0 520 520d 650 650

1957 0 0 520 520d 650 650

1958 0b 0b 520 520d 650 650

1959 0b 0b 520 520d 650 650

1960 0 0 520 520d 650 650

1961 0b 0b 520 520d 650 650

1962 0 0 520 520e 650 650e

1963 0 0 240 265f 650 650

1964 0 0 240 313f 650 650

1965 0 0 240 408f 650 820f

1966 0 86f 240 565f 806g 988f

1967 0 246f 240 745f 932g 1,128f

1968 0 87f 240 769f 876g 1,086f

1969 0 175f 240 1,182f 1,030g 1,296f

1970 0 0 240 671f 680g 932f

1971 0 0 240 809f 722g 950f

1972 0 0 240 512f 652g 988f

1973 0 0 240 701f 650 932f

1974 0 0 240 721f 650 876f

1975 0 0 240 1,298f 650 870f

1976 0 0 240 497f 650 958f

1977 0 0 240 752f 650 904f

1978 0 0 240 722f 650 652f

1979 0 0 240 819f 650 650

1980 0 0 240 1,595f Not applicable Not applicable

1981 0b 0b 240 240 Not applicable Not applicable

1982 0b 0b 85 555 Not applicable Not applicable

1983 325 510 105 134 Not applicable Not applicable

1984 0 0 60 85 Not applicable Not applicable

1985 0 0 60 125 Not applicable Not applicable

1986 0 0 60 60 Not applicable Not applicable

1987 0 0 60 114 Not applicable Not applicable

1988 0 310 72 149 Not applicable Not applicable

1989 0 0 60 81 Not applicable Not applicable

1990 0 9 60 91 Not applicable Not applicable

1991 0 51 60 85 Not applicable Not applicable

1992 0 30 60 71 Not applicable Not applicable

1993 0 22 60 118 Not applicable Not applicable
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 56 of 78

Year

Beta

50th percentileb

Beta

95th percentileb

Gamma

50th percentile

Gamma

95th percentile

Neutron

50th percentilec

Neutron

95th percentilec

1994 0 20 62 138 Not applicable Not applicable

1995 0 70 60 68 Not applicable Not applicable

1996 0 0 60 65 Not applicable Not applicable

1997 0 0 60 60 Not applicable Not applicable

1998 0 0 205 245 Not applicable Not applicable

1999 0 115 60 160 Not applicable Not applicable

2000 0 150 60 110 Not applicable Not applicable

2001 0b 0b 60 76 Not applicable Not applicable

2002 0 49 60 68 Not applicable Not applicable

2003 0 150 201 394 Not applicable Not applicable

2004 0b 0b 63 118 Not applicable Not applicable

2005 0 204 60 123 Not applicable Not applicable

a. Sources: Values from McCartney [2006]; data from Ames [1952–1954, 1953–1954a,b to 1957–1959, 1958–1959, 1959–

1960a,b, 1960b, 1961b, 1962–1964 to 1965], HPS [1966 to 1972], Ames [1973], HPS [1974], Ames/HPS [1975, 1976, 

1976–1977, 1978–1981, 1979–1981], Ames/Landauer [1980], HPS [1981], Landauer [1982–1983 to 1994], Siemens 

[1995], ICN [1996], Hokel [1998], Landauer [1999a to 2000], Simpson [2001, p. 2], Beckel [2001a to 2002], Landauer 

[2003, 2004], Beckel [2004a to 2005], Global Dosimetry Solutions [2006]. 

b. No beta doses were measured or reported in 1952, 1958, 1959, 1961, 1981, 1982, 2001, or 2004.

c. Neutron dosimeters were not assigned after 1979.

d. As discussed in Section 6.2.1, the practice of assigning doses equal to the MRD is roughly equivalent to the standard 

method for correction for missed dose for monitored workers [NIOSH 2007b]. Therefore, no additional correction to 

gamma doses is required for monitored workers for the period from 1955 to 1961. However, this method resulted in 

adjusted doses that were less than the missed dose assigned to unmonitored workers (50th-percentile value), so the 

higher missed dose was assigned.

e. The dosimeter exchange frequency in 1962 is uncertain; therefore, the missed dose for a biweekly frequency was 

assumed to be favorable to claimants.

f. The 95th-percentile value was assumed to be received all in 1 month; missed dose was added for the other 11 months. 

g. The 50th-percentile value was assumed to be received all in 1 month; missed dose was added for the other 11 months. 

Table 6-4. Potential missed dose for monitored workers (mrem).

