Page 1 of 81

DOE Review Release 08/25/2023

Y-12 Plant – Occupational Internal Dose

ORAUT-TKBS-0014-5 Rev. 04

Effective Date: 08/03/2023

Supersedes: Revision 03

Subject Expert(s): Darin R. Hekkala and Dale D. Thomas III

Document Owner 

Approval: Signature on File

Stephanie L. Felch, Document Owner

Approval Date: 08/02/2023

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 07/31/2023

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 07/31/2023

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 07/31/2023

Approval: Signature on File

Timothy D. Taulbee, Associate Director for Science

Approval Date: 08/03/2023

FOR DOCUMENTS MARKED AS A TOTAL REWRITE OR REVISION, 

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

New Total Rewrite Revision
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 2 of 81

 

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

03/17/2004 00 New document to establish TBD for the Y-12 National Security 

Complex – Occupational Internal Dose. First approved issue.

Initiated by William E. Murray.

05/10/2005 01 First revision to include table numbers and titles in the body of the 

document. Approved issue of Revision 01. Initiated by William E. 

Murray.

10/11/2005 01 PC-1 Page change initiated to incorporate the definition of U.S.C. on page 

6 and details for the definition of a DOE facility on page 7. No 

sections were deleted. First approved page change revision for 

Revision 01. Initiated by William E. Murray.

01/12/2006 01 PC-2 No changes occurred as a result of formal internal review. Page 

change initiated to make minor changes to text in Section 5.1, page 

7. Deletes and adds text about the default value for the Th-228/Th232 equilibrium ratio for lung counts in Section 5.3, page 31. Adds 

new Figure 5.4. - Variation of activity in natural thorium after a single 

purification on page 31. Also, changes date in C. M. West reference 

on page 36. No sections were deleted. Training required: As 

determined by the Task Manager. Initiated by William E. Murray.

02/14/2006 01 PC-3 Approved page change revision. Page changes initiated to 

incorporate definitions and directions for dose reconstruction for nonpresumptive cancers that are excluded from the 1943 through 1947 

Special Exposure Cohort. Affected pages are: page 7, page 9 in 

Section 5.1, pages 20 and 21 in Section 5.2.1.1, and page 36 in the 

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

determined by the Task Manager. Initiated by William E. Murray.

08/03/2006 02 Revised to incorporate additional directions for dose reconstruction 

for non-presumptive cancers that are excluded from the 1943 

through 1947 Special Exposure Cohort. Approved issue of Revision 

02. Removed Section 5.4 from the text and Table of Contents, as the 

most recent page change addresses 1943 to 1947, and coworker 

data are used for 1948 to 1950, as specified in the coworker TIB.

Incorporates NIOSH formal review comments. This revision results in 

no change to the assigned dose and no PER is required. Training 

required: As determined by the Task Manager. Initiated by William E. 

Murray.

10/10/2006 02 PC-1 Page change revision initiated to incorporate Y-12 National Security 

Complex Special Exposure Cohort class descriptions in newly added 

Section 5.1.3 on page 10. Minor changes occurred on pages 6, 8, 

and 9. Added one reference on page 37. Incorporates one new 

acronym, editorial changes on pages 8, 9, and 10, modifications to 

SEC instructions in Section 5.1.3.1 on page 10. Two references were 

deleted from the Reference Section. Incorporates internal and 

NIOSH formal review comments. No sections were deleted. This 

revision results in no change to the assigned dose and no PER is 

required. Training required: As determined by the Task Manager.

Initiated by William E. Murray.
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EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

03/12/2012 03 Revision to add new SEC designation to Section 5.1.3 to clarify, 

correct, and update the SEC class of covered employees. Added 

U-232 exception to Section 5.2.2. Added internal coworker study 

(formerly ORAUT-OTIB-0029) as Attachment B. Revision initiated to 

add Attributions and Annotations section and make minor editorial 

corrections. Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by 

Steven R. Reed.

08/03/2023 04 Revision to incorporate SEC-00250, address outstanding priority 

ABRWH issues, and bring into ORAUT-PROC-0031 compliance as 

specified during DCAS/ORAUT discussions. Changed title of 

document to use Y-12 Plant rather than Y-12 National Security 

Complex. Incorporates formal internal and NIOSH review comments. 

