ORAU TEAM

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DOE Review Release 01/18/2012

Document Title:

Site Profile for Ames Laboratory

Document Number: ORAUT-TKBS-0055

Revision: 03

Effective Date: 01/03/2012

Type of Document: TBD

Supersedes: Revision 02

Subject Expert(s): Donald E. Bihl

Approval: Signature on File Approval Date: 01/02/2012

Jackson R. Ellis, Document Owner

Concurrence: Signature on File Concurrence Date: 01/03/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 12/29/2011

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 12/29/2011

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 01/03/2012

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 2 of 90

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.

Signature on File 08/12/2008

Karen S. Kent, Document Owner

Signature on File 08/23/2008

John M. Byrne, Task 3 Manager

Signature on File 08/12/2008

Edward F. Maher, Task 5 Manager

Signature on File 08/19/2008

Kate Kimpan, Project Director

Signature on File 08/20/2008

James W. Neton, Associate Director for Science

12/18/2009 01 Approved revision to add LAT 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 prior to 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 

prior to 1953. Table 5-1 was separated into three tables. Changes 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 3 of 90

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.

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.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 4 of 90

TABLE OF CONTENTS

SECTION TITLE PAGE

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

1.0 Introduction ............................................................................................................................10

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

1.2 Scope.........................................................................................................................11

1.3 Special Exposure Cohort............................................................................................11

1.3.1 August 13, 1942, through December 31, 1954............................................. 11

1.3.2 January 1, 1955, through December 31, 1970 ............................................. 12

1.3.3 January 1, 1955, through December 31, 1960 ............................................. 13

1.3.4 Class Recommended by NIOSH for Addition to the SEC: 

August 13, 1942, through December 31, 1970............................................. 13

2.0 Site Description......................................................................................................................15

2.1 Facilities .....................................................................................................................15

2.1.1 Physical Chemistry Annex 1......................................................................... 17

2.1.2 Chemistry Building (Gilman Hall).................................................................. 17

2.1.3 Physics Building (Physics Hall) .................................................................... 17

2.1.4 Physical Chemistry Annex 2......................................................................... 17

2.1.5 Metallurgy Building (Wilhelm Hall)................................................................ 18

2.1.6 Research Building (Spedding Hall)............................................................... 18

2.1.7 Office and Laboratory Building ..................................................................... 18

2.1.8 Synchrotron Building (Spangler Geotechnical Laboratory) ........................... 18

2.1.9 Metals Development Building....................................................................... 19

2.1.10 Ames Laboratory Research Reactor (Applied Science 

Complex) ..................................................................................................... 19

2.2 Operations..................................................................................................................19

2.2.1 Uranium Metal Operations ........................................................................... 19

2.2.2 Thorium Metal Operations............................................................................ 20

2.2.3 Other Operations.......................................................................................... 21

2.3 Source Terms and Personnel Monitoring....................................................................21

3.0 Occupational Medical Dose.................................................................................................... 23

4.0 Occupational Environmental Dose – Introduction................................................................... 24

4.1 Uranium/Thorium Production Period, 1942 to 1953 ....................................................24

4.2 Synchrotron Operations Period, 1949 to 1971............................................................24

4.3 ALRR Operations Period, 1965 to 1977......................................................................26

4.4 Ames Laboratory Research and Development Activities, 1942 to Present.................. 29

4.5 Significant Environmental Event .................................................................................31

4.6 Environmental Contamination, 1953 to Present..........................................................31

4.7 Summary of Environmental External Doses and Intakes ............................................ 32

5.0 Occupational Internal Dose.................................................................................................... 34

5.1 Uranium Exposure......................................................................................................34

5.1.1 Estimating Internal Doses from Uranium Metal ............................................ 34

5.1.1.1 Uranium Inhalation......................................................................... 35

5.1.1.2 Uranium Ingestion.......................................................................... 37

5.1.1.3 Resuspension During Periods with No Uranium 

Operations..................................................................................... 37
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5.1.1.4 Uranium Excretion Plots for Use with Bioassay Data..................... 38

5.2 Thorium Exposure from Thorium Contamination After 1954 ....................................... 39

5.3 Tritium Exposure ........................................................................................................42

5.3.1 Tritium Exposure for Monitored Workers, 1965 to 1981................................ 42

5.3.2 Tritium Dose to Workers, 1982 to Present.................................................... 46

5.3.3 Tritium Dose to Unmonitored Workers, 1965 Through 1981......................... 46

5.4 Fission Product Intakes ..............................................................................................47

5.4.1 Fission Product Intakes from Early Fuel Research....................................... 47

5.4.2 Fission Product Intakes from Research Reactor Operations 

and Decontamination and Decommissioning................................................ 48

5.5 Summary of Internal Dose Recommendations............................................................49

6.0 Occupational External Dose................................................................................................... 52

6.1 Introduction.................................................................................................................52

6.2 External Radiation Dosimeters and Records ..............................................................52

6.2.1 Historical Administrative Practices ............................................................... 52

6.2.2 Dosimetry Technology ................................................................................. 55

6.2.2.1 Beta/Photon Dosimeters................................................................ 56

6.2.2.2 Neutron Dosimeters....................................................................... 56

6.2.3 Dosimeter Calibration Procedures................................................................ 56

6.2.3.1 Beta/Photon Dosimeters................................................................ 57

6.2.3.2 Neutron Dosimeters....................................................................... 57

6.2.4 Workplace Radiation Fields.......................................................................... 57

6.2.4.1 Beta Radiation............................................................................... 58

6.2.4.2 Photon Radiation ........................................................................... 58

6.2.4.3 Neutron Radiation.......................................................................... 58

6.2.5 Dosimeter Response to Radiation Fields ..................................................... 59

6.2.5.1 Beta/Photon Film Dosimeter Response ......................................... 59

6.2.5.2 Beta/Photon TLD Response .......................................................... 59

6.2.5.3 Neutron Dosimeter Response........................................................ 59

6.2.5.4 Neutron Dose Weighting Factor..................................................... 60

6.3 Recommendations for Ames Laboratory Worker External Dose 

Reconstruction ...........................................................................................................60

6.3.1 Unmonitored External Dose ......................................................................... 61

6.3.1.1 Unmonitored Doses Before 1953................................................... 61

6.3.1.2 Coworker Doses – 1952 to the Present ......................................... 63

6.3.2 Missed External Dose for Monitored Workers .............................................. 65

6.3.3 Uncertainty in Photon Doses........................................................................ 65

6.3.4 Skin Dose .................................................................................................... 66

6.3.5 Extremity Dose............................................................................................. 67

6.3.6 Radiation Dose Fraction............................................................................... 67

6.4 Organ Dose................................................................................................................67

7.0 Attributions and Annotations .................................................................................................. 68

References ........................................................................................................................................74

Glossary ............................................................................................................................................84
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 6 of 90

LIST OF TABLES

TABLE TITLE PAGE

1-1 Feasibility findings for SEC-00038, January 1, 1942, through December 31, 1954 ................ 12

1-2 Feasibility findings for SEC-00075, January 1, 1955, through December 31, 1970 ................ 12

1-3 Feasibility findings for SEC-00166, January 1, 1955, through December 31, 1960 ................ 13

2-1 Timeline of AEC activities at Ames Laboratory, 1942 through 1954 ....................................... 20

4-1 Survey of fenceline gamma dose rates around the synchrotron facility with the 

beam directed west................................................................................................................ 26

4-2 Annual environmental fission product intakes from the hot laboratory in the 

Chemistry Building (1943–1981) ............................................................................................ 30

4-3 Environmental dose measurements at the four Ames facilities............................................... 31

4-4 Gamma analysis of soil samples............................................................................................ 32

4-5 Summary of 24 soil samples for beta and alpha activity ......................................................... 32

4-6 Environmental external doses ................................................................................................ 32

4-7 Summary of environmental intakes ........................................................................................ 33

5-1 Chemistry Building uranium intakes ....................................................................................... 35

5-2 Physical Chemistry Annex 1 uranium intakes......................................................................... 35

5-3 Physical Chemistry Annex 2 uranium intakes......................................................................... 35

5-4 Resuspension during periods with no uranium operations...................................................... 38

5-5 Air concentrations and daily intakes of 232Th at Wilhelm Hall after 1954................................. 40

5-6 Results of tritium coworker analysis ....................................................................................... 46

5-7 Default annual fission product intakes for researchers from hot laboratory in the 

Chemistry Building (1943–1951) ............................................................................................ 48

5-8 Default intakes for Ames Laboratory ...................................................................................... 50

6-1 Dosimeter type, period of use, exchange frequency, MRD, and MDL .................................... 53

6-2 Neutron dose energies, percentages, and associated ICRP correction factors ...................... 60

6-3 Parameters for application of unmonitored external doses from 1942 through 1953 .............. 62

6-4 Annual external exposures from uranium for workers in Chemistry Building .......................... 62

6-5 Annual external exposures from uranium for workers in Physical Chemistry Annex 

1.............................................................................................................................................62

6-6 Annual external exposures from uranium for workers in Physical Chemistry Annex 

2.............................................................................................................................................62

6-7 Assigned dose for unmonitored workers ................................................................................ 64

6-8 Potential missed dose for monitored workers......................................................................... 65
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LIST OF FIGURES

FIGURE TITLE PAGE

2-1 ISU campus map showing locations of Ames Laboratory buildings ........................................ 16

2-2 ISU campus map showing locations of reactor facilities and Synchrotron Building................. 16

4-1 Wind rose for Des Moines, Iowa, 1950 to 1955...................................................................... 25

4-2 Ames Laboratory Research Reactor Site ............................................................................... 27

4-3 Environmental monitoring in the vicinity of the ALRR ............................................................. 28

5-1 Urinary excretion of uranium expected from 365-day chronic intake at 6,061 pCi/d 

inhalation and 124 pCi/d ingestion ......................................................................................... 38

5-2 Urinary excretion of uranium expected from 730-day chronic intake at 6,061 pCi/d 

inhalation and 124 pCi/d ingestion ......................................................................................... 38

5-3 Urinary excretion of uranium expected from 1,826-day chronic intake at 6,061 

pCi/d inhalation and 124 pCi/d ingestion ................................................................................ 39

5-4 Bioassay record example 1.................................................................................................... 45

5-5 Bioassay record example 2.................................................................................................... 45

5-6 Distribution of worker doses from tritium in 1970.................................................................... 47

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 ......................................................................................................... 56

6-2 Flowchart for monitored and unmonitored workers................................................................. 66
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 8 of 90

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ALRR Ames Laboratory Research Reactor

ANL-E Argonne National Laboratory–East

AP anterior-posterior

AWE atomic weapons employer

BNL Brookhaven National Laboratory

Ci curie

CFR Code of Federal Regulations

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

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

g gram

GSD geometric standard deviation

HEPA high-efficiency particulate air

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

INL Idaho National Laboratory

IREP Interactive RadioEpidemiological Program

ISU Iowa State University

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

kV kilovolt

L liter

LAT lateral

LBNL Lawrence Berkeley National Laboratory

m meter

MDA minimum detectable activity

MDL minimum detection level 

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mi mile

min minute

mL milliliter

mR milliroentgen
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 9 of 90

mrad millirad

MRD minimum recordable dose

mrem millirem

mrep millirep

MTR Materials Test Reactor

MW megawatt

N/A Not applicable

NCA Nucleonic Corporation of America

NCRP National Council on Radiological Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NTA Eastman Kodak nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

pCi picocurie

POC probability of causation

R&D research and development

s second

SEC Special Exposure Cohort

SLAC Stanford Linear Accelerator Center

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

TASF Technical and Administrative Services Facility

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

yr year

β beta

γ gamma

μCi microcurie

μg microgram

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 10 of 90

1.0 INTRODUCTION

1.1 PURPOSE

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] guidelines established under 

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

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

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2010a).

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2010a):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 11 of 90

1.2 SCOPE

This site profile provides information about U.S. Atomic Energy Commission (AEC), Energy Research 

and Development Administration (ERDA), and DOE operations at Ames Laboratory that pertains to 

radiation exposures for monitored or unmonitored workers. Section 2 provides a description of the 

site and operations that pertain to possible radiation exposures and discusses radiation source terms. 

Section 3 provides guidance for the determination of occupational medical dose. Section 4 provides 

guidance for the determination of dose to workers outside radiological facilities due to releases of 

radioactive materials to the environment. Section 5 provides guidance for the determination of intakes 

of radionuclides inside facilities. Section 6 provides guidance for the determination of external doses 

from measured doses or 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. 

1.3 SPECIAL EXPOSURE COHORT

The Secretary of Health and Human Services has designated four classes of employees at Ames 

Laboratory for addition to the Special Exposure Cohort (SEC) authorized under EEOICPA. 

NIOSH added classes to the SEC in 2006, 2007, and 2010 to cover three separate groups of 

employees based on work location and job description. While the classes added in 2006 and 2007 

included specific workers performing specific tasks in designated buildings, the 2010 class 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 area(s). In 2011, NIOSH 

designated a fourth class (SEC-00185) that encompasses all previous Ames SEC periods and 

designates all Ames employees (including predecessor agencies, contractors, and subcontractors) as 

included. Additionally, SEC-00185 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.

Although NIOSH cannot bound doses for certain areas and periods as described in the SECs, internal 

and external data that become available for an individual claim (and that can be interpreted using 

existing NIOSH dose reconstruction processes or procedures) and applicable dose reconstruction 

methods that are defined in Sections 4, 5, and 6 of this site profile, will be used to complete partial 

dose reconstructions for workers who worked during the SEC periods but are not eligible for the SEC.

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

within the SEC periods. There are varying types of exposures that can be applicable during the SEC 

periods to dose reconstructions for employees who do not qualify for the SEC(s). The periods in 

which doses can and cannot be reconstructed for a particular claim should be identified when writing 

the Dose Reconstruction Report. For dose reconstruction of claims with employment during the SEC 

period (see Section 1.3.4), all Ames employees are included and the dose reconstruction is a “partial 

reconstruction” by default. This designation should be included in the Dose Reconstruction Report.

1.3.1 August 13, 1942, through December 31, 1954

This SEC includes DOE employees or DOE contractor or subcontractor employees who worked at the 

Ames Laboratory in one or more of the following facilities or locations: Chemistry Annex 1 (also 

known as the old women’s gymnasium and Little Ankeny), Chemistry Annex 2, Chemistry Building 

(also known as Gilman Hall), Research Building, or Metallurgical Building (now Harley Wilhelm Hall), 

from January 1, 1942, through December 31, 1954, for a number of workdays that total at least 250 

workdays, or in combination with workdays within the parameters (excluding aggregate workday 

requirements) established for one or more classes of employees in the SEC, and who were monitored 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 12 of 90

or should have been monitored. As stated above and defined in Section 1.3.4, this SEC class has 

effectively been expanded to include all Ames employees, contractors, and subcontractors who 

worked in any area of the Ames Laboratory at Iowa State University, and the start date for this SEC 

was changed to August 13, 1942.

NIOSH evaluated the feasibility for completing dose reconstructions for employees at Ames 

Laboratory from August 13, 1942, through December 31, 1954, and found that the monitoring records, 

process descriptions, and source term data available are not sufficient to perform complete dose 

reconstructions for the SEC (SEC-00038) class of employees (NIOSH 2006). Table 1-1 summarizes 

the results of the feasibility findings for each exposure source for the period from August 13, 1942, 

through December 31, 1954.

Table 1-1. Feasibility findings for SEC-00038, August 13, 1942, through December 

31, 1954. (NIOSH 2006, Table 7-1).