Dosimeter Period

Exchange 

frequencya

MDL 

skin

MDL 

deep MDL neutron

Film badge 

β, γ, n

1952–1962 Biweekly 40b 40b 50b

Film badge 

β, γ, n

1963–1981 Monthly 40b 40b 50b,c

TLD β, γ 1981–present Quarterly 30 30 Not applicablec

a. Exchange frequencies were established from dosimetry reports.

b. Estimated MDL typical of film dosimeter capabilities [Wilson 1960, 1987; Cardarelli et al. 1993; 

Lalos 1989; Wilson et al. 1990].

c. Neutron dosimeters were not assigned after 1979 due to the removal of fuel from the ALRR in 

October 1979.

6.3.2 Missed External Dose for Monitored Workers

If external dose data are found in a worker’s file, dose reconstructors should assign a missed photon 

dose based on the MDL/2 method and the number of exchange periods [NIOSH 2007b] listed in 

Table 6-4 for the dosimetry systems. These missed doses are included in the values in Table 6-3. 

A flow chart is provided in Figure 6-2 to guide dose reconstructors in applying the above tables and 

recommendations. The Monitored Worker side of the figure should be used if there are record files for 

the entire employment period. If data for any exposure period are missing, the Unmonitored Worker 

side of the figure should be used for the missing period only. If there are no dose record files for a 

worker, the Unmonitored Worker side of the figure should be used for the entire employment period. 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 57 of 78

Figure 6-2. Flowchart for monitored and unmonitored workers. Attachment A contains an extended 

description of this image.

6.3.3 Uncertainty in Photon Doses

For the usual analysis of measured film badge doses, MDLs in the literature range from about 30 to 

50 mrem for beta/photon irradiation [Wilson et al. 1990]. It is possible to read a photon dose of 

100 mrem to within ±15 mrem if the exposure involved photons with energies between several keV 

and several MeV [Morgan 1961]. The estimated standard error in recorded film badge doses from 

photons of any energy is ±30% [ORAUT 2006f]. The estimated uncertainty in doses recorded by TLDs 

is ±20% from 1982 to 1984 and ±10% since 1985 with NVLAP accreditation [Landauer 1985–1986].
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 58 of 78

6.3.4 Skin Dose

Before 1981, the beta and skin dose records included beta doses only; that is, only nonpenetrating 

beta dose is recorded. In 1981 and subsequent years, the beta and skin dose has been calculated as 

the sum of the whole-body penetrating dose and the nonpenetrating dose [Landauer 1985–1986]. For 

cases in which no nonpenetrating dose was recorded, the skin dose is assumed to be equal to the 

whole-body penetrating dose. Guidance on determining skin dose can be obtained from ORAUTOTIB-0017, Interpretation of Dosimetry Data for Assignment of Shallow Dose [ORAUT 2005].

6.3.5 Extremity Dose

Assignment to and use of extremity dosimeters by Ames Laboratory workers has been inconsistent. In 

some years, there are no reported extremity dose results; in many years, there are only a few results, 

which suggests a less-than-rigorous extremity dosimetry program given the types and quantities of 

radioactive materials handled. Results are so sparse that it must be concluded that extremity doses 

were essentially unmonitored. 

6.3.6 Radiation Dose Fraction

Section 6.2.4 addresses the recommended energy ranges and fractions for Ames Laboratory dose 

according to facilities, processes, or activities as required by the IREP.

6.4 ORGAN DOSE

Organ doses are calculated using the external dose reconstruction implementation guidelines for 

irradiation geometry and dose conversion factors [NIOSH 2007b]. For photons measured with film 

badges and early TLD use (before 1985), use the conversion factor from exposure to organ dose.

Beginning in 1985, when NVLAP accreditation was obtained, deep dose equivalent should be used.

For neutron doses, use the conversion factors from deep dose equivalent to organ dose for anteriorposterior irradiation from NIOSH [2007b]. 

7.0 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction.

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item.

[1] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022. 

Normally, the accelerator would have been operated at reduced beam currents and with 

targets that produced much lower radiation dose rates on and off site. This was verified by 

examining the dosimetry records of the synchrotron staff, which showed no unusually high 

exposures during the May 1961 survey or at any other time.

Background reading in this area (synchrotron) before beam being turned on was 0.5 to 

1.0 mR/hr [Ames 1961a]. Apparently, the background reading was taken near the synchrotron 

where the radiation level was slightly elevated (due to activation products) above the normal 

natural background level. However, the exact location of the background reading cannot be 

determined. The fence line readings were taken some distance from the background reading 

location and should not have been influenced by the elevated background near the 

accelerator. In any event, the elevated background dose rates were not subtracted from the 

fence line dose rates reported in Table 4-1.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 59 of 78

[2] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

It is assumed that generally about one-third of the total staff time is spent on nonoperational 

activities, such as experiment setup, equipment startup, maintenance, repairs, etc. Thus, 

about two-thirds of the total time is available for operations.