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

determined by Objective Manager. Initiated by John M. Byrne and 

authored by Stephanie L. Felch.
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 4 of 81

 

TABLE OF CONTENTS

SECTION TITLE PAGE

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

5.1 Introduction .............................................................................................................................. 9

5.1.1 Purpose...................................................................................................................... 10

5.1.2 Scope......................................................................................................................... 10

5.1.3 Special Exposure Cohort............................................................................................ 10

5.2 Uranium Solubility in the Lung................................................................................................ 13

5.2.1 Particle Size Information............................................................................................. 14

5.2.2 Uranium Alpha Activity as a Function of Enrichment................................................... 14

5.2.3 Temporal Pattern of Uranium Exposures.................................................................... 15

5.2.4 Other Radionuclides of Concern Including Recycled Uranium Contaminants ............. 16

5.2.4.1 Unmonitored Radionuclides from Recycled Uranium ................................... 16

5.3 In Vitro Urine Analysis............................................................................................................ 23

5.3.1 Coverage.................................................................................................................... 23

5.3.2 Sample Collection....................................................................................................... 23

5.3.3 Minimum Detectable Activities.................................................................................... 24

5.3.4 In Vitro Bioassay Methods for Individual Radionuclides.............................................. 27

5.3.4.1 Uranium ....................................................................................................... 27

5.3.4.2 Plutonium..................................................................................................... 29

5.3.4.3 Tritium.......................................................................................................... 30

5.3.4.4 Other Radionuclides..................................................................................... 30

5.3.5 Fecal Sample Analysis ............................................................................................... 30

5.4 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting Practices.............. 30

5.4.1 Whole Body Counting................................................................................................. 30

5.4.2 Chest Counting........................................................................................................... 31

5.4.2.1 Uranium ....................................................................................................... 31

5.4.2.2 Other Actinides ............................................................................................ 32

5.4.2.3 Other Radionuclides..................................................................................... 34

5.5 Internal Dose Reconstruction Guidance................................................................................. 34

References ........................................................................................................................................ 37

Glossary ............................................................................................................................................ 44

ATTACHMENT A OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS............ 49

ATTACHMENT B INTERNAL DOSIMETRY CO-EXPOSURE DATA........................................... 56
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LIST OF TABLES

TABLE TITLE PAGE

5-1 Enrichment assumptions for dose assessment for analytical techniques................................ 15

5-2 Y-12 uranium receipts summary presenting average levels of three predominant RU 

contaminants through plant history, 1953 to 1999 .................................................................. 17

5-3 Total Y-12 RU receipts summary ........................................................................................... 17

5-4 Primary contaminants (at 3-year decay) in VHEU fuels and of concern in internal dose 

reconstruction ........................................................................................................................19

5-5 Comparison of the production of three selected isotopes in VHEU spent fuel with levels 

reported in HEU shipments to Y-12........................................................................................ 19

5-6 Ranges of reported contaminant values in the main processing facilities ............................... 20

5-7 Typical uranium enrichment materials at Y-12 with calculated and inferred RU 

contaminant levels .................................................................................................................21

5-8 Recommended defaults based on upper levels of expected ranges....................................... 22

5-9 LD values................................................................................................................................26

5-10 233U to 232U activity ratios........................................................................................................ 29

5-11 Reported lung counting detection limits for 235U ..................................................................... 31

5-12 Reported lung counting detection limits for 238U ..................................................................... 32

5-13 Calculated lung counting detection limits for 212Pb and 228Ac, 1979 to 1991 ........................... 33

A-1 Uranium urinalysis detection limits ......................................................................................... 53

A-2 Other in vitro detection limits .................................................................................................. 54

A-3 Reported in vivo detection limits for other radionuclides......................................................... 54

B-1 Monthly uranium 24-hour urinary excretion rate analyses, 1952 to 1988................................ 59

B-2 Type F uranium intake periods and rates ............................................................................... 65

B-3 Type M uranium intake periods and rates .............................................................................. 65

B-4 Type S uranium intake periods and rates ............................................................................... 66

B-5 Type F uranium intake periods and rates ............................................................................... 66

B-6 Type M uranium intake periods and rates .............................................................................. 67

B-7 Type S uranium intake periods and rates ............................................................................... 67

LIST OF FIGURES

FIGURE TITLE PAGE

5-1 212Pb MDAs at various CWTs for a 30-minute count............................................................... 34

B-1 Assumed uranium intake, 01/01/1947 to 12/31/1969, 50th-percentile results, type F ............. 68