Source of exposure

Dose reconstruction 

is feasible

Dose reconstruction 

is NOT feasible

Internal dose:

Uranium X

Thorium/plutonium X

Thoron X

External dose:

Uranium beta/gamma X

Thorium/plutonium beta/gamma X (except 1953 and 1954)

Neutron X

Occupational medical X-ray X

1.3.2 January 1, 1955, through December 31, 1970

This SEC includes sheet metal workers, physical plant maintenance and associated support staff 

(includes 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, for the period from January 1, 1955, through 

December 31, 1970, and who were employed for a number of workdays aggregating at least 250 

workdays, either solely under this employment or in combination with workdays within the parameters 

(excluding aggregate workday parameters) established for other classes of employees included in the 

SEC. As stated above and defined in Section 1.3.4, this SEC class has effectively been expanded to 

include all Ames employees, contractors, and subcontractors who worked in any area of the Ames 

Laboratory at Iowa State University.

NIOSH evaluated the feasibility for completing dose reconstructions for employees at Ames 

Laboratory from January 1, 1955, through December 31, 1970. NIOSH found that the monitoring 

records, process descriptions, and source term data available are not sufficient to perform complete 

dose reconstructions for the SEC (SEC-00075) class of employees (NIOSH 2007a). Table 1-2 

summarizes the results of the feasibility findings for each exposure source for the period from January

1, 1955, through December 31, 1970.

Table 1-2. Feasibility findings for SEC-00075, January 1, 1955, through December 

31, 1970. (NIOSH 2007a, Table 7-1).

Source of exposure

Dose reconstruction 

is feasible

Dose reconstruction 

is NOT feasible

Internal

Th-232 and progeny X

Ambient environmental X

External
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Gamma X

Beta X

Neutron N/A

Ambient environmental X

Occupational medical X-ray X

1.3.3 January 1, 1955, through December 31, 1960

This SEC includes all employees of DOE, its predecessor agencies, and its contractors and 

subcontractors who worked in any area of DOE facilities on the Ames Laboratory Campus from 

January 1, 1955, through December 31, 1960, for a number of workdays aggregating at least 250 

workdays, occurring either solely under this employment, or in a combination with workdays within the 

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

NIOSH evaluated the feasibility for completing dose reconstructions for employees at Ames 

Laboratory from January 1, 1955, through December 31, 1960, and found that the monitoring records, 

process descriptions, and source term data available are not sufficient to perform complete dose 

reconstructions for the SEC (SEC-00166) class of employees (NIOSH 2010b). Table 1-3 summarizes 

the results of the feasibility findings for each exposure source for the period from January 1, 1955, 

through December 31, 1960.

Table 1-3. Feasibility findings for SEC-00166, January 1, 1955, through December 

31, 1960. (NIOSH 2010b, Table 7-5).

Source of exposure

Dose reconstruction 

is feasible

Dose reconstruction 

is NOT feasible

Internal dose:

Uranium and progeny X

Thorium and progeny X

Other radionuclides (Research 

Building)

X

External dose:

Gamma X

Beta X

Neutron X

Occupational medical X-ray X

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

evaluated class, 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 January 1, 1955 through December 31, 1960, but who do not qualify for inclusion in 

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

1.3.4 Class Recommended by NIOSH for Addition to the SEC: August 13, 1942, through

December 31, 1970

This SEC includes all employees of DOE, its predecessor agencies, and its contractors and 

subcontractors who worked in any area of DOE facilities on the Ames Laboratory at Iowa State 

University during the period from August 13, 1942, through December 31, 1970, for a number of 

workdays aggregating at least 250 workdays, occurring either solely under this employment, or in a 

combination with workdays within the parameters established for one or more other classes of 

employees in the SEC.

This SEC, SEC-00185 (NIOSH 2011), does not make any new information available for the feasibility 

of performing dose reconstructions between the dates of August 13, 1942, and December 31, 1970. 
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NIOSH previously designated the three classes associated with Sections 1.3.1, 1.3.2, and 1.3.3. 

NIOSH has determined that site-specific and claimant-specific data available for Ames Laboratory for 

this entire period are insufficient to enable it to determine that a specific work group was not 

potentially exposed to radioactive material releases or possible subsequent contamination. Based on 

this information, NIOSH has determined that the previously proposed SEC class definitions cannot be 

based on or limited to job titles or duties.

The guidance (and limitations) for performing dose reconstructions using the information in this site 

profile have not changed. However, the class definitions for the SECs in Sections 1.3.1 and 1.3.2 

have effectively been expanded to include all Ames employees, contractors, and subcontractors who 

worked in any area of the Ames Laboratory at Iowa State University.
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2.0 SITE DESCRIPTION

The Ames Laboratory site consists of a number of buildings at Iowa State University (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 Iowa State College 

(Fulmer 1947). Ames Laboratory was established by the AEC in May 1947 (Karsjen 2003). The 

Ames Project/Laboratory played a key role in the production of strategic nuclear materials for the 

Manhattan Project and the AEC.

Early in 1942, prior to 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 Iowa State College 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 used 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 tons 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, more than 65 tons of pure thorium metal and thorium compounds were produced by the 

Ames Laboratory (Ames 1960a).

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

Laboratory handled a number of 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 Iowa State College that were used 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). 

In November 1945, the buildings used by the Ames Project were designated as 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). In 

May 1947, the 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. Construction of a 

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

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

(Spedding Hall) and the Chemistry Building (Gilman Hall); the TASF 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. 
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Figure 2-1. ISU campus map showing locations of Ames Laboratory buildings. The former 

sites of Physical Chemistry Annex 1 and Annex 2 are also shown. Source: ISU (2006).

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

Synchrotron Building. Source: ISU (2006).
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To avoid confusion between the historical building names and the current names, as shown in 

Figures 2-1 and 2-2, both names are used in the following sections with the historical name first and 

the current name in parentheses, where applicable.

2.1.1 Physical Chemistry Annex 1

The production of uranium metal was conducted in Physical Chemistry Annex 1, which was 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 tons of pure uranium and more than 300 tons 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 

1979). 

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, including 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 

conducted in the “hot laboratory” between 1942 and 1951 (Ames 1960a). The Chemistry Building 

was decontaminated and surveyed in May 1976 (Voss 1979).

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).

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 the recovery of uranium from scrap uranium metal turnings collected from other Manhattan 

Project sites. Operations in this building through December 1945 produced more than 300 tons of 

recovered uranium metal (Fulmer 1947). Operations ended in 1953 when the building was converted 

to a plumbing shop; it was razed in 1972 (Ames 1985). 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 1979). Part of the General Services Building now covers part of the former building site 

(see Figure 2-1). 
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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 directed toward the 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 zirconium-tin alloy phase diagram (Ames 1962a). The subject of reactor coolants was studied, 

as were 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 instruments 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 

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.

2.1.7 Office and Laboratory Building

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

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

chemists and physicists, and a large physical sciences reading room (Ames 1951). 

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 radioisotopes for nuclear spectroscopy (Ames 1962a). Operations at the 

Synchrotron Building ended in June 1971; the equipment was decommissioned in the early 1990s.
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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 processes developed. One major process was the production of 

very pure metals, and the production of small quantities of these metals 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, liquidliquid 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 a 5-MW, heavy-water-moderated research reactor 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). 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 surrounding 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 of tritium per liter 

(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 thorium metal (1943 to 1953). A number of 

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).
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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 and 6. 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 and 6. Local environmental contamination around Annex 2 

was assumed to be similar to that around Annex 1, and this is addressed in Section 4.1. 

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 and 6. 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 through 1954 is 

presented in Table 2-1.

Table 2-1. Timeline of AEC activities at Ames Laboratory, 1942 through 1954.

Operation Datesa Building Activity

Uranium metal production Feb 1942–Aug 1942 Chemistry (Gilman Hall) Process development

Uranium metal production Aug 1942–Dec 1942 Chemistry Production

Uranium metal production Sept 1942–Aug 1945 Annex 1 Production

Uranium scrap recovery Late 1943–early 1944 Chemistry Process development

Uranium scrap recovery Early 1944–Dec 1953 Annex 2 Production
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Uranium metal casting Sept 1942–Aug 1945 Annex 1 Production

Thorium metal production Aug 1943–Jun 1946 Annex 1 Process development

Thorium metal production Jun 1946–late 1949 Annex 1 Production

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

Studies of plutonium properties Jun 1943–Dec 1947 Chemistry & Metallurgy Research

Plutonium/fission product 

separations

Summer 1943–Dec 1954 Chemistry & Research 

(Spedding Hall)

Research and hot cell 

work

Thorium metal casting Jun 1946–late 1949 Annex 1 Production

Thorium metal casting Early 1950–Apr 1953 Metallurgy Production

Studies of uranium and thorium 

properties

Early 1942–Dec 1954 Chemistry Research

Development of analytical 

procedures

Early 1943–Dec 1954 Chemistry Research

Annex 1 demolition 1953 Annex 1 Demolition

a. Exact start-end dates vary depending on the reference. Listed dates are the most common or consensus.

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, 125Sb, and five isotopes of 

ruthenium and rhodium. Research on the parent-daughter relationship of 90Sr/90Y 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 members were exposed to a number of radiation sources, 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 1979). 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.

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 fluoride 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 potential 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 
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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 were estimated. When reactor 

operations started in 1965, a routine tritium bioassay program was started, and tritium intakes can be 

reliably estimated. Occupational internal dose is addressed in Section 5. 

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 

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, suggesting 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). 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, are used to estimate earlier radiation doses. 

Records of external radiation exposures received from 1955 to the present are used in Section 6 to 

evaluate occupational external dose.
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3.0 OCCUPATIONAL MEDICAL DOSE

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

obtained at covered facilities (ORAUT 2011b). The X-ray equipment for occupational medical 

examinations of Ames Laboratory staff members was located at the Iowa State Student Health 

Center/College Hospital (Voss 1957a), which is not a covered facility under EEOICPA. Therefore, no 

occupational medical doses are included.
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4.0 OCCUPATIONAL ENVIRONMENTAL DOSE – INTRODUCTION

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

facilities (e.g., 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 Ames Laboratory Research Reactor (ALRR). While most radionuclides when 

inhaled would give a dose to particular organs in the body, tritium gas would give a dose to the whole 

body (Voigt 1973). These radionuclides are addressed in the following sections.

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

workers were badged (Martin 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). 

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 4.5. 

4.1 URANIUM/THORIUM PRODUCTION PERIOD, 1942 TO 1953

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

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

(see Figure 2-1) from early 1944 through December 1945, and some operations continued until 1954. 

Thorium production occurred first in Physical Chemistry Annex 1 from 1944 through 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 1963).

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

outside the facilities; that is, input materials arrived at the facilities and were processed, and products 

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). 

There were no documents found that stated the room and hood ventilation stacks on the facilities had

filters. Concentrations of uranium dust were measured in the operation rooms (Voss 1978). To 

estimate a bounding dose outside the facilities, it was assumed that losses of 0.1% of the uranium or 

thorium as dust in a facility were emitted continuously and dispersed from ground level in accordance 

with local and regional meteorological conditions (see Figure 4-1) (Voss 1981) and a standard 

Gaussian atmospheric dispersion computer model (Napier et al. 2004). When resuspension is 

included, the daily intake rate of uranium (modeled as 234U, type M or S) was 5 pCi. This intake 

applies outside Physical Chemistry Annex 1 for August 1942 through August 1945 and outside 

Physical Chemistry Annex 2 for 1944 through 1954. In addition, daily intakes of 232Th, 228Th, and 

228Ra at 0.07 pCi each apply to Annex 1 for June 1946 through 1949 and to the Metallurgy Building for 

1950 through April 1953. These are upper bound intakes so the distribution is constant (Napier 

2006a). 

4.2 SYNCHROTRON OPERATIONS PERIOD, 1949 TO 1971 

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

1949 through June 1971 (see Figure 2-2). Unlike the uranium and thorium facilities, this facility was 
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Figure 4-1. Wind rose for Des Moines, Iowa, 1950 to 1955. Source: Voss (1981).

fenced, providing a defined exclusion area outside the building. Dosimeters were not provided to 

anyone at the facility until late 1952 (Martin 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 [1]. No routine measurements of direct gamma or neutron 

radiation were made outside the buildings, but a detailed survey of the facility, including fenceline

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). 

The fenceline gamma dose rates measured in the radiation survey (Ames 1961a) were worst case 

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

maximum dose rates at the fenceline [2]. 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 members (tens of milliroentgen per hour) 

(Ames 1961a). 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) 

(Martin 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 

been about 2,000 hr/yr [3]. 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) [4].
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Table 4-1. Survey of fenceline gamma dose rates around the 

synchrotron facility with the beam directed west (Ames 1961a, 

p. 7).

Location Dose rate (mrem/hr)a Location

Dose rate 

(mrem/hr)a

1 0.75–1 14 1.8

2 3.5 15 1.8

3 7 16 1.8

4 6 17 1.6

5 5.5 18 1.5

6 7 19 1.5

7 4.75 20 1.75

8 3.5 21 1.6

9 2.6 22 1.6

10 1.5 23 1.5

11 2 24 0.75–1

12 2 25 4.5

13 1.8

a. Background reading in the beam direction (see locations 3-6) before turning 

on the beam was 0.5 to 1.0 mrem/hr.

A topographical map that was part of the 1961 radiation survey (Ames 1961a) 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) 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 (NCRP 2003). If it is 

conservatively assumed that all of 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 ALRR OPERATIONS PERIOD, 1965 TO 1977

The ALRR started operations in February 1965 and continued through December 1977 (Ames 1967; 

Voigt 1981). The facility 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 employees working inside the ALRR fence were provided dosimeters. 

However, not all dosimetry records are identified with names. Therefore, not all workers at the ALRR 

have recorded doses. In addition, environmental doses from gaseous effluents released 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 in the 

vicinity of 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 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 27 of 90

Figure 4-2. Ames Laboratory Research Reactor Site. Source: Ames (1974).
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 28 of 90

Figure 4-3. Environmental monitoring in the vicinity of the ALRR. Source: Voss (1975).

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 (Napier et al. 2004; Napier 2006b), was used to determine external dose rates from 

the 41

Ar and inhalation intake estimates for the tritium.

At the fenceline 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. An offsite worker who worked full time at this location would not have been 

exposed for more than 2,080 hr/yr, which would result in a submersion dose of about 1 mrem/yr.

At the fenceline 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. An offsite worker who worked full time at this location 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.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 29 of 90

4.4 AMES LABORATORY RESEARCH AND DEVELOPMENT ACTIVITIES, 1942 TO 

PRESENT

R&D activities have been carried out at Ames Laboratory from 1942 to the present. It was the R&D 

capabilities of Iowa State College that attracted the attention of the Manhattan Engineer District in 

1942 (Ames 1960a). Buildings owned by the government (AEC, ERDA, and now DOE) in addition to 

the uranium/thorium, synchrotron, and research reactor buildings discussed above, are on the main 

campus of ISU (see Figure 2-1). 

The Ames Laboratory facilities consist of (Ames 1996):

• Metals Development Building

• Spedding Hall (formerly the Research Building)

• Wilhelm Hall (formerly the Metallurgy Building)

• TASF

Facilities owned by ISU but leased to the Government include (Ames 1996):

• Gillman Hall (formerly the Chemistry Building)

• Zaffarano Physics Addition

• Office and Laboratory Building

As can be seen in Figure 2-1, these R&D facilities are an integral part of the ISU campus and are, 

therefore, open to staff, students, and public traffic [5]. There were no restrictions on the movement of 

this traffic outside the facilities. In addition, there was no monitoring of radiation exposure or 

contamination of the personnel or environmental areas around and among these facilities. As 

research facilities, the quantities of radioactive materials involved in the R&D work were small 

compared to the production facilities.