[3] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

The radiation conditions during the fence line radiation survey on May 16, 1961, produced 

dose rates greater than 1 mR/hr at occupiable locations inside the Synchrotron Building. Fulltime occupancy in such a location would give an annual dose of greater than 2,000 mrem. A 

review of dosimetry records for synchrotron personnel indicated no one received such a dose 

during 1961 or at any other time. The average annual dose was less than 200 mrem. The 

average annual dose outside the Synchrotron Building would have been much lower.

[4] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

The air sample results were indistinguishable from background and were therefore not useful. 

Of the surface survey results in areas somewhat accessible to people, the highest result was 

1,224 dpm; however, there was no record of the surface area smeared. Surface activity 

measurements are typically reported in units of activity per 100 cm2

. Because of that 

uncertainty, the 1,224-dpm value was raised to 2,000 dpm/100 cm2 to represent an upper 

bound.

[5] Sharfi, Mutty M. ORAU Team. Principal Health Physicist. June 2022. 

Indoor resuspension factors range from 10-6 to 10-3 [ORAUT 2012]. The factor 10-4 was chosen 

for this application to represent an upper bound for chronic intake from surfaces not 

immediately accessible and, therefore, not often disturbed. 

[6] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

Hokel et al. [1998] indicated the thorium was assumed to be in equilibrium with its progeny.

The intake dose conversion factors account for buildup/decay of 232Th progeny in the body 

after the intake, but the activity of the progeny already in the air at intake has to be treated as 

separate intake radionuclides. Because of their short half-lives, 228Ac is included in dose 

calculations for 228Ra and all the progeny from 228Th are included in the dose calculations for 

228Th, so only 228Ra and 228Th must be entered as intakes separate from 232Th. 

[7] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

Review of Hokel et al. [1998] suggested there were areas with higher contamination, such as 

crawlspaces and vertical void spaces in rooms that were rarely, if ever, occupied. The 

statement in Section 5.2 that contamination was 10 to 100 times greater generalizes the 

discussion in Hokel et al. on this subject.

[8] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

Hokel et al. [1998] contains statements that rarely accessed areas that are still contaminated 

are carefully controlled, and that workers wear protective equipment and are monitored by 

health physics technicians; therefore, the 6-pCi/d chronic intake is an upper bound.

[9] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

MDAs or recording levels varied at the major AEC sites in the 1960s and 1970s from 1 μCi/L 

at the Hanford Site, Savannah River Site, and Los Alamos, to 0.02 μCi/L at Oak Ridge, to 

0.0002 μCi/L at Idaho National Laboratory. The Hanford and Savannah River values were 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 60 of 78

recording levels with actual laboratory MDAs generally being lower. Review of data recording 

methods at Ames Laboratory indicated that 0.1 μCi/L was the likely MDA or recording level. 

[10] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

The uranium and thorium metal production processes at Ames Laboratory established that 

workers were exposed to external radiation doses between 1942 and 1952. The only records 

of radiation dosimetry measurements during this period are Tybout [1944] and Voss [1949], 

which are very incomplete. Either the radiation doses were not measured or essentially no 

records were kept. In either case, the workers during this period were unmonitored.

[11] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

Voss [1949] recorded radiation exposures measured by pocket chambers for 15 individuals 

during a 2-week period in February 1949. These were the only pocket chamber dosimeter 

records found for the 1942-to-1952 period. The record suggests that pocket chambers were 

available during this period, but either they were not used or, if they were, records of the 

dosimeter readings were not kept.

[12] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

From 1965 to 1981, the film badge dosimetry records reflect badge numbers only. Rosters of 

names and badge numbers for bioassay records were found and it was possible to correlate 

some badge numbers and individual names. However, there were many badge numbers for 

which a name could not be assigned.

[13] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

As the spreadsheets were assembled, some addition and subtraction errors in the dosimetry 

records were noted. In all cases, the result that gave the highest dose was assumed and 

recorded in the spreadsheets.

[14] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

As stated in Section 6.2.1, recordkeeping from January 1962 to June 1963 was less than 

adequate. The only records found for 1962 were some calibration data from April 1962 and the 

annual summary data for 1962 [Ames 1962–1964]. It was assumed that the in-house film 

badge system continued through December 1962, with film badges exchanged on a biweekly 

frequency; however, no records could be found to confirm this assumption. Monthly film badge 

service was provided for all Ames Laboratory personnel with the potential for radiation 

exposure by the Atomic Film Badge Corporation from July 1963 to March 1965. No records 

were found for the first half of 1963. 