B-2 Assumed uranium intake, 01/01/1970 to 12/31/1988, 50th-percentile results, type F ............. 69

B-3 Assumed uranium intake, 01/01/1947 to 02/28/1978, 50th-percentile results, type M ............ 70

B-4 Assumed uranium intake, 01/01/1947 to 04/30/1952, 50th-percentile results, type M ............ 70

B-5 Assumed uranium intake, 08/01/1953 to 11/30/1953, 50th-percentile results, type M ............ 71

B-6 Assumed uranium intake, 11/01/1956 to 02/28/1959, 50th-percentile results, type M ............ 71

B-7 Assumed uranium intake, 06/01/1961 to 12/31/1962, 50th-percentile results, type M ............ 72

B-8 Assumed uranium intake, 10/01/1968 to 12/31/1972, 50th-percentile results, type M ............ 72

B-9 Assumed uranium intake, 03/01/1978 to 09/30/1984, 50th-percentile results, type M ............ 73

B-10 Assumed uranium intake, 10/01/1984 to 12/31/1988, 50th-percentile results, type M ............ 73

B-11 Predicted uranium urinary excretion from eight independent intakes, 50th percentile, 

1952 to 1988..........................................................................................................................74

B-12 Predicted uranium urinary excretion from eight independent intakes, 84th percentile, 

1952 to 1988..........................................................................................................................75

B-13 Assumed uranium intake, 01/01/1947 to 02/28/1978, 50th-percentile results, type S............. 76
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B-14 Assumed uranium intake, 01/01/1947 to 04/30/1952, 50th-percentile results, type S............. 76

B-15 Assumed uranium intake, 08/01/1953 to 11/30/1953, 50th-percentile results, type S............. 77

B-16 Assumed uranium intake, 11/01/1956 to 02/28/1959, 50th-percentile results, type S............. 77

B-17 Assumed uranium intake, 06/01/1961 to 12/31/1962, 50th-percentile results, type S............. 78

B-18 Assumed uranium intake, 10/01/1968 to 12/31/1972, 50th-percentile results, type S............. 78

B-19 Assumed uranium intake, 03/01/1978 to 09/30/1984, 50th-percentile results, type S............. 79

B-20 Assumed uranium intake, 10/01/1984 to 12/31/1988, 50th-percentile results, type S............. 79

B-21 Predicted uranium urinary excretion from eight independent intakes, 50th percentile, 

1952 to 1988..........................................................................................................................80

B-22 Predicted uranium urinary excretion from eight independent intakes, 84th percentile, 

1952 to 1988..........................................................................................................................81
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 7 of 81

 

ACRONYMS AND ABBREVIATIONS 

ABRWH Advisory Board on Radiation and Worker Health

AMAD activity median aerodynamic diameter

AWE Atomic Weapons Employer

CEDE committed effective dose equivalent

CER Center for Epidemiologic Research

Ci curie

cpm counts per minute

CWT chest wall thickness

d day

DCF dose conversion factor

DF decontamination factor

DL decision level

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

F fast (absorption type)

FMPC Feed Materials Production Center

g gram

GSD geometric standard deviation

hr hour

HEU highly enriched uranium

HTO tritiated water

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

kg kilogram

L liter

LC critical level

LD detection level

LEU low-enriched uranium

M moderate (absorption type)

MDA minimum detectable activity

mg milligram

mL milliliter

mm millimeter

mrem millirem
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NaI(Tl) thallium-drifted sodium iodide

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NU natural uranium

ORAU Oak Ridge Associated Universities

ORAUT ORAU Team

pCi picocurie

PER program evaluation report 

pH negative logarithm of the hydrogen ion concentration, a measure of alkalinity or acidity

ppb parts per billion

ppm parts per million

R&D research and development

RU recycled uranium

RWP radiation work permit

S slow (absorption type)

SEC Special Exposure Cohort

SMC Specific Manufacturing Capability

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

TBD technical basis document

TIB technical information bulletin

UCC Union Carbide Corporation

U.S.C. United States Code

VHEU very highly enriched uranium

wt% weight percent

yr year

§ section or sections

μCi microcurie

μg microgram

μm micrometer
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5.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

time for inclusion in dose reconstruction.