There was insufficient information about releases from the R&D facilities to determine intakes directly. 

Principal sources would have been uranium, thorium, and fission products from the hot canyon/hot 

cell in the Research Building. Because of the smaller amounts of radioactive materials in the R&D 

facilities in comparison with those in the production facilities, the releases from the R&D facilities were 

assumed to be one one-hundredth of the releases from the production facilities [6]. That assumption 

resulted in daily environmental intakes of 5 × 10-2 and 7 × 10-4 pCi/d for uranium and thorium, 

respectively. A review of environmental intakes at somewhat related sites [Argonne National 

Laboratory–East (ANL-E), Brookhaven National Laboratory (BNL), and Lawrence Berkeley National 

Laboratory (LBNL)] was made. Neither ANL-E nor BNL listed any environmental intakes of uranium 

or 232Th (ORAUT 2006a, 2010a). LBNL listed environmental intakes of unspecified gross alpha and 

gross beta emitters. From 1939 through 1999, the gross alpha intakes ranged from 6.6 × 10-3 to 

2.7 × 10-1 pCi/d, and the gross beta intakes ranged from 1.5 × 10-1 to 63 pCi/d (ORAUT 2010b). The 

highest gross alpha intake occurred in 1996 and appears to be an anomaly; the second highest gross 

alpha intake was 5.9 × 10-2 pCi/d. Therefore, the proposed uranium intake is comparable to the 

second highest gross alpha intake from LBNL and, therefore, appears to be a reasonable upper 

bound. For environmental intakes for workers exposed around the R&D facilities, a daily intake of 

5 × 10-2 pCi of 234U and 7 × 10-4 pCi each of 232Th, 228Ra, and 228Th was assumed [7]. Because the 

nature of the research materials is not known, absorption types F, M, or S for the uranium and M or S 

for the thorium/radium isotopes were assumed [8].

A hot laboratory was operated in the Chemistry Building but 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 means of ion exchange columns. A final report on 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 30 of 90

plutonium R&D was published in July 1946 (Fulmer 1947), at which time plutonium R&D and the 

personnel who conducted it were transferred to Los Alamos National Laboratory (LANL; Payne 1992). 

A hot laboratory, capable of handling 5 curies through the adsorption process, was 

designed and built. The method proved successful in a number of runs. The uranium 

and plutonium were separated from one another and from the fission products using 

1 kg samples of uranium, from the pile, which had an activity of 5 curies. (Fulmer 

1947)

Releases were estimated from this operation by making the following assumptions (see Bihl 2006 for 

details of the calculation):

• A total annual throughput of the laboratory of 50 Ci/yr

• An airborne fraction of 0.002 for boiling liquids

• Filtration efficiency for average particle sizes of 99.5% [high-efficiency particulate air (HEPA) 

filters were just being developed during this period]

These assumptions produce a total release of fission products of 5 × 108 pCi/yr. 

Using the least dispersive approach recommended by the National Council on Radiation Protection 

and Measurements (NCRP) in Publication 123 (NCRP 1996), the ground-level annual average air 

concentration would have been 13.2 pCi/m3

. Using the inhalation rate of 2,400 m3/yr and converting 

to a daily intake results in 87 pCi/d.

Using the recommended fission product ratios for 180-day cooled fuel from Fission and Activation 

Product Assignment for Internal Dose-Related Gross Beta and Gross Gamma Analyses (ORAUT 

2007a), an 87-pCi/d intake of fission products is assigned to specific radionuclides as listed in 

Table 4-2. 

 Table 4-2. Annual environmental fission

product intakes from the hot laboratory

in the Chemistry Building (1943–1981).

Radionuclide

Relative 

fraction

Intake 

(pCi/d)

Ce-141 0.0221 1.92

Ce-144 0.2191 19.1

Cs-134 0.0054 0.470

Cs-137 0.0208 1.81

Eu-155 0.0014 0.122

Fe-55 0.0172 1.50

Nb-95 0.2492 21.7

Pm-147 0.0546 4.75

Ru-103 0.0321 2.79

Ru-106 0.0844 7.34

Sr-89 0.0558 4.85

Sr-90 0.0157 1.37

Y-91 0.0911 7.93

Zr-95 0.1311 11.4

Because maximizing assumptions were used for most of the parameters in this analysis, the 

distribution is an upper bound (constant). The absorption type for each radionuclide that is most 

favorable to claimants, as listed in International Commission on Radiological Protection (ICRP) 
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Publication 68 (ICRP 1995), should be used, with the exception that only type F should be used for 

strontium. The date of the first use of the hot laboratory was not found. Because it received fuel from 

the Chicago Pile, the earliest reasonable use would have been 1943 [9]. It was replaced by a hot cell 

with improved ventilation and filtration in the Research Building in 1951, which was removed from 

service and decontaminated in 1982. Releases from the hot cell are not documented but would have 

been smaller than those estimated above for the hot laboratory. It is favorable to claimants to apply 

the Table 4-2 intakes for each year from 1943 through 1981 [10]. 

4.5 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 through 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 completetreatment 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 that there was little hazard to the public or the employees of the sewer plant (Hayes

1956). Table 4-3 is a summary of environmental dose measurements at the four facilities in Ames, 

Iowa:

Table 4-3. Environmental dose measurements at the four 

Ames facilities (mrem/yr). Source: Voss 1979.

Lawn site Maximuma Average

Sewer plant 780 400

Cemetery 440 350

Airport 530 350

Parkway 530 350

a. Background in central Iowa is about 300 mrem/yr.

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

would reduce the doses in Table 4-3 by a large fraction, making them below the current public dose 

limit of 100 mrem/yr (10 CFR Part 20).

All the locations in Table 4-3 are outside the ISU campus and all are 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. However, the 

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

materials while making radiation measurements and collecting environmental samples for analysis. It 

is reasonable to assume that radiation exposures to the health physics staff were monitored and 

offsite doses were included in their recorded occupational doses. Thus, no additional dose is 

recommended for occupational environmental dose for any workers.

4.6 ENVIRONMENTAL CONTAMINATION, 1953 TO PRESENT 

Gamma spectral analysis of samples of environmental media was not initiated 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 (Ames 1985). Synchrotron operations ceased in 1971, and the building 
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was decommissioned in 1990 [11]. 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 4-4. 

Table 4-4. Gamma analysis of soil samples (Voss 1975).

Site Sample weight (g) Cs-137 (pCi/g)a

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. 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 4-5.

Table 4-5. Summary of 24 soil samples for 

beta and alpha activity (pCi/g). (Voss 1975).

Activity Betaa Alphab

Average 10.56 0.61

High 13.01 0.96

Low 7.66 0.38

Background 11.9 0.76

a. Beta detection limit was 0.25 pCi/g.

b. 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 position 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.

4.7 SUMMARY OF ENVIRONMENTAL EXTERNAL DOSES AND INTAKES

Tables 4-6 and 4-7 summarize the conclusions of the preceding sections on environmental external 

doses and intakes.

Table 4-6. Environmental external doses (to be applied only to unmonitored workers).

Location Dates

Dose 

(mrem/yr)a Distribution

Skyshine from Synchrotron Building 1949–June 1971 25 Constant

Ar-41 from ALRR 1965–1977 1 Constant

All other R&D Buildings (see list in Section 4.4) All Negligible N/A

a. The energy range for all environmental external dose is assumed to be 100% 30–250 keV (NIOSH 2007b).
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 33 of 90

Table 4-7. Summary of environmental intakes.

a

Dates Radionuclide/absorption Intake (pCi/d) Distribution

August 1942–1953 U (assume U-234); type M or S 5 Constant

1954–present U (assume U-234); type F, M, or S 0.05 Constant

June 1943–April 1953 Th-232, Ra-228, Th-228; all type M 0.07 each Constant

1954–present Th-232, Ra-228, Th-228; type M or S 0.0007 each Constant

1965–1977 Tritium 2,700 Constant

1943–1981 Fission products per Table 4-2 Per Table 4-2 Constant

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 instructions in Table 5-8. 
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5.0 OCCUPATIONAL INTERNAL DOSE

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

fission products. Based on the Evaluation Report for Petition SEC-00038 (NIOSH 2006) for the 

period from August 13, 1942, through December 31, 1954, it was determined that internal doses from 

the production and casting of thorium metal for DOE employees or contractor or subcontractor 

employees who worked in Chemistry Annex 1 (also known as the old women’s gymnasium and Little 

Ankeny), Chemistry Annex 2, Chemistry Building (now Gilman Hall), Research Building, or Metallurgy

Building (now Harley Wilhelm Hall), cannot be reconstructed. Therefore, only doses from thorium 

contamination left over after 1954 are addressed here. As stated in Section 1.3, this SEC class has 

been expanded to include all Ames employees, contractors, and subcontractors. 

In addition, based on the Evaluation Report for Petition SEC-00075 (NIOSH 2007a) for the period 

from January 1, 1955, through December 31, 1970, it was determined that internal doses from 

maintenance and renovation activities of the thorium production areas in Wilhelm Hall (a.k.a. the 

Metallurgy Building or “Old” Metallurgy Building) cannot be reconstructed for sheet metal workers, 

physical plant maintenance, and associated support staff (includes all maintenance shop personnel of 

Ames Laboratory) and supervisory staff who were monitored or should have been monitored. 

Therefore, internal doses from these activities to this class of workers cannot be assessed using the 

approach described in Section 5.2. As stated in Section 1.3, this SEC class has been expanded to 

include all Ames employees, contractors, and subcontractors.

In addition, based on the Evaluation Report for Petition SEC-00166 (NIOSH 2010b) for the period 

from January 1, 1955, through December 31, 1960, it was determined that internal doses from 

radionuclides other than uranium in the Hot Canyon and Hot Cell (or the Cave) of the Research 

Building (Spedding Hall) cannot be determined for any workers. Further, this petition was expanded 

to include all workers at Ames Laboratory during this period due to the lack of information on work 

locations of employees.

5.1 URANIUM EXPOSURE

Few data on uranium exposure were found (Stone 1951); however, the SEC petition evaluation 

reports determined that doses could be estimated (NIOSH 2006, 2007a, 2010b). If uranium bioassay 

data are available for a worker, they should be used. However, because it is unlikely that uranium 

bioassay data will be available, default intakes were determined. Because even workplace sampling 

data were unavailable, this analysis used data from the documents: Site Profiles for Atomic Weapons 

Employers (AWE) that Worked Uranium Metals (Battelle 2011) and The Industrial Hygiene of Uranium 

Refining (Christofano and Harris 1960) to estimate doses to workers in the Chemistry Building, 

Physical Chemistry Annex 1, and Physical Chemistry Annex 2. These two documents are 

representative of the potential intakes that workers at Ames Laboratory might have had because the 

processes were often developed at the Laboratory and remained similar to processes used at AWE 

sites. The two processes that were used here to estimate doses are Metal Reduction (Christofano 

and Harris 1960) and Uranium Scrap Recovery and Casting (Battelle 2011).

5.1.1 Estimating Internal Doses from Uranium Metal

Historical aspects of the production of uranium metal at Ames Laboratory are discussed in Section 2. 

Although there was much historical information on the general methods for producing uranium metal 

(Fulmer 1947), there was very little specific information useful for determining intakes from uranium. If 

uranium bioassay data are available for a worker, they should be used to estimate the intake. 

However, it is likely that bioassay data will not be available. In that case, the information below has 

been developed from Battelle (2011) and Christofano and Harris (1960) as methods for assigning 

internal dose to workers.
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The next two sections describe estimated doses for workers from August 13, 1942, through 

December 31, 1954. Additional information is provided to calculate dose to workers from radioactivity 

left over from the sites where uranium had been processed after 1954. Because there were 

numerous studies of various compounds of uranium during this period, all three absorption types (fast, 

moderate, and slow) were possible. 

5.1.1.1 Uranium Inhalation

From 1942 through August 1945, individuals working in Physical Chemistry Annex 1 (also known as 

Little Ankeny) can be assigned doses from inhalation of uranium [12]. Similarly, individuals who 

worked in the Chemistry Building during this period or in Physical Chemistry Annex 2 from 1944 

through December 1945 should be assigned doses from inhalation of uranium [13]. Job titles of 

researchers acknowledged for their work in uranium production (Fulmer 1947) include (1) chemist, (2) 

associate chemist, (3) junior chemist, (4) research assistant, (5) junior research assistant, 

(6) physicist, (7) analyst, (8) assistant physicist, (9) associate director, and (10) director. However, it 

is not entirely clear how much time these researchers spent in the area where production was 

performed.

Because it is not clear if there were clerical, janitorial, or nontechnical personnel and other types of 

researchers working in these buildings, and it is not known what precautions might have been taken 

for contamination control, it can be assumed that all individuals who worked in the buildings had some 

potential for exposure to uranium.

The data in Tables 5-1 and 5-2 were derived from data in Christofano and Harris (1960). In 

Christofano and Harris (1960), there is a description of the process for metal reduction that is similar 

to the process used at Ames Laboratory for production of uranium metal (Fulmer 1947). The primary 

difference appears to be that at Ames Laboratory the process used granulated calcium metal and at 

AWE sites the process used magnesium. Fulmer (1947) describes the process using magnesium, 

and it appears to be similar enough to be representative of the intakes at Ames Laboratory. 

Table 5-1. Chemistry Building uranium intakes (pCi/d).

Period Inhalation Ingestion

Aug 1942–December 1953 8.5a,b 0.09b

a. No data were available for determination of intakes in the Chemistry Building; 

therefore, it was assumed that research activities would have one-hundredth the 

intake of production activities since uranium metal production was moved to the 

Physical Chemistry Annex 1.

b. Values are for workers assumed to work in research or production full time. For 

supervisors, assume one-quarter of the intake; for all other employees (clerical, 

janitorial, security, etc), assume one-tenth of the supervisor’s intake.

Table 5-2. Physical Chemistry Annex 1 uranium intakes (pCi/d).

Period Inhalation Ingestion

August 1942–August 1945 853a 8.7a

a. Values are for workers assumed to work in research or production full time. 

For supervisors, assume one-quarter of the intake; for all other employees 

(clerical, janitorial, security, etc), assume one-tenth of the supervisor’s intake.

Table 5-3. Physical Chemistry Annex 2 uranium intakes (pCi/d).

Period Inhalation Ingestion

January 1944–December 1950 6,061a 124a

January 1951–December 1953 5,556a 114a

a. Values are for workers assumed to work in research or production full time. For 

supervisors, assume one-quarter of the intake; for all other employees (clerical, 

janitorial, security, etc), assume one-tenth of the supervisor’s intake.
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Christofano and Harris (1960) provides data for multiple stages of the production operation for 

operators. Using the techniques outlined in Default Assumptions and Methods for Atomic Weapons 

Employer Dose Reconstructions (Battelle 2007), the values in the above tables were calculated from 

Table 8 of Christofano and Harris (1960). However, review of Ames Laboratory documentation did 

not reveal any information specific enough for determination of who would be responsible for what 

aspects of the process and for how long. Therefore, the value used for determining inhalation intakes, 

which was from Christofano and Harris (1960, Table 8), is for the Bomb Preparation operator, who is 

assumed to work a 2400-hour year. This value was the highest intake rate for the metal reduction 

process. This number is then scaled for the potential for intake (see Tables 5-1 and 5-2). 