[15] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

The early history of research at Ames Laboratory was described by Fulmer [1947]; later history 

was described in Struss and Hawkins [1986]. These documents discussed numerous research 

projects that involved many beta-emitting radionuclides. The quantities used in research were 

much smaller than the quantities involved in production operations.

[16] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

Commercial film badge service was supplied by BNL from April 1954 to June 1957, by NCA

from July 1957 to June 1963, by Atomic Film Badge Corporation from July 1963 to March 

1965, and by HPS from June 1965 to September 1981 (see Table 6-1).

[17] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

The maximum beam energy at the Ames Laboratory synchrotron was 80 MeV; the maximum 

beam energy at the SLAC synchrotron was 200 MeV. Both accelerators produced neutrons, 
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 61 of 78

but the spectrum at SLAC included higher energy neutrons, so the neutron radiation fields 

were not identical. 

[18] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

The SLAC site profile [ORAUT 2007] used a neutron dose correction factor of 1.53 ±0.14 to 

account for unmeasured neutrons with energies below 800 keV. Although the neutron spectra 

of the Ames Laboratory and SLAC synchrotrons were not identical, they were similar enough 

to justify the assumption of a neutron dose correction factor of 2 for the Ames synchrotron. 

[19] Ellis, Jackson R. ORAU Team. Senior Health Physicist. July 2022.

From 1965 to 1981, the film badge dosimetry records reflect badge numbers only. Rosters of 

names and badge numbers for bioassay records were found and it was possible to correlate 

some badge numbers and individual names. However, there were many badge numbers for 

which a name could not be assigned. These records indicate that some workers were 

monitored, but they have to be considered unmonitored because the records for specific 

individuals cannot be retrieved.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 62 of 78

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GLOSSARY

absorption type

Categories for materials according to their rates of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively insoluble in the 

respiratory tract (slow solubilization).

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

bioassay

Measurement of amount or concentration of radionuclide material in the body (in vivo 

measurement) or in biological material excreted or removed from the body (in vitro 

measurement) and analyzed for purposes of estimating the quantity of radioactive material in 

the body. Also called radiobioassay.

bremsstrahlung

Electromagnetic radiation released as a result of inelastic scattering of a moving charged 

particle within the nucleus of an atom. X-rays produced in a typical medical X-ray tube 

frequently originate from inelastic scattering of accelerated electrons in the anode material.

contamination

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sieverts; other types of dose are in 

units of rad, rep, or grays.

dose equivalent (DE, H) 

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose. 

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See film dosimeter and thermoluminescent dosimeter. 

dosimetry

Measurement and calculation of internal and external radiation doses.
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dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

enriched uranium

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7% by mass. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade 

uranium contains greater than 90% 235U.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

film

(1) In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. (2) X-ray film.

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

fission product

(1) Radionuclides produced by fission or by the subsequent radioactive decay of 

radionuclides. (2) Fragments other than neutrons that result from the splitting of an atomic 

nucleus.

gamma radiation

Electromagnetic radiation (photons) that originates in atomic nuclei and accompanies many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are 

identical to X-ray photons; the difference is that X-rays do not originate in the nucleus.

hot cell

Shielded laboratory for handling of radioactive materials with the aid of remotely operated 

manipulators. The walls and windows are made of materials that protect workers from 

radiation.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds.

internal dose

Dose received from radioactive material in the body (e.g., plutonium or uranium) that was 

inhaled, ingested, absorbed, or injected through a wound.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 
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positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

beta radiation, gamma radiation, photon radiation, and X-ray radiation.

minimum detectable amount (MDA)

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error).

minimum detection level (MDL)

See minimum detectable amount.

missed dose

(1) In relation to external dose, dose to monitored workers that was not measured or recorded 

due to such factors as a missing or damaged dosimeter or a result below the detection limits of 

the dosimeter. Missed dose is especially important in the early years of radiation monitoring, 

when relatively high detection limits were combined with short exchange periods. (2) In 

relation to internal dose, potential dose that could have been received by a bioassay program 

participant but, because of limitations in the monitoring system, was undetected.