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

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

dose reconstructions for covered employees [NIOSH 2010]:

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

conventional structures, and

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

reasons.
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5.1.1 Purpose

This TBD describes internal dosimetry systems and practices at the Y-12 Plant. This document 

discusses historical and current practices in relation to the evaluation of internal exposure data for 

monitored and unmonitored workers. For convenience, this TBD uses the Y-12 Plant or Y-12 to refer 

to this site.

5.1.2 Scope

The Y-12 Plant, operated by Consolidated Nuclear Security, LLC and its predecessors, has played an 

important role in the development of the U.S. nuclear weapons program. Operations at the plant have 

evolved from the separation and enrichment of uranium to the manufacture and assembly of nuclear 

weapons components to stockpile stewardship and maintenance. This TBD is part of the Y-12 Plant 

Site Profile, which describes plant facilities and processes, historical information, and environmental 

data in relation to dose reconstruction for Y-12 workers. It contains supporting documentation to assist 

in the reconstruction of occupational internal doses resulting from these activities.

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

the beginning of operations at the Y-12 Plant. One objective of this document is to provide supporting 

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

radiation exposures covered by the EEOICPA legislation. In addition, this document presents the 

technical basis of methods used to prepare Y-12 worker dose information for input into the NIOSH 

Interactive RadioEpidemiological Program (IREP).

At the Y-12 Plant, uranium isotopes in various chemical and physical forms have been the primary 

contributors to internal radiation doses to workers since November 1943. Y-12 missions have involved 

other radionuclides and, at times, some uranium compounds could have contained impurities with 

radiological health implications. These implications and default analyses are discussed in this TBD.

However, the primary focus on internal dose control at Y-12 has been on uranium compounds and 

alloys over a wide range of 235U enrichment. Therefore, this section begins with an overview of 

concepts that apply broadly to the history of uranium work at Y-12. Attachment A summarizes some of 

the key points from the following sections. Attachment B provides co-exposure intakes for applications 

when bioassay records are unavailable or incomplete.

There were programs in the early years of Y-12 operations for which currently available information is 

insufficient to provide general guidance for internal dose reconstruction. At the peak of 235U 

enrichment operations from 1943 to 1945, there were 1,156 “large mass spectrographs,” called 

calutrons, in operation. This enrichment production technique was shut down in 1946.

The 86-Inch Cyclotron began operating on November 11, 1950, and operated until the early 1980s.

The 86-Inch Cyclotron was later used to create neutron-deficient radionuclides as a part of the 

research and development (R&D) program noted above. Some of these R&D efforts involved 

plutonium, and workers could have plutonium bioassay results in their records [Patterson et al. 1957]. 

When claim information indicates that a Y-12 worker was involved with research activities involving 

the calutron, cyclotron (accelerator), fusion work, or plutonium (except in the case of recycled uranium 

[RU] exposure, which is addressed in this section), consideration must be given to possible exposure 

to radionuclides other than uranium [Wilcox 1999]. 

5.1.3 Special Exposure Cohort

The Secretary of the U.S. Department of Health and Human Services has issued six designations of 

classes of Y-12 employees as additions to the SEC.
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March 1, 1943, to December 31, 1947

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

who worked in uranium enrichment operations or other radiological activities at the Y-12 

facility in Oak Ridge, Tennessee from March 1943 through December 1947 and who 

were employed for a number of work days aggregating at least 250 work days, either 

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

(excluding aggregate work day requirements) established for other classes of employees 

included in the SEC [Leavitt 2005, p. 3]. 

NIOSH determined that it lacked access to sufficient information to estimate either the maximum 

radiation dose incurred by any member of the class or to estimate such radiation doses more 

precisely than a maximum dose estimate [NIOSH 2005].

A second designation extended the class to include all workers in all areas of Y-12 during this period:

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

contractors or subcontractors who worked at the Y-12 Plant in Oak Ridge, Tennessee 

from March 1, 1943 through December 31, 1947 for a number of work days aggregating 

at least 250 work days occurring either solely under this employment or in combination 

with work days within the parameters established for one or more other classes of 

employees in the Special Exposure Cohort [Leavitt 2008, p. 3].