Battelle (2011) provides data for a process of scrap recovery that is similar to the scrap recovery 

process described in Fulmer (1947). Data from Battelle (2011, Table 7.8) are used in Table 5-3 for 

Ames Laboratory workers who worked in the Physical Chemistry Annex 2 building. The value used in 

Battelle (2011, Table 7.8) represents the most conservative intake for the scrap recovery operations, 

similar to those performed in Physical Chemistry Annex 2.

Research in the Chemistry Building began in January 1942. For workers involved only in research 

from January through July 1942 in the Chemistry Building, an exposure of one-tenth of that of the 

workers involved in the production operations is assumed, which corresponds to smaller quantities of 

uranium [14]. This period is prior to the beginning of the covered period for the EEOICPA statute, 

which is August 1942, the start of the Manhattan Engineer District, known later as the Manhattan 

Project. The process developed by this research was moved to the Physical Chemistry 1 building in 

July/August of 1942. There are very few details on the research activities in the Chemistry Building 

after July 1942; therefore, it was assumed that another one-tenth fraction should be applied (for a 

1/100 reduction overall from operations). These intakes should be applied to researchers in the 

Chemistry Building through 1953, when production ended.

Individuals supervising the production processes were assumed to be exposed for one-fourth of the 

time of the production staff [15].

For workers not directly associated with uranium metal research or production, an exposure of onetenth of that of the supervisors was assumed (see Tables 5-1, 5-2, and 5-3) [16].

The intakes in Tables 5-1, 5-2, and 5-3 were compared to the few actual bioassay results found for 

workers in approximately the same period. Chapter 7 in Stone (1951), “Uranium Excretion Studies,” 

provided data from a series of uranium bioassays obtained from Ames Laboratory workers in 1944 

and 1945. Of special interest was a series of samples from the supposedly highest exposed worker 

at Ames Laboratory and samples from the most highly exposed group of workers at the Laboratory

(21 samples from 11 workers). An intake evaluation was performed on the results for the highest 

exposed worker assuming chronic intake from the start of that person’s employment and absorption 

type M uranium (the document indicated the person was exposed to UF4). The estimated intake was 

1,200 μg/d or 820 pCi/d assuming natural uranium. This is consistent with the intake estimate in 

Table 5-2 for Physical Chemistry Annex 1 and quite a bit lower than that in Table 5-3 for Annex 2. 

The average bioassay result for the group of highest exposed workers was 75 μg/L. Assuming 

chronic intake for 1 year before the bioassay, the estimated intakes were: 

• Absorption type F: 390 μg/d, 260 pCi/d

• Absorption type M: 1,670 μg/d, 1,100 pCi/d

• Absorption type S: 45,400 μg/d, 31,000 pCi/d.

If the geometric mean of the data is used, the estimated intakes are smaller; if the highest bioassay 

result of the set is used, the estimated intakes are 2.7 times greater. For type M, this range is still 

consistent with the intakes in Tables 5-1, 5-2, and 5-3. The type S intake estimates are greater than 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 37 of 90

those in the tables but it is unlikely anyone was actually exposed to just type S for long periods. 

Uranium tetrafluoride and metal were the principal forms of uranium in the production facilities. 

Another small set of urinalysis data was found involving five or six samples from six workers from 

July 13 to September 4, 1944 (Tybout 1944). The data were hard to read and most results were listed 

as 0.00 mg/L. Using the data from the worker with the highest result and the most nonzero results 

and assuming chronic intake for 1 year of absorption type M uranium, the estimated intake was 

850 μg/d (580 pCi/d). The estimated intake was slightly less if the intake period started in August 

1942. These results provide a measure of confidence that the default intakes in Tables 5-1, 5-2, and 

5-3 are favorable to claimants [17].

5.1.1.2 Uranium Ingestion

Ingestion intakes are determined using NIOSH (2004) and are listed in Tables 5-1, 5-2, and 5-3. To 

describe this process in brief, the daily ingestion rate from food contamination is given by I = 0.0985A

, where I is the daily ingestion rate in picocuries per day and A is the average air concentration in 

picocuries per cubic meter. The ingestion rate I must be adjusted for the fact that the Integrated 

Modules for Bioassay Analysis (IMBA) computer program assumes chronic intakes, even during 

weekends, and that the number of hours worked in a year changed over time (Battelle 2007). The 

adjustments result in an IMBA chronic intake rate I Ah 5

IMBA 3.373 10− = × where IIMBA is the daily 

intake, A is the median air concentration using the specified units, and h is the number of hours in a 

work year. Making similar adjustments to Neton’s equations for incidental hand-to-mouth ingestion 

(NIOSH 2004), the chronic IMBA daily intake rate is I Ah 5

IMBA 3.425 10− = × where IIMBA is the daily 

chronic intake rate, A is the median dust concentration, and h is the number of hours in a work year. 

The total ingestion rate is the sum of the food contamination and incidental hand-to-mouth ingestion 

rates and is II Ah 5

MBA 6.798 10− = × where the variables are the same as defined above. This intake 

applies to the period from 1942 through August 1945. The rationale used in inhalation intakes for 

reduced fractions for other workers is applied to ingestion intakes as well. 

5.1.1.3 Resuspension During Periods with No Uranium Operations

There was a potential for internal exposure to resuspended material from AEC work during non-AEC 

operations soon after the actual operations. To estimate exposure from resuspended materials, this 

analysis assumed that surfaces in the building became contaminated by deposition of uranium dust 

during operations similar to metal working operations. The daily intake rates for the Chemistry 

Building can be estimated by reducing the generic exposure estimate from ORAUT (2006e) by a 

factor of 10 to account for the conclusion of the work, the standard laboratory precautions that were in 

place, and the production time at about 50% of full time in the Chemistry Building. The maximum 

intake from inhalation for the researchers in metal production operations (Physical Chemistry Annex 

1) and scrap recovery (Annex 2) were used to estimate intakes for these facilities. The method used 

in Battelle (2011) was applied to the maximum intake described for each process. Table 5-4 lists the 

intake in picocuries per day for each building for the specified periods. These intakes apply to anyone 

in the building, regardless of job description. NIOSH (2004) indicates that the ingestion rate, in terms 

of disintegrations per minute for an 8-hour workday, can be estimated by multiplying the air 

concentration in disintegrations per minute per cubic meter by a factor of 0.2. The ingestion intakes 

are also provided in Table 5-4. 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 38 of 90

Table 5-4. Resuspension during periods with no uranium operations (pCi/d).

Period

Chemistry Building

inhalation/ingestion

Physical Chemistry 

Annex 1

inhalation/ingestion

Physical Chemistry 

Annex 2

inhalation/ingestion

September 1945–December 1953 N/A 17.5/1.6 N/A

January 1954–May 1976 4.1/0.68 N/A 124.7/11.2 (only 

through 1972), then 0a

a. Physical Chemistry Annex 2 was razed in 1972.

5.1.1.4 Uranium Excretion Plots for Use with Bioassay Data

Figures 5-1, 5-2, and 5-3 are plots for expected urinary excretion assuming chronic intakes at the 

highest intake values (scrap recovery) for both inhalation and ingestion, for chronic intake periods of 

365, 730, and 1,826 days, respectively. If bioassay results are available for a worker, they can be 

compared to the excretion curves in the figures if efficiency techniques are being used; for instance, 

for a noncompensable efficiency case based on the default intakes, the Energy Employee’s actual 

bioassay should be less than the applicable curve in the figures [18]. Types F, M, and S are defined 

as fast, medium, and slow, respectively, in the figures.

Figure 5-1. Urinary excretion of uranium expected from 365-day

chronic intake at 6,061 pCi/d inhalation and 124 pCi/d ingestion.

Figure 5-2. Urinary excretion of uranium expected from 730-day

chronic intake at 6,061 pCi/d inhalation and 124 pCi/d ingestion.

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

10 100 1000 10000

Days Since Start of Chronic Intake

Excretion, pCi/d 

Type F

Type M

Type S

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

10 100 1000 10000

Days Since Start of Chronic Intake

Excretion, pCi/d 

Type F

Type M

Type S
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 39 of 90

Figure 5-3. Urinary excretion of uranium expected from 1,826-day

chronic intake at 6,061 pCi/d inhalation and 124 pCi/d ingestion.

5.2 THORIUM EXPOSURE FROM THORIUM CONTAMINATION AFTER 1954

As discussed in Section 1.3, NIOSH determined that it was not feasible to complete dose 

reconstruction for internal dose for thorium, plutonium, and thoron for 1942 through 1954 (NIOSH 

2006) and, therefore, internal dose is not addressed here for those radionuclides during those years. 

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 plutonium after 1954 would have been significant.

However, Metallurgy Building (Wilhelm Hall) was the location of the main activities with thorium and is 

still in use. As discussed in Section 1.3, it is not feasible to determine internal doses for sheet metal 

workers, physical plant maintenance and associated support staff (includes all maintenance shop 

personnel of Ames Laboratory), and supervisory staff who performed maintenance and renovation 

activities in Wilhelm Hall (Old Metallurgy Building) during the period of January 1, 1955, through 

December 31, 1970. This method can be used, however, to provide an estimate of internal 

exposures for others who performed work in this building for non-SEC claims. 

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 through the early 1990s, surveys were conducted in Wilhelm Hall to determine 

locations of contamination left from the early production years. This information is in Hokel et al. 

(1998). The following discussion using data from Hokel et al. and Klevin provides an estimate for 

intakes by workers in Wilhelm Hall from 1955 to the present. 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. There was one set 

of data that had floor surveys made in 1988, but these locations all had fixed activity, indicating that 

this was from beta radiation (see Appendix 6 of Hokel et al.). For this reason, the recommendation 

for assessing dose from thorium from 1954 through the present was provided. The accessible areas 

of the building, including rooms, air ducts, hallways, stairwells, transformer rooms, etc., were 

surveyed starting in 1984. 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 sub-basement 

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

10 100 1000 10000

Days Since Start of Chronic Intake

Excretion, pCi/d 

Type F

Type M

Type S
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 40 of 90

pipe tunnels and in large vertical void spaces in the stairwells. All of these 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 high for generally accessible areas and 

with the possibility of the contamination becoming resuspended. Although other survey numbers in 

the report with removable contamination are higher, they are in locations that are inaccessible (in a 

drain line), hard to reach (inside a drawer), or small (edge of a sink), or the contamination was fixed.

Table 5-5. Air concentrations and daily intakes of 232Th at Wilhelm Hall 

after 1954.

Year t (yr)

Air 

concentration 

(dpm/m3)

Chronic 

inhalation 

intake (pCi/d)

Chronic 

ingestion intake 

(pCi/d)

1955a 0.5 448 1330 40.4

1956a 1.5 392 1160 35.3

1957a 2.5 343 1020 30.9

1958a 3.5 300 887 27.0

1959a 4.5 262 776 23.6

1960a 5.5 229 679 20.6

1961a 6.5 200 594 18.1

1962a 7.5 175 519 15.8

1963a 8.5 153 454 13.8

1964a 9.5 134 397 12.1

1965a 10.5 117 347 10.6

1966a 11.5 103 304 9.24

1967a 12.5 89.7 266 8.08

1968a 13.5 78.4 232 7.07

1969a 14.5 68.6 203 6.18

1970a 15.5 60.0 178 5.40

1971 16.5 52.5 155 4.73

1972 17.5 45.9 136 4.13

1973 18.5 40.1 119 3.61

1974 19.5 35.1 104 3.16

1975 20.5 30.7 90.9 2.76

1976 21.5 26.8 79.5 2.42

1977 22.5 23.5 69.5 2.11

1978 23.5 20.5 60.8 1.85

1979 24.5 18.0 53.2 1.62

1980 25.5 15.7 46.5 1.41

1981 26.5 13.7 40.7 1.24

1982 27.5 12.0 35.6 1.08

1983 28.5 10.5 31.1 0.946

1984 29.5 9.18 27.2 0.827

1985 30.5 8.03 23.8 0.724

1986 31.5 7.02 20.8 0.633

1987 32.5 6.14 18.2 0.553

1988 33.5 5.37 15.9 0.484

1989 34.5 4.70 13.9 0.423

1990 35.5 4.11 12.2 0.370

1991 36.5 3.59 10.6 0.324

1992 37.5 3.14 9.31 0.283

1993 38.5 2.75 8.14 0.248

1994 39.5 2.40 7.12 0.217

1995 40.5 2.10 6.23 0.189

1996 41.5 1.84 5.45 0.166

1997 42.5 1.61 4.76 0.145
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 41 of 90

Year t (yr)

Air 

concentration 

(dpm/m3)

Chronic 

inhalation 

intake (pCi/d)

Chronic 

ingestion intake 

(pCi/d)

1998 43.5 1.41 4.17 0.127

1999 44.5 1.23 3.64 0.111

2000 45.5 1.08 3.19 0.0969

2001 46.5 0.941 2.79 0.0847

2002 47.5 0.823 2.44 0.0741

2003 48.5 0.719 2.13 0.0648

2004 49.5 0.629 1.86 0.0567

2005 50.5 0.550 1.63 0.0496

2006 51.5 0.481 1.43 0.0434

2007 52.5 0.421 1.25 0.0379

2008 53.5 0.368 1.09 0.0332

2009 54.5 0.322 0.953 0.0290

2010 55.5 0.282 0.834 0.0254

a. Dose reconstructors should not use data from 1955 to 1970 to determine internal 

doses for employees affected by SEC-00075 (NIOSH 2007a).

Hokel et al. (1998) stated that the thorium in Wilhelm Hall is in equilibrium.

For estimating dose to individuals working in Wilhelm Hall from 1955 to the present, 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 [19]. This means the 232Th activity was approximately 200 dpm/100 cm2

. Applying a 

resuspension factor of 10-4 [20]:

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, maybe minor moving of 

equipment or work benches followed by cleaning, etc. This method of internal dose estimation does 

not apply to workers in the buildings or areas designated in SEC-00075 (NIOSH 2007a).

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: 

1. 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. 

2. The general air concentrations included one room that was 10 times higher than all the

 others and included the lunch room. The highest room was clearly related to operations so

Air concentration = (200 dpm/100 cm2)(10-4/m )(100 × 100 cm2/m2) = 2 dpm/m3
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 42 of 90

that datum was excluded, as was the lunch room, 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 Ao and 2 as A(t) and assuming t is from January 1, 1955, to 

November 15, 1995, for a total of 14,929 days,

A(t) was solved for the midpoint of each year from 1995 to the present. The air concentration was 

converted to intake using

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 1955 to the present. 

The dose reconstructor needs to include equal intakes of 228Ra and 228Th to account for the progeny 

radionuclides [21]. The associated ingestion intakes would be: 

The annual 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 [22]. 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 provided in Table 5-5 [23]. However, this assumption is not valid for employees who worked

in the buildings or areas described in SEC-00075 (NIOSH 2007a).

5.3 TRITIUM EXPOSURE

The ALRR operated from February 1965 through December 1977. The reactor was then 

decommissioned, which was completed in 1981. The ALRR was a 5-MW, heavy-water-moderated 

research reactor. 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

A number of 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 (Voss 1971)

describes potential sources of tritium exposure to Laboratory personnel. 

A(t) = Aoexp(-λt). λ = 3.7 × 10-4/d or 0.134/yr.

intake (pCi/d) = (air conc. dpm/m3)(1 pCi/2.22 dpm)(1.2 m3/hr)(2,000 hr/yr)/365d/yr)

(0.2)(air concentration in dpm/m3) ÷ 2.2 dpm/pCi = X pCi/d each for 232Th, 228Ra, and 228Th.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 43 of 90

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 the 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 

on a semiannual basis 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 on the basis of their 

work with sources of tritium).

6. All personnel who showed tritium concentration greater than 1,500 dpm/min/mL of urine were 

resampled weekly until the detected activity level was below 1,500 dpm/min/mL of urine.