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements.

natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcuries 

per gram). See uranium.

neutron (n)

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

nuclear track emulsion, type A (NTA)

Film made by the Eastman Kodak Company that is sensitive to fast neutrons. The developed 

image has tracks caused by neutrons that become visible under oil immersion with about 

1,000-power magnification. The number of tracks in a given area is a measure of the dose 

from that radiation.

occupational dose

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

occupational medical dose

Dose from X-ray procedures performed for medical screening of workers as part of an 

occupational health program. Doses from X-rays used to diagnose diseases or injuries, even if 

incurred on the job, are not considered occupational and are therefore not eligible to be 
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included in dose reconstruction under the Energy Employees Occupational Illness 

Compensation Program Act of 2000.

personal dose equivalent [Hp(d)] 

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeters 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole body (deep) doses. These are noted 

as Hp(0.07) and Hp(10), respectively. In 1993 the International Commission on Radiological 

Measurement and Units recommended Hp(d) as the dose quantity for radiological protection.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.

progeny

Nuclides that result from decay of other nuclides. Also called decay products and formerly 

called daughter products.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation. 

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

roentgen (R)

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulomb per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0 degrees Celsius and standard atmospheric 

pressure. An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad 

in soft tissue for higher energy photons (generally greater than 100 kiloelectron-volts).

routine monitoring

Monitoring carried out at regular intervals during normal operations.

synchrotron

Roughly circular particle accelerator in which the particles travel in synchronized bunches at a 

fixed radius.
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thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose. Thermoluminescent dosimeters replaced film dosimeters at 

essentially all U.S. Department of Energy sites beginning in the 1960s.

unmonitored dose

Potential unrecorded dose that could have resulted because a worker was not monitored. See 

missed dose.

uranium (U)

Heavy, metallic, and radioactive element with atomic number 92. Most natural uranium as 

found in ores is 238U with trace levels of other isotopes. Uranium-235 (0.7% of natural uranium) 

is fissile by itself and used in nuclear weapons as well as reactors. Uranium-238 (99.3% of 

natural uranium) is fissionable by fast neutrons and used in nuclear reactors. Natural uranium 

contains a minute amount of 234U. See enriched uranium and natural uranium.

whole body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose.

X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
Document No. ORAUT-TKBS-0055 Revision No. 04 Effective Date: 09/18/2023 Page 78 of 78

ATTACHMENT A 

EXTENDED DESCRIPTION OF FIGURE 6-2 FLOWCHART

This attachment describes Figure 6-2 for accessibility by visually impaired readers. It is a flowchart to 

guide dose reconstructors in constructing doses for monitored or unmonitored workers. There are two 

pathways, one for monitored workers and one for unmonitored workers. 

The monitored workers pathway is on the left with five boxes indicating sequential steps. 

1. The first box instructs the reader to, for 1953 to present, tabulate beta, photon, and neutron 

doses of record for each year.

2. The second box instructs the reader to multiply beta/photon doses by uncertainty correction 

factor and contains a footnote (a) that gives the uncertainty correction factors as ±30% and 1.3 

for beta/gamma film badges, ±20% and 1.2 for TLDs before 1985, and ±10% and 1.1 for TLDs 

after NVLAP accreditation in 1985.

3. The third box instructs the reader to add adjustments for missed and zero doses after 1961, 

with a footnote (b) that indicates that adjustments for missed and zero doses have been made 

to the dose of record for years before 1962 and informs the reader that more information is in 

Section 6.2.1. 

4. The fourth box instructs the reader to multiply the neutron dose of record by correction factor 

of 2 and parenthetically states that this dose does not apply to missed dosed. A footnote (c) is 

included that states the correction factor is 2 for NTA film badges and instructs the reader to 

see Section 6.2.5.3. 

5. The fifth box instructs the reader to adjust the neutron dose using the ICRP [1991] weighting 

factor by multiplying by 1.91 with a footnote (d) that states that the neutron weighting factor 

adjustment is for neutron energies from 0.1 to 2 MeV.

The unmonitored workers pathway is on the right with three drop boxes indicating sequential steps. 

1. The first box instructs the reader to tabulate annual assigned doses from Table 6-3 and 

instructs the reader to see the text in Section 6.3.1 for explanation and guidance.

2. The second box instructs the reader to multiply the assigned neutron dose by a correction 

factor of 2. A footnote (c) is included that states the correction factor is 2 for NTA film badges 

and instructs the reader to see Section 6.2.5.3. 

3. The third box instructs the reader to adjust the neutron dose using the ICRP [1991] weighting 

factor by multiplying by 1.91 with a footnote (d) that states that the ICRP [1991] neutron 

weighting factor adjustment is for neutron energies from 0.1 to 2 MeV.

A note under the footnotes states that the photon and neutron organ dose conversion factors are not 

included in this flowchart and must be applied separately.