NIOSH determined that it lacked the internal dosimetry data necessary to reconstruct the internal 

exposures from uranium enrichment and other radiological activities. In addition, NIOSH determined 

that, because exposure potential might not have been limited to only specific buildings or groups of 

workers at Y-12, the SEC class definition should include all employees in all areas of Y-12, and all 

employees of DOE, its predecessor agencies, and their contractors and subcontractors who worked 

at Y-12 during the period from March 1, 1943, to December 31, 1947 [NIOSH 2008]. NIOSH 

considers it feasible to adequately reconstruct the occupational medical dose for Y-12 workers for the 

covered period. Although no exposure data are available for the medical X-rays performed during this 

time, surrogate data were used to develop an exposure matrix for the Y-12 site profile [Oak Ridge 

Associated Universities (ORAU) Team (ORAUT) 2007a]. 

January 1, 1948, to December 31, 1957

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

who were monitored or should have been monitored for:

(1) thorium exposures while working in Building 9201-3, 9202, 9204-1, 9204-3, 9206, 

or 9212 at Y-12 for a number of work days aggregating at least 250 work days 

from January 1948 through December 1957 or in combination with work days 

within the parameters (excluding aggregate work day requirements) established 

for one or more classes of employees in the SEC; or

(2) radionuclide exposures associated with cyclotron operations in Building 9201-2 

at Y-12 for a number of work days aggregating at least 250 work days from 

January 1948 through December 1957 or in combination with work days within 

the parameters (excluding aggregate work day requirements) established for one 

or more classes of employees in the SEC [Leavitt 2006, p. 3].

NIOSH determined that the specific biological monitoring data, air monitoring information, process and 

radiological source information, and surrogate data from similar operations at other sites that would 

allow it to estimate, with sufficient accuracy, the potential internal radiological exposures to thorium 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 12 of 81

 

and cyclotron radionuclides are not sufficient to complete dose reconstructions for the class [NIOSH 

2006a,b].

A second designation extended the class to include all workers in all areas of Y-12 during this period:

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

contractors and subcontractors who worked at the Y-12 facility in Oak Ridge, Tennessee, 

during the period from January 1, 1948 through December 31, 1957, for a number of 

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

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

or more other classes of employees in the Special Exposure Cohort [Sebelius 2011, 

p. 3].

NIOSH determined that, due to undocumented worker movements across the site, limited 

worker-specific information about work locations, and a determination by DOL that employment 

records do not indicate work locations, it is unable to eliminate any specific worker from potential 

exposure scenarios based on assigned work location. NIOSH found that a determination cannot 

always be made about the specific area an employee worked in or whether an employee should have 

been monitored for radiological exposures. Accordingly, NIOSH determined that it was necessary to 

remove the area-specific and monitoring criteria from the class description associated with thorium 

and cyclotron exposures for the period. NIOSH determined that it was also necessary to expand the 

SEC class definition to include all employees and areas of Y-12 during this period [NIOSH 2011]. In 

its evaluation [NIOSH 2011], NIOSH found that internal exposure could be reconstructed for all 

employees from January 1, 1948, to December 31, 1957, with the exception of internal exposure to 

thorium and cyclotron radionuclides. NIOSH found that external exposure, including occupational 

medical doses, could be reconstructed for all employees between January 1948 and December 31, 

1957.

January 1, 1958, to December 31, 1976

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

contractors and subcontractors who worked at the Y-12 Plant in Oak Ridge, Tennessee, 

during the period January 1, 1958, through December 31, 1976, for a number of work 

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

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

classes of employees in the Special Exposure Cohort [Azar 2019a, p. 3].

NIOSH determined that it lacked sufficient information to allow it to estimate with sufficient accuracy 

the internal exposures to thorium (and associated progeny) for the period from January 1, 1958, to 

December 31, 1976, and 241Pu for the period from January 1, 1958, to December 31, 1966 [NIOSH 

2018].

Although NIOSH has determined that there are in vivo monitoring data for thorium (i.e., lung counts) 

during the period from January 1, 1958, to December 31, 1976, these data are recorded in total 

thorium mass. NIOSH is unable to use these data to determine the associated quantities of 232Th, 228Th, and 228Ra. For this reason, NIOSH cannot determine with sufficient accuracy the internal 

exposures that might be represented by each thorium lung measurement. NIOSH has also evaluated 

the available gross alpha air monitoring data at the Y-12 Plant and determined that they cannot be 

used to accurately reconstruct internal exposure from thorium and associated progeny. NIOSH has 

not identified biological monitoring data specific to 241Pu that can be used to reconstruct 241Pu

exposure during the period from January 1, 1958, to December 31, 1966. NIOSH has identified 

sufficient monitoring data to reconstruct 241Pu beginning in 1967 [NIOSH 2018].
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January 1, 1977, to July 31, 1979

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

contractors and subcontractors who worked at the Y-12 Plant in Oak Ridge, Tennessee, 

during the period between January 1, 1977, and July 31, 1979, for a number of work 

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

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

classes of employees in the Special Exposure Cohort [Azar 2019b, p. 3].