7. Samples were taken at the termination of employment of persons assigned to work areas 

where monitoring for tritium was required.

8. As a check and control, termination samples were taken from other personnel (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.

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. Thus, 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.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 44 of 90

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 assignment of dose that were discussed in Voss (1971) are:

1. The dose assignment system had two separate conditions of radiation dose delivery to the 

body: (a) 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 (b) the radiation 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/min/mL in urine, sampling 

continued until the urine sample level was less than 1,500 dpm/min/mL.

3. For any person on the routine and incident monitoring program, a basic dose assignment was 

made. This basic dose was that associated with the maintenance of a level of 

750 dpm/min/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 on the basis of the effective 

half-life 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.

A number of other documents discussed tritium dose assessment, but they are similar to the 

information provided above. “Personnel Monitoring Program” (Ames 1974) provides the Tritium Dose 

Calculation Procedures used at the time.

Personnel in the tritium monitoring program should have bioassay and dose records. Tritium 

bioassay records were found for 1965 through 1981, although those for 1965 through 1968 appear to 

be incomplete.

Figures 5-4 and 5-5 show examples of what a bioassay record in a worker’s file might look like. The 

data that can be found 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 employee.).

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 were operations that were performed and sampling 

was based on potential for intake.

5. Date that individual was employed.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 45 of 90

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 5-4 is for the first quarter and shows the 

dose for quarter 1. For each subsequent quarter, 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 5-4. Bioassay record example 1. 

Figure 5-5. Bioassay record example 2. 

Documents from the early 1970s (Voss 1971; Ames 1974) indicate that a person in the tritium 

monitoring program would have been assigned a default radiation dose of 38 mrem/yr (or it would 

have been prorated for the time monitored during the year); however, a review of the available dose 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 46 of 90

records indicated that the policy to assign a 38-mrem/yr dose to everyone on the tritium program was 

not always followed.

No information was found that indicated the analysis method or the minimum detectable activity 

(MDA); therefore, it is recommended that 0.1 μCi/L (220 dpm/mL) be used as the MDA. This is a 

reasonable assumption compared with MDAs at other sites during this period and is consistent with 

the way the data were recorded [24].

5.3.2 Tritium Dose to Workers, 1982 to Present

An interoffice memorandum (Voss 1987) 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). 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 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-hour workweek (Struss 1987). 

The dose reconstructor can consider assigning a dose of 8.6 mrem/yr for personnel working in the 

Applied Science Center in either the reactor room or the pump room if there are no bioassay data 

available.

5.3.3 Tritium Dose to Unmonitored Workers, 1965 Through 1981

A coworker analysis of worker tritium bioassays was conducted on bioassay samples from 1965 

through 1981 to provide default doses for unmonitored workers. Results of the coworker analysis are 

listed in Table 5-6, columns 2 and 3. If there is an indication that an Energy Employee worked in the 

ALRR during 1965 through 1981 but was not monitored for tritium, the annual doses listed in 

Table 5-6, column 4, should be assigned. These doses apply to all organs. Assume a lognormal 

distribution with the geometric standard deviations (GSDs) as listed in Table 5-6, column 5. 

Table 5-6. Results of tritium coworker analysis [25]. 

Year

Coworker results Assigned values

Annual dose (mrem) GSDa Annual dose (mrem) GSDa

1965 0.44 1.6 0 N/A

1966 0.54 1.5 1 3

1967 0.46 1.9 0 N/A

1968 0.23 2.3 0 N/A

1969 1.9 7.9 1.9 7.9

1970 0.88 6.9 1 6.9

1971 1.3 15 1.3 15

1972 1.3 12 1.3 12

1973 1.2 18 1.2 18

1974 1.9 8.1 1.9 8.1

1975 1.3 17 1.3 17

1976 1.0 9.5 1.0 9.5

1977 1.6 25 1.6 25

1978 1.7 11 1.7 11

1979 1.0 7.5 1.0 7.5

1980 1.0 9.8 1.0 9.8

1981 1.1 4.7 1.1 4.7

a. Use a GSD of 3.0 if the calculated value is less than 3 in Table 5-6.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 47 of 90

The tritium coworker doses were determined using methods described in Analysis of Coworker 

Bioassay Data for Internal Dose Assignment (ORAUT 2005a) and Tritium Calculated and Missed 

Dose Estimates (ORAUT 2004). Each annual dose was the geometric mean of the lognormal 

distribution fitted to the distribution of individual workers‘ annual doses. Figure 5-6 shows an example 

distribution and fit for 1970. Annual doses less than 1 mrem are insignificant and can be ignored 

when assigning doses; a minimum GSD of 3.0 was assigned if the calculated value was less than 3. 

5-6. Distribution of worker doses from tritium in 1970.

5.4 FISSION PRODUCT INTAKES 

5.4.1 Fission Product Intakes from Early Fuel Research

As described in Section 4.4, 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 means of ion exchange columns. 

A hot laboratory, capable of handling 5 curies through the adsorption process, was 

designed and built. The method proved successful in a number of runs. The uranium 

and plutonium were separated from one another and from the fission products using 1 

kg samples of uranium from the pile, which had an activity of 5 curies. (Fulmer 1947) 

Intakes by workers in the hot laboratory were estimated using the approach for calculating intakes 

described in NUREG-1400, Air Sampling in the Workplace (Hickey et al. 1993). The equation from 

NUREG-1400 is:

where

I = intake

Q = source term for 1 year = assumed to be 50 Ci/yr

R = release fraction = 0.01 for a liquid (the material comes in as a solid, but would be liquefied 

for research activities)

C = confinement factor = 0.1 assuming material was handled in some containment

I = Q × 10-6 × R × C × D
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 48 of 90

D = dispersibility factor = 10 for heating or chemical reactions

As described in Section 4, using the recommended fission product ratios for 180-day cooled fuel from 

Fission and Activation Product Assignment for Internal Dose-Related Gross Beta and Gross Gamma 

Analyses (ORAUT 2007a), the 1,370-pCi/d intake of fission products is assigned to specific 

radionuclides, as listed in Table 5-7.

Table 5-7. Default annual fission product intakes 

for researchers from hot laboratory in the 

Chemistry Building (1943–1951).

Radionuclide

Relative 

fraction

Intake 

(pCi/d)

Ce-141 0.0221 30.3

Ce-144 0.2191 300

Cs-134 0.0054 7.4

Cs-137 0.0208 28.5

Eu-155 0.0014 1.9

Fe-55 0.0172 23.6

Nb-95 0.2492 341

Pm-147 0.0546 74.8

Ru-103 0.0321 44

Ru-106 0.0844 115.6

Sr-89 0.0558 76.4

Sr-90 0.0157 21.5

Y-91 0.0911 124.8

Zr-95 0.1311 179.6

Because maximizing assumptions were used for most of the parameters in this analysis, the 

distribution is an upper bound (constant). The absorption type for each radionuclide that is most 

favorable to claimants, as listed in ICRP Publication 68 (ICRP 1995), should be used, with the 

exception that only type F should be used for strontium [26]. The date of first use of the hot laboratory 

was not found. Because it received fuel from the Chicago Pile, the earliest reasonable use would 

have been 1943. The hot laboratory was replaced by a hot cell in the Research Building in 1951. 

Therefore, these intakes should be applied for each year from 1943 through 1951. As discussed 

above, SEC-00166 (NIOSH 2010b) states that intakes from the hot cell cannot be determined for 

radionuclides other than uranium. Due to the smaller amounts of material and the improved 

ventilation and containment of the Research Building over its predecessor discussed in Sections 2 

and 4, respectively, workers in the Research Building who did not work with the hot cell or laboratories 

associated with the hot cell should be assigned environmental internal and external doses.

5.4.2 Fission Product Intakes from Research Reactor Operations and Decontamination 

and Decommissioning

There was no evidence 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 (Voigt 1981) indicated that there was a 

negligible amount of fission products in the building after shutdown. 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.

I = 50 × 10-6 × 0.01 × 0.1 × 10 = 5 × 10-7 Ci/yr = 5 × 105 pCi/yr

= 5 × 105 pCi/yr/365 d/yr = 1,370 pCi/d
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 49 of 90

During D&D operations, radiation protection appeared to be acceptable for the time (Voigt 1981) and 

external exposure and internal exposure for tritium were tracked. There is no evidence that bioassay 

for other radionuclides was performed. However, during D&D of the reactor, which included 

dismantlement, cutting, grinding, etc., it can be expected that some intakes from activation products 

occurred.

A reasonable estimate of intakes from D&D of the reactor can be made by using the gross beta air 

concentration limit from 1977 (1 × 10-9 μCi/mL) (ERDA 1977) based on the most conservative beta 

emitter (90Sr).

For workers involved in the actual decontamination or demolition, we assumed:

Assuming an occupancy factor of 0.5 (1,000 hr/yr exposure) and a breathing rate of 1.2 m3

/hr.

Assign this intake to the D&D workers for the appropriate years between 1978 and 1981. For 

supervisors and others not directly involved in the work, assign one-fourth of the intake.

5.5 SUMMARY OF INTERNAL DOSE RECOMMENDATIONS

Default occupational intakes for workers without the applicable bioassay data are summarized in 

Table 5-8.

(1 × 10-9 μCi/mL) (1 × 106 pCi/μCi) (1 × 106 mL/m3

) = 1,000 pCi/m3

(1,000 pCi/m3) (1.2 m3

/hr) (1,000 hr/yr) ÷ 365 d/yr = 3,300 pCi/d.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 50 of 90

Table 5-8. Default intakes for Ames Laboratory (if no bioassay results are available) [27].

Job category 

or task/Building

Dose calculation parameters IREP input parameters

Period Material Mode

Absorption 

type or f1 pCi/d

Distribution 

type

Parameter 

1

Researcher/Chemistry Buildinga Aug 1942–

Dec 1953

Natural 

uranium

Chronic 

inhalation

F, M, or S 8.5 Constant Dose

Researcher/Chemistry Buildinga Aug 1942–

Dec 1953

Natural 

uranium

Chronic 

ingestion

0.02 with F,M; 

0.002 with S

0.09 Constant Dose

All employees in Chemistry Building Jan 1954–

May 1976 

Natural 

uranium

Chronic 

Inhalation

F, M, or S 4.1 Constant Dose

All employees in Chemistry Building Jan 1954–

May 1976

Natural 

uranium

Chronic 

Ingestion

0.02 with F,M; 

0.002 with S

0.68 Constant Dose

Researcher in hot lab/Chemistry Building 1943–1951 Fission 

products

Chronic 

inhalation

Use most favorable 

to the claimant; for 

Sr-90 use F

See Table 5-7 Constant Dose

Researcher, production technician, anyone 

involved daily with uranium in Annex 1a Aug 1942–Aug 1945

Natural 

uranium

Chronic 

inhalation

F, M, or S 853 Constant Dose

Researcher, production technician, anyone 

involved daily with uranium in Annex 1a Aug 1942–Aug 1945

Natural 

uranium

Chronic 

ingestion

0.02 with F,M; 

0.002 with S 

8.7 Constant Dose

All employees in Annex 1 Sep 1945–

Dec 1953

Natural 

uranium

Chronic 

inhalation

F, M, or S 17.5 Constant Dose

All employees in Annex 1 Sept 

1945–

Dec 1953

Natural 

uranium

Chronic 

ingestion

0.02 with F,M; 

0.002 with S 

1.6 Constant Dose

Researcher, production technician, anyone 

involved daily with uranium in Annex 2a Jan 1944–Dec 1953

Natural 

uranium

Chronic 

inhalation

F, M, or S 6,061 through 

1950, 5,556 from 

1951 to 1953

Constant Dose

Researcher, production technician, anyone 

involved daily with uranium in Annex 2a Jan 1944–Dec 1953

Natural 

uranium

Chronic 

ingestion

0.02 with F,M; 

0.002 with S

124 through 1950; 

114 from 1951 to 

1953

Constant Dose

All employees in Annex 2 Jan 1954–

1972

Natural 

uranium

Chronic 

inhalation

F, M, or S 124.7 Constant Dose

All employees in Annex 2 Jan 1954–

1972

Natural 

uranium

Chronic 

ingestion

0.02 with F,M; 

0.002 with S

11.2 Constant Dose

Anyone routinely in Wilhelm Hall 

(Metallurgy Building)

1955–

present

Th-232 Chronic 

inhalation

M or S See Table 5-5 Constant Dose

Anyone routinely in Wilhelm Hall 1955–

present

Th-232 Chronic 

ingestion

5E-4 with M; 2E-4 

with S

See Table 5-5 Constant Dose

Anyone routinely in Wilhelm Hall 1955–

present

Ra-228 Chronic 

inhalation

M See Table 5-5 Constant Dose

Anyone routinely in Wilhelm Hall 1955–

present

Ra-228 Chronic 

ingestion

0.2 See Table 5-5 Constant Dose
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 51 of 90

Job category 

or task/Building

Dose calculation parameters IREP input parameters

Period Material Mode

Absorption 

type or f1 pCi/d

Distribution 

type

Parameter 

1

Anyone routinely in Wilhelm Hall 1955–

present

Th-228 Chronic 

inhalation

M or S See Table 5-5 Constant Dose

Anyone routinely in Wilhelm Hall 1955–

present

Th-228 Chronic 

ingestion

5E-4 with M; 

2E-4 with S

See Table 5-5 Constant Dose

Reactor workers and D&D workers 1965–1981 Tritium Total of all 

modes

N/A N/A Lognormal, 

see Table 5-

6 for GSD

Dose; see 

Table 5-6

Workers involved with D&D of reactor, 

including former reactor workersa 1978–1981 Co-60 or Zn-65

Chronic 

inhalation

Type S 3,300 Constant Dose

Workers involved with D&D of reactor, 

including former reactor workersa 1978–1981 Co-60 or Zn-65

Chronic 

ingestion

Type S 660 Constant Dose

Anyone routinely in Applied Science Center 1982–

present

Tritium Total of all 

modes

N/A N/A Constant 8.6 

mrem/yr

a. Values are for workers assumed to work in research or production full time. For supervisors, assume 0.25 of the intake; for all other employees (clerical, janitorial, 

security, etc) assume one-tenth of the supervisor’s intake (0.025 of the intake in the table).
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 52 of 90

6.0 OCCUPATIONAL EXTERNAL DOSE

Workers at Ames Laboratory received external radiation doses between 1942 and 1952 that were 

largely unmonitored [28]. Pocket chambers were available that might have been used to monitor 

external doses, but very few records could be found [29]. 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 through 1981) when individual 

names were not recorded with dosimeter numbers and many records cannot be clearly associated 

with individual workers [30]. 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 of 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 Excel® 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 

(Martin 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 

Martin (2006c) should be used if there is any doubt about completeness. 

As described in Section 1.3, some portions of external dose cannot be adequately estimated for 

certain workers from August 13, 1942, through December 31, 1954 (NIOSH 2006). 

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 to be less than adequate. A number of 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 was started in February 1953; beta/gamma film badges were exchanged 

on a biweekly frequency with the results tabulated monthly and summarized annually (Ames 1954a). 

Dosimetry services evolved and improved over the following 50 years; the characteristics of the 

dosimeters used are summarized in Table 6-1.