NIOSH determined that it lacked sufficient information to allow it to estimate with sufficient accuracy 

internal exposures to thorium (and associated progeny) for the period from January 1, 1977, to 

July 31, 1979, due to a lack of sufficient thorium-related lung count data for this period [NIOSH 2019].

Although NIOSH has determined that there are in vivo monitoring data for thorium (i.e., lung counts) 

during the period from January 1, 1977, to July 31, 1979, these data are recorded in total thorium 

mass. NIOSH is unable to use these data to determine the associated quantities of 232Th, 228Th, and 

228Ra. For this reason, NIOSH cannot determine with sufficient accuracy the internal exposures that 

might be represented by each thorium lung measurement. Air-sampling data specific to thorium for 

the period under evaluation (January 1, 1977, to December 31, 1994) are not available to NIOSH 

[NIOSH 2019, p. 42]. Consistent with its findings in the earlier Y-12 Plant evaluation reports, NIOSH 

finds that there is sufficient monitoring and source term information available to reconstruct external 

dose, including occupational medical dose, with sufficient accuracy for all Y-12 Plant employees 

during the period from January 1, 1977, to July 31, 1979.

SEC Classes Summary

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

classes, NIOSH intends to use any internal and external monitoring data that might become available, 

and can be interpreted using its existing dose reconstruction processes or procedures, for an 

individual claim. Therefore, for individuals employed at Y-12 during the period from March 1, 1943, to 

July 31, 1979, who do not qualify for inclusion in the SEC, dose reconstructions may be performed 

using these data as appropriate. All dose reconstructions for all workers having employment during an 

SEC period are considered partial dose reconstructions.

For the March 1, 1943, to December 31, 1947 period, NIOSH concludes that external doses for 

workers directly involved with the calutron uranium enrichment and the occupational medical dose for 

all workers can be reconstructed [NIOSH 2005, 2008]. NIOSH found that external exposure, including 

occupational medical doses, could be reconstructed for all employees from January 1, 1948, to July 

31, 1979, as determined in the evaluations of SEC-00018 [NIOSH 2005], SEC-00028 [NIOSH 2006a], 

SEC-00098 [NIOSH 2008], SEC-00186 [NIOSH 2011], SEC-00251 [NIOSH 2018], and SEC-00250 

[NIOSH 2019]. 

5.2 URANIUM SOLUBILITY IN THE LUNG

The uranium compounds that have been used at Y-12 range from highly soluble to very insoluble 

[ORAUT 2007b]. Health physicists have long recognized that nephrotoxicity is the primary hazard 

associated with soluble uranium compounds (depleted through low enrichment). According to Sterner 

and Riley [1946], exposures to soluble compounds were monitored from the closing days of World 

War II by clinical tests of renal function and by fluorometric tests for uranium in urine. However, very 

few urinalysis data have been found for Y-12 before 1948.

Beginning in September 1994, the plant was placed in a stand-down mode. This effectively stopped 

all routine work (and chronic exposure potential) in the process areas. Only minimal walkthroughs and 

area checks continued during the stand down, which lasted until August 1998. The stand down 

significantly influenced the available source term for exposures. Before stand down, the Y-12 program 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 14 of 81

 

default modeling assumption was class Q (90% Super-W, 10% Y) [Barber and Forrest 1995]. During 

the stand down, the Y-12 program default assumption of chronic exposure was modified to assume 

acute exposures occurring at the midpoint of a quarterly sampling frequency. No wet chemical 

operations were performed during this 4-year period. Therefore, no soluble component was being 

produced and materials that were stopped in process were allowed to oxidize, which provided a larger 

component of type S material as an exposure source. In August 1998, a partial resumption of 

activities occurred; wet chemical operations were not restarted and were still not restarted as of 

October 2003. The predominant material encountered after partial restart in August 1998 was uranium 

oxide, which fecal sampling has shown to be more consistent with 100% type S material. Based on 

the changing workplace conditions after partial restart in 1998, the default assumptions were modified 

to return to chronic exposure but to use type S solubility [Eckerman and Kerr 1999]. 