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, through 1952 when film badge service was begun. However, the records of pocket 

chamber results that were found apply only to a brief period in February 1949 and only for a few 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 53 of 90

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

Dosimeter type, 

provider Period

Exchange 

frequencya

MRD (mrem)b MDL (mrem)c

Skin β/γ deep Neutron Skin Deep Neutron

Pocket Chambers, 

Ames Laboratory

1942–1952d Daily 5 5

β/γ film, Ames 

Laboratory in-house 

system

Feb 1953–May 1953 Biweekly 10 40 40

June 1953–Dec 1953 25 40 40

Jan 1954–Feb 1957 25 25 40 40

Mar 1957–Dec 1961 10 10 40 40

β/γ/NTA film, BNL Apr 1954–June 1957 Biweekly 15 15 10 40 40 (e)

β/γ/NTA film, NCA July 1957–June 1963 Biweekly 10 10 10 40 40 (e)

β/γ/NTA film, Atomic 

Film Badge 

Corporation

July 1963–Mar 1965 Monthly 10 10 10 40 40 (e)

Pocket Chambers, 

Ames Laboratory

Apr 1965–June 1965 Daily 5 5

β/γ/NTA film, Health 

Physics Services 

July 1965–Oct 1979 Monthly 10 5 28 40 40 (e)

β/γ film, Health 

Physics Services

Nov 1979–Sept 1981 Monthly 10 5 40 40

β/γ TLDs, Landauer June 1980–Dec 1981 Monthly & 

quarterly

40 10 30 30

β/γ TLDs, Landauer Jan 1982–Dec 1994 Quarterly 40 10 30 30

β/γ TLDs, Siemens Jan 1995–June 1996 Quarterly 10 10 30 30

β/γ TLDs, ICN 

Dosimetry Service

July 1996–Sept 1998 Quarterly 10 10 30 30

β/γ TLDs, Landauer Oct 1998–Dec 2004 Quarterly 40 10 30 30

β/γ TLDs, Global 

Dosimetry

Jan 2005–present Quarterly 10 10 30 30

a. The exchange frequency was established from dosimetry reports. 

b. MRD = minimum recordable dose; based on minimum doses recorded on dosimetry reports.

c. Estimated minimum detection level (MDL) typical of film dosimeter capabilities (Wilson 1960, 1987; NIOSH 1993; NRC 

1989; Wilson et al. 1990).

d. Exposures from August 13, 1942, through December 31, 1952, that resulted from thorium, its progeny, and plutonium 

cannot be reconstructed for workers who worked in buildings or areas included in SEC-00038 (NIOSH 2006). However, 

external dose from uranium can be estimated from 1942 to 1953 (Section 6.3.1).

e. For years of NTA film use, between 1954 and 1979, the adjusted neutron dose is calculated using correction factor of 2.

individuals (Voss 1949). Thus, it is only possible to estimate external doses before 1953. External 

doses at Ames Laboratory before 1953 are discussed in Section 6.3.1.

An upper bound of the total external exposures cannot be made for SEC-00038 workers because 

external doses from beta and gamma radiation resulting from exposure to thorium and its daughters 

or plutonium cannot be reconstructed due to a lack of information on the percentage of thorium 

daughter in-growth (up to 1954, when thorium operations ended) (NIOSH 2006). However, external 

dose from potential exposure to uranium from 1942 to 1953 can be estimated.

An in-house film badge system was established at Ames Laboratory in the fall of 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 through March 1953, but records are incomplete (Ames

1954a). Regular film badge service began in February 1953 for 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 1954a). 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 period in 

February, and a 4-week period in December 1954; otherwise, the exchange periods were 2 weeks 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 54 of 90

each (Ames 1954b, 1955). The exchange frequency during 1955 and 1956 was every 2 weeks, with 

no exceptions (Voss 1956a,b). There was a 4-week exchange period in May and June 1957; all other 

exchange periods in 1957 were 2 weeks (Ames 1958). 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 1959, 1960b). The exchange frequency during 1960 was every 2 weeks without exception 

(Ames 1961b). During 1961, there were nine 3-week exchange periods and 12 2-week periods 

(Ames 1962b). 

Initially, the minimum recordable dose (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 1957b, 1958, 1959, 1960b, 1961b, 1962b). 

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 1954a). Corrections that 

are favorable to claimants for this exposure that was not received by personnel are recommended in 

Martin (2006c). 

In the fourth quarter of 1953 and continuing through 1961, beta and gamma doses were reported 

separately, but were added to give a total dose (Ames 1954a). Between 1954 and 1961, neutron 

doses were added to the beta and gamma doses to give a total dose (Ames 1954a,b, 1955, 1956a,b, 

1958, 1959, 1960b, 1961b, 1962b). 

In the annual summaries for 1953 through 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 (Voss 1954a,b, 1955, 1956a,b,). From March 1957 through 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 1958, 1959, 1960b, 1961b, 1962b). 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 through 1961.

Beginning in April 1954 and continuing through June 1957, a beta/gamma/neutron film badge service 

was provided by BNL for the synchrotron staff (Ames 1954b, 1955). 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 dose 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 1954b, 1955, 1962b).

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 1958). 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 dose 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 1958).

The data discussed above for the period from 1953 through 1961 were compiled on a spreadsheet 

and the inconsistencies were corrected or minimized (Martin 2006c). For example, if beta, gamma, 

and neutron doses were 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 would be favorable to claimants [31].
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 55 of 90

Dosimetry recordkeeping from January 1962 through June 1963 was less than adequate. The only 

records found for 1962 were some calibration data from April and the annual summary data (Ames

1963). It was assumed that the in-house film badge system continued through December 1962 with 

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 through March 

1965. No records were found for the first half of 1963. The annual summary for 1963 reflects results 

from July through December 1963 only; therefore, all Ames Laboratory workers are assumed to have 

been unmonitored for the first half of 1963 (Ames 1964). The annual summary for 1964 accurately 

reflected the sum of the monthly dosimeter readings during 1964 (Voss 1965). 

Dosimetry records for January through 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 members were monitored with pocket chambers on a daily basis and 

records were compiled in-house from April through June 1965 (Ames 1966). Film badge service was 

provided by Health Physics Services from July 1965 through September 1981. The monthly service 

included beta, gamma, and neutron film badges for reactor and synchrotron staff members and 

beta/gamma film badges for the remaining Laboratory personnel (Ames 1966). The MRDs were 

10 mrad for beta, 5 mR for gamma, and 28 mrem for neutrons (Ames 1966).

TLD services for beta and gamma dosimetry were provided by Landauer from June 1980 through 

June 1995. Similar services have been provided in subsequent years by Siemens, ICN Dosimetry 

Service, 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; 

Siemens 1996; ICN 1997). Dosimetry records compiled since 1981 are considered to be 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).

6.2.2 Dosimetry Technology

Ames Laboratory used film badge dosimeters from 1952 through 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 used 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 (ORNL) multielement dosimeter; the neutron film was nuclear track emulsion type 

A (NTA) (ORAUT 2006b). Similar film badge holders and film were provided by NCA, Atomic Film 

Badge Corporation, and Health Physics Services 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 [32]. From 1980 to 

the present, beta/gamma TLD service has been provided by Siemens, ICN Dosimetry Service, and 

Landauer for the periods listed in Table 6-1. These services have been accredited by NVLAP since 

1985 (Landauer 1985).
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 56 of 90

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, Goldstein, and Wollan 1944) and one for a sensitive 

DuPont 555 emulsion in the multielement badge (Thornton, Davis, and Gupton 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 over-response 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, Goldstein, and Wollan 1944) and sensitive DuPont 

555 emulsion in multielement film badge (Thornton, Davis, and 

Gupton 1961).

6.2.2.2 Neutron Dosimeters

NTA film was used by the dosimetry services provided to Ames Laboratory from 1954 through 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 (IAEA 1990; ORAUT 2006b). 

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 [33]. The suppliers of neutron dosimeters were 

generally relied on for neutron calibrations.

10-2 10-1 10 |0

Photon Energy (MeV)

0.1

1

10 Relative Response

Hp(10)

Two-Element Film Badge

Multi-Element Film Badge
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 57 of 90

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 through 1961 (Ames 1974). The 

beta calibration was accomplished by the exposure of 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 the exposure 

of films to doses of gamma radiation from a 10-mg 226Ra source that ranged between 10 and 1,000 

mR (Ames 1954b, 1955). 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 1963). From June 1963 to March 1965, when the Atomic Film Badge Corporation 

provided film badge services, monthly gamma calibrations were done by the exposure of films to 

doses of gamma radiation from a 10-mg 226Ra source that ranged from 100 to 10,000 mR (Ames

1964, 1965). There was no evidence of beta or neutron calibrations during this period. 

From July 1965 through September 1981, when film badge services were provided by Health Physics 

Services, a set of calibration control film badges were exposed to beta and gamma sources with each 

biweekly set of personnel film badges (Ames 1966, 1971, 1974; Ames/HPS 1978, 1979, 1981). The 

calibration procedures were the same as those used during the period from 1953 through 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 through June 1957, when film badge services were provided by BNL, a set of four 

calibration badges was included in each monthly set of films (Voss Ames 1954b, 1955, 1956a,b). The 

calibration badges were exposed to known doses of beta, gamma, and neutron radiation. The beta 

exposures were to a milled uranium plate, the gamma exposures were to a 10-mg 226Ra source, but 

records of the source of neutron exposures were not found (Voss Ames 1954b, 1955, 1956a,b).

From July 1957 through 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 1958, 1959, 1960b, 

1961b, 1962b, 1963, 1964). 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 1961b). From June 1963 through 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 Health Physics Services 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 rose from the production of uranium and thorium metal, 

the operation of the synchrotron and ALRR, and research activities in laboratories. With few 

exceptions, the following sections show that, for external dose reconstruction, all beta radiation fields 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 58 of 90

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 [34]. Assuming that 100% of the radiation fields are within these 

ranges is a simplifying conservative assumption that is generally 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, Slaback, and Birky 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, 

Slaback, and Birky 1998). 

A large number of beta-emitting radioisotopes were handled in research work at Ames Laboratory 

throughout its history [35]. 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 that occurred at the ALRR were monitored by 

urine bioassay, which is addressed in Section 5 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 from 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 gammaemitting radioisotopes 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 generally favorable to claimants [36]. 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. Neutron doses cannot be 

reconstructed for the SEC-00038 class of employees from 1942 through 1954 (NIOSH 2006). As 

stated in Section 1.3, this SEC class has been expanded to include all Ames employees, contractors, 

and subcontractors.

The primary sources of neutrons at Ames Laboratory were the synchrotron, which operated from 1949 

through 1971, and the ALRR, which operated from 1965 through 1977. In both facilities, fast, 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 59 of 90

intermediate, and thermal neutrons would have been present in the workplace [37]. 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 evidence of neutrons in the workplace is the many nonzero fast neutron doses measured 

with NTA film worn by personnel at each facility (Martin 2006c). Assuming that 100% of the neutrons 

were in the 0.1-to-2-MeV energy range is a simplifying conservative assumption that is generally 

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 and 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 [38]. 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 (Ames 1974). 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 that expected from the source terms 

and personnel monitoring described in Section 2.4. The recorded beta and gamma doses can be 

considered reliable for dose reconstruction [39].

6.2.5.2 Beta/Photon TLD 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).

6.2.5.3 Neutron Dosimeter Response

The neutron doses received by synchrotron staff members 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 2010a). The higher neutron doses at ALRR might have been due to less 

than adequate management of neutron shielding in the experimental areas [40]. 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 the 

Stanford Linear Accelerator Center (SLAC) was determined to be similar to the 80-MeV synchrotron 

at Ames Laboratory in relation to neutron exposures (ORAUT 2007a). The Materials Test Reactor 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 60 of 90

(MTR) at the Idaho National Laboratory (INL) was determined to be similar to the ALRR with respect 

to neutron exposures (ORAUT 2010c). Neutron dosimeters with NTA film were used at both facilities 

and neutron spectral data are available for estimation of 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 (2006c) 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 

but their neutron spectra were not identical [41]. To allow for differences in the neutron spectra and 

other variables, a neutron dose correction factor of 2 is recommended for Ames synchrotron workers 

[42]. 

The MTR at INL, which operated from 1952 to 1970, was similar to the ALRR in that it was fueled with 

enriched uranium and it had a power level of 30 MW (ORAUT 2010c). 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 (2010a). 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 2006c). 

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 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, as discussed in NIOSH (2006). 

Table 6-2, which is from ORAUT (2006d), 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 AMES LABORATORY WORKER EXTERNAL DOSE 

RECONSTRUCTION 

Dose reconstruction for Ames Laboratory workers is based on the information above, which requires 

assessment of additional dose to 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

• 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. 03 Effective Date: 01/03/2012 Page 61 of 90

6.3.1 Unmonitored External Dose

At Ames Laboratory, the concept of unmonitored worker will have to be expanded to include 

monitored but records not found [43]. Unmonitored doses were determined in two periods by two 

different methods. Previous work on exposures to workers at AWE uranium facilities (Battelle 2011,

Christofano and Harris 1960, and ORAUT 2010d) was used to calculate unmonitored doses before 

1953, while coworker dose based on a review of Ames Laboratory dosimetry data was used to 

construct coworker doses beginning in 1952. 

For unmonitored doses in 1952 and 1953 (for which both methods provide unmonitored doses), the 

dose reconstructor should choose the method that best represents the employee’s exposure based 

on work location and job description.

6.3.1.1 Unmonitored Doses Before 1953

Before the use of film badges, which began in 1953, external doses at Ames Laboratory were 

monitored using pocket dosimeters, but not generally recorded. Minimal records were found of these 

pocket dosimeter results (Voss 1949). To estimate reasonable external doses from the uranium 

processes before 1953, the methods described in three documents (Battelle 2011, Christofano and 

Harris 1960, and ORAUT 2010d) were used. These documents are representative of the potential 

external exposures encountered at Ames Laboratory because the processes developed there were 

similar to processes at AWE sites.

Two processes were considered to be the most similar to the processes in use at Ames Laboratory; 

Scrap Recovery in Battelle (2011, Section 2.1) is similar to processes in use in Physical Chemistry 

Annex 2, and Metal Reduction in Christofano and Harris (1960) is similar to processes in Physical 

Chemistry Annex 1 and the Chemistry Building.

The operation and processes of these three facilities are described in Section 2. However, there is 

not enough information to develop specific exposure potential parameters to the occupations in these 

facilities from 1942 through 1953. It is safe to assume that all workers in these facilities received 

exposures directly from the material being used and from airborne and surface contamination in the 

facilities. Battelle (2011) and Christofano and Harris (1960) provide estimates of exposure for typical 

scenarios that were similar in many AWE uranium facilities. These estimates are based on several 

assumptions that were designed to determine reasonable exposures for workers in these categories: 

operators (production workers), general laborers, supervisors, and clerical. For this site profile, 

security and janitorial personnel have been placed in the clerical category. Ames Laboratory had a 

number of personnel with occupations listed as researcher, junior researcher, research technician, 

etc. It is assumed that these research workers were exposed to doses that were one-tenth those of 

the production workers [44].

Exposure levels in these documents were converted to annual doses. These doses are provided for 

the Chemistry Building, Physical Chemistry Annex 1, and Physical Chemistry Annex 2 in Tables 6-4, 

6-5, and 6-6 below. External doses from submersion in airborne contamination were less than 0.001 

rem/yr and, therefore, were insignificant. Doses from surface contamination on the floor were less 

than 0.001 rem/yr in Battelle (2011) and, therefore, these doses were not included in Annex 2 doses 

(Table 6-6). In addition, doses in Battelle (2011) were slightly less after 1950 than in the years before. 

Based on information available about Annex 2, there is no indication that production in that facility 

decreased after 1950. Therefore, the higher doses were applied for all years of operation.