For a workplace as varied as Y-12, it is clear that no single solubility or particle size would apply to all 

workers. Further, accurate assignment of the uranium lung absorption type to a given bioassay result 

was considered virtually impossible because of uncertainties about chemical form and limitations of 

the personnel tracking system [Barber and Forrest 1995, p. 10]. As of 2001, Y-12 implemented a 

radiation work permit (RWP) tracking system that tracks the locations of workers, their types of work, 

and the types of bioassay needed [ORAUT 2007b, p. 24]. 

Due to the many different processes and types of work performed at Y-12, and the potential range of 

solubility types associated with materials and work locations, the absorption type can be based on the 

monitoring data, assumptions that are favorable to claimants, or both.

5.2.1 Particle Size Information

Several particle size studies in uranium process areas have been conducted at Y-12 [e.g., Struxness

1953; Steckel and West 1966; Barber and Forrest 1995; Veinot 2003, Section 5]. For different times 

and different processes, reported particle sizes ranged from less than 1 to greater than 10 μm 

(physical). Steckel and West [1966] reported a positive correlation between uranium oxide particle 

size and process temperature. Barber and Forrest [1995] used an 8-μm activity median aerodynamic 

diameter (AMAD) based on particle size measurements as the basis for the class Q dosimetry system 

used in the 1990s. (The class Q system is described as 10% class Y and 90% modified class W. The 

modification consists of increasing the class W 50-d compartment to 120 days.)

Since 2000, Y-12 has implemented the latest guidance provided by the ICRP including the 

Publication 66 [ICRP 1994] lung model and Publication 78 [ICRP 1998] methods and models. This is 

the justification for the change to the 5-μm AMAD as the default particle size. Y-12 uses the newer 

ICRP Publication 60 [ICRP 1991] weighting factors [Snapp 2003b, p. 119]. For Y-12 dose 

reconstructions, the default particle size distribution is 5 μm AMAD.

5.2.2 Uranium Alpha Activity as a Function of Enrichment

At Y-12, uranium enrichments range from depleted uranium (DU; less than 0.71 wt% 235U) to very 

highly enriched uranium (VHEU; greater than 93 wt% 235U). Section 5.2.4.1 lists the uranium isotopic 

mixes with RU contaminants that were typical of uranium materials at Y-12. The potential dosimetric 

complexity is simplified by the similarity of the dose conversion factors (DCFs) for the most important 

uranium isotopes and by the quantities measured in the in vitro bioassay and chest-counting 

programs [BWXT Y-12 2000]. 

For work with enriched compounds and alloys, urinalysis programs after 1950 reported either total 

uranium alpha counts or isotopic uranium results, both of which reflect the dosimetric potential of the 

isotopes generally handled [AEC 1959]. Early in vivo chest-counting results were reported either as 

μg of 235U or mg of 238U based on whether an exposure was believed to be from enriched uranium 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 15 of 81

 

(EU), natural uranium (NU), or DU [Cofield 1959a; Scott and West 1975]. Based on these reporting 

techniques, the assumptions in Table 5-1 can be made about data interpretation for the periods from 

1948 to 1989 for urinalysis and from 1961 to 1989 for lung counting.

Table 5-1. Enrichment assumptions for dose assessment for analytical techniques.

a

Analytical technique 

(urinalysis 1948–present) 

(lung count 1961–1989) Measurement information

Default enrichment 

assumption

Urine by fluorometry (usually for an 

individual who worked in areas with 

NU or DU; before 1950, also used 

for workers in EU areas).

Alpha dpm can be calculated from 

mass. 

93% before 1950, NU from 

1950 to 1989

Urine by gross alpha counting 

(usually for an individual who 

worked in areas with EU).

Alpha dpm is the sum of all uranium 

alphas.

93% enrichment

Urine by alpha spectrometry

(10/1989–present).

Alpha dpm/d is reported isotopically 

and total uranium is determined by 

summing isotopic results based on 

the guidance in ORAUT 2018. 

93% enrichment

Lung count μg U-235 93% enrichment

Lung count μg U-238 NU

a. Sources: AEC [1959]; Cofield [1959a]; Scott and West [1975].