The annual doses in the tables below can be prorated as necessary based on the work period of the 

employee. Doses from the tables should be applied with the parameters from Table 6-3 for deep and 

shallow doses.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 62 of 90

Table 6-3. Parameters for application of unmonitored external doses 

from 1942 through 1953.

Exposure category

Overestimate energy 

distribution

Exposure 

type GSD

Whole body Photons

100% 30 to 250 keV

Lognormal 5

Skin Electrons >15 keV Lognormal 5

Exposure from 

contaminated floor

Photons

100% 30 to 250 keV

Lognormal 4.5

Tables 6-4, 6-5, and 6-6 provide the annual doses that can be applied to workers at Ames Laboratory 

before 1953. 

Table 6-4. Annual external exposures from uranium for workers in Chemistry Building (mrem/yr).

Period Job title Whole bodya

Skin of hands 

or forearmsa

Skin other than 

hand or 

forearma

Exposure from 

contaminated 

floora

August 1942–

December 

1945

Production 

worker/operator

N/Ab N/Ab N/Ab N/Ab

Researcher 930 2,900 1,600 41

Supervisor 2,300 7,200 4,100 100

Clerical, security, 

janitorial

230 720 410 10

General laborer 4,600 14,000 8,100 210

January 1946–

December 

1953c

Production 

worker/operator

N/Ab N/Ab N/Ab N/Ab

Researcher 93 290 160 4.1

Supervisor 230 720 410 10

Clerical, security, 

janitorial

23 72 41 1

General laborer 460 1,400 810 21

a. This value can be prorated based on the fraction of year worked by the AWE employee.

b. In general, there were no operators in this building; consider that they might have worked in Physical Chemistry Annex 1 

or 2 instead, and apply doses from the tables below.

c. After production period ended in 1945, research is assumed to have one-tenth the exposure potential.

Table 6-5. Annual external exposures from uranium for workers in Physical Chemistry Annex 1

(mrem/yr).

Period Job title Whole bodya Skin of hands or forearmsa

Skin other than 

hand or 

forearma

Dose from 

contaminated 

floora

August 1942–

August 1945

Production 

worker/operator

9,300 29,000 16,000 410

Researcher 930 2,900 1,600 41

Supervisor 2,300 7,200 4,100 100

Clerical, security, 

janitorial

230 720 410 10

General laborer 4,600 14,000 8,100 210

a. This value can be prorated based on the fraction of year worked by the AWE employee.

Table 6-6. Annual external exposures from uranium for workers in Physical Chemistry Annex 2

(mrem/yr).

Period Job title Whole bodya

Skin of hands or 

forearmsa

Skin other than 

hand or forearma

January 1944–

December 1953

Production worker/operator 2,500 280,000 25,000

Researcher 250 28,000 2,500
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 63 of 90

Supervisor 220 14,000 1,100

Clerical, security, janitorial 22 0 0

General laborer 450 138,000 12,000

a. This value can be prorated based on the fraction of year worked by the AWE employee.

6.3.1.2 Coworker Doses – 1952 to the 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 coworker 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. Coworkers are 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 2011a).

The general approach to applying coworker 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 had the worker been 

monitored (50th percentile) or if the monitored worker was clearly identified in the dosimetry records 

(95th percentile) (ORAUT 2011a).

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. In general, the 50th-percentile dose 

can be used to bound doses for those workers when professional judgment indicates the worker was 

likely to have been exposed to intermittent low levels of external radiation (ORAUT 2011a). 

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), coworker data should be applied in the years for which 

records are missing (ORAUT 2011a). The 95th-percentile dose should be assigned to those workers 

who could have been regularly exposed.

The coworker data study for Ames Laboratory included all available dosimetry records from 1952 

through 1981 (Martin 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). The results of the analysis are summarized in Table 6-7. The missed dose 

recommended for monitored workers in Table 6-8 was added to the 50th- and 95th-percentile values 

in Table 6-7. 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 2011a). 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 2005b). These external doses cannot be applied for 

the SEC-00038 class of employees (NIOSH 2006), which covers the period from August 13, 1942, 

through December 31, 1954. As stated in Section 1.3, this SEC class has been expanded to include 

all Ames employees, contractors, and subcontractors.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 64 of 90

Table 6-7. Assigned dose for unmonitored workers (mrem/yr) (McCartney 2006).

Year

Beta Gamma Neutron

Raw data files

References

50th 

percentile

95th 

percentile

50th

percentile

95th 

percentile

50th 

percentile

95th

percentile

1952 N/Aa N/Aa 130 268b 163c 163c Ames (1954a)

1953 0 0 520 590b 650c 650c Ames (1954a)

1954 0 0 520 737b 650 650 Ames (1954b, 1955)

1955 0 0 520 520b 650 650 Ames (1956a)

1956 0 0 520 520b 650 650 Ames (1956b)

1957 0 0 520 520b 650 650 Ames (1957)

1958 0a 0a 520 520b 650 650 Ames (1959)

1959 0a 0a 520 520b 650 650 Ames (1960b)

1960 0 0 520 520b 650 650 Ames (1961b)

1961 0a 0a 520 520b 650 650 Ames (1962b)

1962 0 0 520 520d 650 650d Ames (1963)

1963 0 0 240 265e 650 650 Ames (1964)

1964 0 0 240 313e 650 650 Ames (1965)

1965 0 0 240 408e 650 820e Ames (1966)

1966 0 86e 240 565e 806f 988e HPS (1966)

1967 0 246e 240 745e 932f 1,128e HPS (1967)

1968 0 87e 240 769e 876f 1,086e HPS (1968)

1969 0 175e 240 1,182e 1,030f 1,296e HPS (1969)

1970 0 0 240 671e 680f 932e HPS (1970)

1971 0 0 240 809e 722f 950e HPS (1971)

1972 0 0 240 512e 652f 988e HPS (1972)

1973 0 0 240 701e 650 932e Ames (1973)

1974 0 0 240 721e 650 876e HPS (1974)

1975 0 0 240 1,298e 650 870e Ames/HPS (1975)

1976 0 0 240 497e 650 958e Ames/HPS (1976)

1977 0 0 240 752e 650 904e Ames/HPS (1977)

1978 0 0 240 722e 650 652e Ames/HPS (1978)

1979 0 0 240 819e 650 650 Ames/HPS (1979)

1980 0 0 240 1,595e N/Ac N/Ac Ames/HPS (1980)

1981 0 0 240 240 N/Ac N/Ac HPS (1981)

1982 0 0 85 555 N/Ac N/Ac Landauer (1982)

1983 325 510 105 134 N/Ac N/Ac Landauer (1983)

1984 0 0 60 85 N/Ac N/Ac Landauer (1984)

1985 0 0 60 125 N/Ac N/Ac Landauer (1985)

1986 0 0 60 60 N/Ac N/Ac Landauer (1986)

1987 0 0 60 114 N/Ac N/Ac Landauer (1987)

1988 0 310 72 149 N/Ac N/Ac Landauer (1988)

1989 0 0 60 81 N/Ac N/Ac Landauer (1989)

1990 0 9 60 91 N/Ac N/Ac Landauer (1990)

1991 0 51 60 85 N/Ac N/Ac Landauer (1991)

1992 0 30 60 71 N/Ac N/Ac Landauer (1992)

1993 0 22 60 118 N/Ac N/Ac Landauer (1993)

1994 0 20 62 138 N/Ac N/Ac Landauer (1994)

1995 0 70 60 68 N/Ac N/Ac Siemens (1995)

1996 0 0 60 65 N/Ac N/Ac ICN (1996)

1997 0 0 60 60 N/Ac N/Ac Hokel (1998)

1998 0 0 205 245 N/Ac N/Ac Landauer (1998)

1999 0 115 60 160 N/Ac N/Ac Landauer (1999)

2000 0 150 60 110 N/Ac N/Ac Landauer (2000)

2001 0a 0a 60 76 N/Ac N/Ac Simpson (2001); 

Beckel (2001a,b, 

2002)
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 65 of 90

Year

Beta Gamma Neutron

Raw data files

References

50th 

percentile

95th 

percentile

50th

percentile

95th 

percentile

50th 

percentile

95th

percentile

2002 0 49 60 68 N/Ac N/Ac Landauer (2002)

2003 0 150 201 394 N/Ac N/Ac Landauer (2003)

2004 0a 0a 63 118 N/Ac N/Ac Beckel (2004a,b, 

2005)

2005 0 204 60 123 N/Ac N/Ac McGuigan (2006)

a. No beta doses were measured or reported in 1952, 1958, 1959, 1961, 1981, 1982, 2001, or 2004.

b. 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 1952 through 1961. However, from 1955 through 

1961, 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.

c. Neutron dosimeters were not provided in 1952 or 1953 and were not needed after 1979.

d. The dosimeter exchange frequency in 1962 is uncertain; therefore, the missed dose for a biweekly frequency was 

assumed to be favorable to claimants.

e. The 95th-percentile value was assumed to be received all in one month; missed dose was added for the other 

11 months of the year.

f. The 50th-percentile value was assumed to be received all in one month; missed dose was added for the other 

11 months of the year.

Table 6-8. Potential missed dose for monitored workers.

Dosimeter Period

Exchange

frequencya

MDL (mrem) Missed annual mean dose (mrem)

Skin Deep Neutron Skin Deep Neutron

Film badge-β,γ,n 1952–1956 Biweekly 40b 40b 50b 0c 0c 0c

1957–1962 Biweekly 40b 40b 50b

Film badge-β,γ,n 1963–1981 Monthly 40 40 50 240 240 300

TLD-β and γ 1981–present Quarterly 30 30 N/Ad 60 60

a. Exchange frequencies were established from dosimetry reports. 

b. Estimated MDL typical of film dosimeter capabilities (Wilson 1960, 1987; NIOSH 1993; NRC 1989; Wilson et al 1990). 

c. As discussed in Section 6.2.1, doses equal to the MRD were added to the recorded dose when zero readings occurred. 

Therefore, no additional correction is required for monitored workers for the period from 1952 through 1961.

d. 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 6-8 for the dosimetry systems. These missed doses are included in the values in Table 6-7. 

A flow chart is provided in Figure 6-2 to guide dose reconstructors in the application of the above 

tables and recommendations. The Monitored Worker side of the figure should be used if there are 

record files for the entire period of employment. 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 

period of employment. 

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 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 66 of 90

Figure 6-2. Flowchart for monitored and unmonitored workers. Source: authors.

and several MeV (Morgan 1961). The estimated standard error in recorded film badge doses from 

photons of any energy is ±30% (ORAUT 2006d). The estimated uncertainty in doses recorded by 

TLDs is ±20% from 1982 through 1984 and ±10% since 1985 with NVLAP accreditation (Landauer 

1985).

6.3.4 Skin Dose

In years 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 

a. Uncertainties and correction factors are ±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, respectively.

b. Adjustments for missed and zero doses have already been made to the dose of record for years before 1962 (see 

Section 6.2.1).

c. Correction factor is 2 for neutron (NTA) film badges (see Section 6.2.5.3).

d. ICRP (1991) neutron weighting factor adjustment for neutron energies from 0.1 to 2 MeV.

Note: Photon and neutron organ dose conversion factors are not included in this flowchart and must be 

applied separately.

Monitored Worker Unmonitored Worker

For 1942 to 1953, use Section 

6.3.1.1 to estimate external 

doses

For 1953 to present, tabulate 

beta, photon, and neutron 

doses of record for each year

Multiply beta/photon doses by 

uncertainty correction factora

Add adjustments for missed 

and zero doses after 1961b

Multiply neutron dose of record 

by correction factor of 2c

Adjust neutron dose for ICRP 

(1991) weighting factor by 

multiplying by 1.91d

Tabulate annual assigned doses 

from Table 6-7. See text in 

Section 6.3.1.2 for explanation 

and guidance.

For 1942 to 1953, use Section 

6.3.1.1 to estimate external 

doses

Multiply assigned neutron dose 

by correction factor of 2c

Adjust neutron dose for ICRP 

(1991) weighting factor by 

multiplying by 1.91d
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 67 of 90

been calculated as the sum of the whole-body penetrating dose and the nonpenetrating dose 

(Landauer 1985). 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 Interpretation of Dosimetry Data for Assignment of Shallow Dose (ORAUT 2005b).

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. 

A standard practice in operational health physics is to use a factor of 10 between whole-body and 

extremity exposures (PNNL 2004). That is, if the measured contact dose rate from a source is 10 

times (or more) the measured dose rate at the location of the whole-body dosimeter, extremity 

dosimeters should be assigned for the work. In the case of missing extremity dose data (or 

unmonitored extremities), the whole-body dose can be multiplied by 10 and the result assigned as a 

conservative extremity dose. If the cancer site involves the hands, forearms, feet, or legs below the 

knees, the conservative calculated extremity dose can be used by dose reconstructors.

When a more accurate evaluation of extremity dose based on whole-body dosimetry readings is 

needed, the dose reconstructor should refer to the Division of Compensation Analysis and Support, 

DCAS-TIB-0010, Technical Information Bulletin: Best Estimate External Dose Reconstruction for 

Glovebox Workers, (NIOSH 2010c) for information regarding adjustments to whole-body doses at 

locations associated with the extremities.

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 Interactive RadioEpidemiological 

Program (IREP).

6.4 ORGAN DOSE

Once the Hp(10) adjusted doses have been calculated for each year, the values are used to calculate 

organ doses of interest using the external dose reconstruction implementation guidelines (NIOSH 

2007b). Consistent with Project direction, dose reconstructors should assume anterior-posterior (AP; 

front to back) geometry for the irradiation geometry and for conversion to organ dose. For photons 

measured with film badges (before 1982), use the conversion factor from exposure to organ dose. 

Beginning in 1982, when all measurements were made using TLDs, deep dose equivalent should be 

used. For neutron doses, use the conversion factors from deep dose equivalent to organ dose for AP 

irradiation from NIOSH (2007b).
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 68 of 90

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 with information to identify the source and justification for each 

associated item. Conventional references, which are provided in the next section of this document, 

link data, quotations, and other information to documents available for review on the Project’s Site 

Research Database.

[1] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

This statement is based on personal experience with a 52-in. cyclotron at the University of 

Colorado, where a bioassay program, routine contamination surveys, and effluent monitoring 

were conducted. Bioassay results were always indistinguishable from background. 

Particulate contamination frequently had a very short half-life and detectable levels were 

almost always less than 10 CFR Part 835, Appendix D values. Airborne effluents also had 

short half-lives and were indistinguishable from background. This information is consistent 

with the Site Profile for SLAC (ORAUT 2007a), which found that the environmental dose from 

accelerator effluents was less than 1 mrem/yr and negligible.

[2] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

Common practice in accelerator health physics is to conduct a periodic fenceline survey of 

maximum radiation dose rates. The purpose of such surveys is to determine the maximum 

offsite impact of accelerator operations and identify special controls or limitations that might be 

needed for safe operation. To produce maximum fenceline dose rates, the maximum beam 

energy and beam current were applied to a target selected for maximum radiation production. 

These operating conditions were unusual and could be detrimental to the accelerator 

equipment, so the maximum dose rate conditions existed only long enough to obtain the 

fenceline survey results. Normally, the accelerator 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.

Ames (1961a) states that “background reading in this area (synchrotron) prior to beam being 

turned on was 0.5 to 1.0 mr/hr.” 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 fenceline readings were taken some distance from the 

background reading location and should have not been influenced by the elevated background 

near the accelerator. In any event, the elevated background dose rates were not subtracted 

from the fenceline dose rates reported in Table 4-1.

[3] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

Experience at other accelerator facilities indicates 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.

[4] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

The radiation conditions during the fenceline radiation survey on May 16, 1961, produced dose 

rates at occupiable locations inside the Synchrotron Building that were greater than 1 mR/hr. 

Full-time 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 that 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. 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 69 of 90

[5] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

The buildings of Ames Laboratory are on the main campus of Iowa State University and are 

not physically separated in any way from other university buildings (i.e., no fence or other 

boundaries). Staff, students, and the public are free to walk around the exterior walls of the 

buildings. 

[6] Shipler, Dillard B. ORAU Team. Principal Health Physicist. February 2007. 

More than 1,000 tons of uranium metal were processed at Ames Laboratory in about 3 years. 

The University’s chemistry laboratories were used for R&D on small quantities, rather than 

production quantities. The estimate that the source term in the laboratories was 1/100th of the 

amount handled in production facilities was made and was shown to be comparable to 

releases from laboratories at the University of California at Berkeley, thereby supporting the 

estimate as reasonable.

[7] Shipler, Dillard B. ORAU Team. Principal Health Physicist. February 2007. 

The source of the intake numbers was explained in the previous attribution. The assumption 

was made that the thorium was in equilibrium with its progeny, which is favorable to claimants. 

The intake-to-organ-dose conversion factors require that 232Th, 228Ra, and 228Th be entered as 

separate intakes.

[8] Bihl, Donald E. ORAU Team. Principal Health Physicist. February 2007. 

Because the chemical forms and, therefore, the absorption types of the uranium and thorium 

were not known, especially for the R&D activities, all the default values in ICRP Publication 68 

(ICRP 1995) were considered possible.

[9] Bihl, Donald E. ORAU Team. Principal Health Physicist. February 2007. 

The first sustained chain reaction occurred in the Chicago Pile on December 2, 1942. The pile 

was dismantled in the spring of 1943, so the first irradiated fuel would not have reached Ames 

Laboratory until after that, probably not until 1944, but assuming 1943 was considered 

favorable to claimants.

[10] Bihl, Donald E. ORAU Team. Principal Health Physicist, February 2007. 

The hot cell was engineered to prevent releases into the workplace and would have had better 

filtration in the exhaust system than a general chemistry lab built in the 1940s. The exact 

nature of the hot laboratory used in the 1940s was not discovered in the records. Presumably, 

it would have involved some methods for recognizing and controlling the spread of 

contamination with some filtration on the exhausts, but it was unlikely that the controls of the 

hot laboratory were as good as those for the hot cell.

[11] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

The decommissioning date of 1990 was determined in a personal communications with (Name 

redacted), Radiation Safety Officer at Ames Laboratory, July 2006.

[12] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2007.

Review of references and discussions in Sections 2.1 and 2.2 indicated the location of 

operations with uranium.

[13] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2007.

Review of references and discussions in Sections 2.1 and 2.2 indicated the location of 

operations with uranium.
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 70 of 90

[14] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2007.

Because there was no information on the amount of time in a day that a worker would be 

exposed to material, an assumption had to be made. The assumptions in the section on 

uranium are the same as the assumptions used in Battelle (2011).

[15] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2007.

Because there was no information on the amount of time in a day that a worker would be 

exposed to material, an assumption had to be made. The assumptions in the section on 

uranium are the same as the assumptions used in Battelle (2011).

[16] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2007.

Because there was no information on the amount of time in a day that a worker would be 

exposed to material, an assumption had to be made. The assumptions in the section on 

uranium are the same as the assumptions used in Battelle (2011).

[17] Bihl, Donald E. ORAU Team. Principal Health Physicist. April 2007.

The calculations in the preceding three paragraphs were conducted by Mr. Bihl using IMBA 

from data in the referenced documents. Although hard to read, the urinalysis values used for 

the most exposed worker from Tybout (1944) were 0.00, 0.06, 0.00, 0.20, 0.00, and 0.10 mg/L 

adjusted to mg/d. The 0.00 values were entered as <0.056 mg/d, assuming the detection level 

was approximately 0.04 mg/L.

[18] Bihl, Donald E. ORAU Team. Principal Health Physicist. February 2007.

As long as the Energy Employee’s actual excretion is less than the excretion plotted in the 

figures, the intake is also less than the default intakes. For a noncompensable case, if the true 

intake is less than the default intake, the default intake can be used for efficiency. 

[19] Bihl, Donald E., and Hickey, Eva Eckert. ORAU Team. Principal Health Physicists. February 

2007.

The air sample results were indistinguishable from background, so were not much help. 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 is 

usually reported in units of activity per 100 cm2

, but the authors could not be sure of that. We 

had to "assume" the survey values were for 100 cm2

. Because of that uncertainty, we 

arbitrarily raised the 1,224-dpm value to 2,000 dpm/100 cm2 to represent an upper bound

clearly. 

[20] Bihl, Donald E. ORAU Team. Principal Health Physicist. February 2007.

Indoors resuspension factors range from 10-6 to 10-3

. The International Atomic Energy Agency

(Clayton 1970) recommends an indoor resuspension factor of 5 × 10-5

. 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. 

[21] Bihl, Donald E. ORAU Team. Principal Health Physicist. February 2007.

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 need to be entered as intakes separate from 232Th. 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 71 of 90

[22] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2007.

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.

[23] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2007.

In Hokel et al. (1998), there were statements that the areas that were rarely accessed 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.

[24] Bihl, Donald E. ORAU Team. Principal Health Physicist. February 2007.

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 (SRS), and LANL, to 0.02 μCi/L at ORNL and 

Lawrence Livermore National Laboratory, to 0.0002 μCi/L at INL. The Hanford and SRS

values were recording levels with actual laboratory MDAs generally being lower. Review of 

the way data were recorded at Ames Laboratory indicated that 0.1 μCi/L was the likely MDA or 

recording level. 

[25] Bihl, Donald E. ORAU Team. Principal Health Physicist. February 2007. 

The table was created by Mr. Bihl using the lognormal tritium dose plots created by Thomas R. 

La Bone, as explained in the text. 

[26] Bihl, Donald E. ORAU Team. Principal Health Physicist. February 2007. 

Strontium is type F with the exception of the titanate form, which is rare. There is no reason to 

suspect strontium titanate was in use at Ames Laboratory. 

[27] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2007.

Table 5-8 is a summary of the intakes described in this section. The data are taken from 

(1) Tables 5-1, 5-2, and 5-3 and the discussions in Sections 5.1.1.1 on uranium inhalation and 

Section 5.1.1.2 on uranium ingestion; (2) Table 5-4 and the discussion in Section 5.1.1.3 on 

resuspension of uranium during periods of no operation; (3) the discussion in Section 5.2 on 

thorium exposure from leftover thorium contamination; (4) doses from tritium as discussed in 

Section 5.3; and (5) Table 5-7 doses from fission product intakes as discussed in Section 5.4.

[28] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

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.

[29] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

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 used, records of the 

dosimeter readings were not kept.

[30] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

During the 1965-to-1981 period, 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 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 72 of 90

correlate some badge numbers and individual names. However, there were many badge 

numbers for which a name could not be assigned.

[31] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

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.

[32] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

After 1979, neutrons were no longer included in the source term at Ames Laboratory and 

neutron dosimeters were no longer needed.

[33] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

As stated in Section 6.2.1, recordkeeping from January 1962 through 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 1963). 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 through March 1965. No 

records were found for the first half of 1963. 

[34] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

With the wide variety of radiation sources used at Ames Laboratory, it is possible that beta, 

photon, and neutron fields with energies outside the assumed ranges existed. However, the 

predominant radiation fields were within the assumed energy ranges and, if it is assumed that 

100% of the radiation fields were within these ranges, it is a conservative assumption that is 

generally favorable to claimants.

[35] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

The early history of research at Ames Laboratory was described by Fulmer (1947); later 

history was described in Ames (1985). These documents discussed numerous research 

projects that involved many beta-emitting radioisotopes. The quantities used in research were 

much smaller than the quantities involved in production operations.

[36] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

The descriptions of research and production operations at Ames Laboratory (Fulmer 1947;

Ames 1985) established the presence of photons in the less-than-30-keV, 30-to-250-keV, and 

greater-than-250-keV energy ranges. However, the predominant photon energy range was 

from 30 to 250 keV. If it is assumed that 100% of the photon radiation fields were within this 

range, it is a conservative assumption that is generally favorable to claimants.

[37] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

Descriptions of the synchrotron (Ames 1967) and the research reactor (Voigt 1981) 

established that neutrons of varying energies were present in these facilities. However, no 

records were found that characterized the neutron spectrum in either facility. If it is assumed 

that 100% of the neutron radiation fields were with the 0.1-to-2-MeV energy range, it is a 

conservative assumption that is generally favorable to claimants.

[38] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

Commercial film badge service was supplied by BNL from April 1954 to June 1957, by 

Nucleonic Corporation of America from July 1957 to June 1963, by Atomic Film Badge 
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Corporation from July 1963 to March 1965, and by Health Physics Services from June 1965 to

September 1981 (see Table 6-1).

[39] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

The film badge dosimeters used at Ames Laboratory were state-of-the-art commercial 

products that were properly designed and calibrated to give reliable dose results. The TLDs 

used since 1980 were commercial products that were NVLAP-accredited and calibrated to 

give reliable dose results. 

[40] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

The neutron doses received by ALRR personnel were somewhat higher when compared with

similar personnel at the High Flux Beam Reactor at BNL (ORAUT 2010a). The ALRR records 

did not include any gamma or neutron radiation surveys around the experimental areas, nor 

was there any evidence of neutron spectral measurements. The lack of survey data suggests 

that management of neutron shielding might have been less than adequate.

[41] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

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, 

but the spectrum at SLAC included higher energy neutrons, thus the neutron radiation fields 

were not identical. 

[42] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

The SLAC Site Profile (ORAUT 2007b) 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. 

[43] Martin, Jerome B. ORAU Team. Senior Health Physicist. February 2007. 

During the 1965-to-1981 period, 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.

[44] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2007.

Because there was no information on the amount of time in a day that a worker would be 

exposed to material, an assumption had to be made. The assumptions in the section on 

uranium are the same as the assumptions used in Battelle (2011).
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 74 of 90

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GLOSSARY

absorbed dose

Amount of energy (ergs or joules) deposited in a substance by ionizing radiation per unit mass 

(grams or kilograms) of the substance and measured in units of rads or grays See dose.

absorption type 

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood. 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 absorbtion into 

blood from the respiratory tract (moderate rate of solubilization), and S: deposited materials 

that are relatively in the respiratory tract (slow solubilization). Also called solubility type. 

acute exposure

Radiation exposure to the body delivered in a short period. See chronic exposure.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2. 

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

becquerel (Bq) 

International System unit of radioactivity equal to 1 disintegration per second; 1 curie equals 

37 billion (3.7 × 1010) becquerels. 

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

Determination of kinds, quantities, or concentrations, and in some cases locations of 

radioactive material in a living body, whether by direct measurement (in vivo measurement) or 

by analysis and evaluation of materials excreted or removed from the body (in vitro

measurement). 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. 

chronic exposure 

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years). 

See acute exposure.
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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 x 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

deep dose equivalent (DDE, Hd, Hp(10)) 

Dose equivalent in units of rem or sievert for a 1-centimeter depth in tissue (1,000 milligrams 

per square centimeter). See dose.

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 sievert; other types of dose are in 

units of roentgens, rads, reps, or grays.

dose equivalent (H, DE) 

Product of absorbed dose in units of rem or sievert 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, neutron film dosimeter, and thermoluminescent 

dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.

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%. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade uranium 

contains greater than 90% 235U.

exposure

In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure.

film 

In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter.

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 
Document No. ORAUT-TKBS-0055 Revision No. 03 Effective Date: 01/03/2012 Page 86 of 90

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

fission products

(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) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) 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 of high energy; the difference is that X-rays do not originate in the nucleus. 

gray (Gy) 

International System unit of absorbed radiation dose, which is the amount of energy from any 

type of ionizing radiation deposited in any medium; 1 Gy equals 1 joule per kilogram or 

100 rads.

high-efficiency particulate air (HEPA) filter 

Dense filter that removes contaminants from air flows before return to the working environment 

or discharge to the outside air (exhaust).

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. 

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, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons.

in vitro bioassay

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo bioassay

The measurements of radioactive material in the human body utilizing instrumentation that 

detects radiation emitted from the radioactive material in the body. 

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds. 

internal dose or exposure

Dose received from radioactive material in the body.

internal dose assessment

Estimation of an intake of radioactive material and the consequent internal radiation dose 

based on measurements in the work environment and/or bioassay. 
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minimum detectable activity or 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)

Lowest amount (mass or activity) of a substance detectable by a specific instrument or 

process. Often assumed to be the level at which a dose is detected at the 2-sigma level (i.e., 

95% of the time). 

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, ground water, 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 (U, U-nat, NU) 

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

weight. The specific activity of this mixture is 2.6 × 107 becquerels per kilogram 

(0.7 picocuries per gram). See uranium.

neutron

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.

neutron film dosimeter

Film dosimeter with a nuclear track emulsion, type A, film packet.

nuclear track emulsion, Type A (NTA)

Film sensitive to fast neutrons made by Eastman Kodak. The developed image has tracks 

caused by neutrons that visible under oil immersion with about 1,000-power magnification.

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 exposure

Exposure to radiation and/or to radioactive material from sources of radiation in a restricted 

area or in the course of employment in which the individual's assigned duties. Occupational 

exposure does not include exposure to background radiation, as a patient from medical 
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practices, from voluntary participation in medical research programs, or as a member of the 

public.

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 

included in dose reconstruction under EEOICPA.

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. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as 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. 

pocket ionization chamber (PIC) 

Cylindrical monitoring device commonly clipped to the outer clothing of an individual to 

measure ionizing radiation. A PIC may be self-reading or require the use of a outside device 

to be able to read the dosimeter. Also called pencil, pocket pencil, pencil dosimeter, and 

pocket dosimeter.

probability of causation (POC)

For dose reconstruction under the Energy Employees Occupational Illness Compensation 

Program Act, the percent likelihood that a worker incurred a particular cancer from 

occupational exposure to radiation.

progeny 

Nuclides that result from decay of other nuclides. Also called decay products and formerly 

called daughter products.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joule per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose. 

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer.
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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.

radiograph

Static images produced on radiographic film by gamma rays or X-rays after passing through 

matter. In the context of EEOICPA, radiographs are X-ray images of the various parts of the 

body used to screen for disease.

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 average American receives 360 millirem a year from background radiation. 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°C and standard atmospheric pressure. An 

exposure of 1 R 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. See special monitoring.

shallow absorbed dose (Ds)

Absorbed dose at a depth of 0.07 centimeters (7 milligrams per square centimeter) in a 

material of specified geometry and composition.

shallow dose equivalent [SDE, Hs, Hp(0.07)] 

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

special monitoring

Monitoring in addition to the routine monitoring program carried out for special purposes such 

as estimating the amount of radionuclide deposited in a person after a known or suspected 

accidental intake or after a known or suspected environmental release. See routine 

monitoring.

synchrotron 

Roughly circular particle accelerator in which the particles travel in synchronized bunches at 

fixed radius.

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated by radiation, release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose. 

unmonitored dose

Potential unrecorded dose that could have resulted because a worker was not monitored. See 

missed dose.
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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 (WB) 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 

(1) See X-ray radiation. (2) See radiograph.

X-ray radiation

Penetrating electromagnetic radiation (photons) of short wavelength (0.001 to 10 nanometers) 

and energy less than 250 kiloelectron-volts. X-rays usually come from excitation of the 

electron field around certain nuclei. Once formed, there is no difference between X-rays and 

gamma rays, but gamma photons originate inside the nucleus of an atom.