For lung counts, a combination of the information in the Type Analysis and Material Type (see 

Section A.1.3 in Attachment A for details) reporting fields can be used to determine if the count was 

believed to be due to NU or DU. For records through 1971, a Type Analysis of 1 with a Material Type 

of 2 or 3 indicates DU, while a Material Type of 7 indicates NU. For records after 1971, a Type 

Analysis of 4 with a Material Type of 2 or 3 indicates DU, while a Material Type of 7 indicates NU.

For dose reconstruction purposes, the individual isotopes of a uranium mixture intake (234U, 235U, 236U, 

and 238U) can be summed under certain circumstances as outlined in ORAUT-OTIB-0060, Internal 

Dose Reconstruction [ORAUT 2018]. Note that 234U DCFs are not appropriate substitutes for 232U, as 

the DCFs for 232U are typically 5-10 times larger than those for 234

U.

5.2.3 Temporal Pattern of Uranium Exposures

As a rule, routine uranium exposures at Y-12 were considered to be of a chronic nature. For example, 

AEC [1959, p. 70] stated, “...our interpretation of urinalysis results and our assignment of internal 

dose assume an exposure under equilibrium conditions of intake and elimination.” For the stand-down 

period from September 1994 until August 1998, acute exposures should be assumed to be the more 

likely mode of exposure. After this time, while complete equilibrium is not expected in modern internal

dosimetry models, the presumption of chronic exposure conditions for uranium remains in place as 

noted in a 2003 report: “The most likely exposure potential for uranium work at the Y-12 Complex is 

chronic in nature” [Snapp 2003b, p. 111]. Because workers were assigned to various buildings and 

processes throughout their employment, it is not practical to characterize intakes based on historical 

operational assignments. The claimant files do not generally contain detailed work assignments, but 

do contain all bioassay data for each worker during the years of operational exposure – primarily in 

vitro (urine) and in vivo (lung counts) bioassay. Radiological materials, including isotopic mixes and so 

forth, are not defined by operational area or building in this TBD, because internal dose reconstruction 

relies primarily on the recorded bioassay data and the defaults for unmonitored isotopes that were 

conservatively established on a plantwide basis (see Table 5-8 in Section 5.2.4.1 for plantwide RU 

contaminant defaults).
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 16 of 81

 

5.2.4 Other Radionuclides of Concern Including Recycled Uranium Contaminants

In addition to 234U, 235U, and 238U, the following radionuclides were identified in the Y-12 Technical 

Basis Document for Internal Dosimetry [Snapp 2003b]: 

• 3

H, 90Sr, and 99Tc; 

• 228Th and 232Th; 

• 232U, 233U, and 236U; 

• 238Pu, 239Pu, and 241Pu; and

• 237Np and 241Am. 

Radionuclides that could interfere with in vivo analysis of uranium and thorium (i.e., 40K and 137Cs) 

were quantified so their effects on the spectra could be taken into account.

Other radionuclides addressed included 60Co and 95Zr/95Nb for organizations outside Y-12 [Cofield 

1961]. 

Many of these radionuclides are accompanied by progeny in various stages of equilibrium. In addition, 

operators of the in vivo analysis equipment reviewed accumulating spectra for interferences from 

medical radioisotopes such as 131I, 99mTc, 67Ga, and 201

Tl.

5.2.4.1 Unmonitored Radionuclides from Recycled Uranium

One of the primary sources of significant contaminants (including the production of other than the 

normal uranium isotopes) that were introduced into the Y-12 plant systems were those associated 

with RU from 1953 until 1999 [BWXT Y-12 2000]. These contaminants were a result of fission and 

activation processes of a variety of uranium enrichment isotope mixtures that were irradiated in 

production and test reactors. After completing their useful life in the reactors, the unused uranium in 

the spent fuel elements or targets was recovered in chemical extraction plants and returned to the 

inventories in the DOE system along with trace quantities of the contaminants (in some instances, 

236U was produced in quantities greater than “trace” levels) [BWXT Y-12 2000]. 

The major source of RU contaminants at Y-12 was the receipt and processing of returned VHEU from 

the Savannah River Site (SRS) and the Idaho Chemical Processing Plant (ICPP) [Rich et al. 2000]. 

Tables 5-2 and 5-3 document the quantities of uranium and the calculated mass quantities of 

plutonium, neptunium, and technetium actually delivered to Y-12. Table 5-4 presents a computergenerated listing of the relative production of a few isotopes of concern, following irradiation of VHEU 

fuel in a high-flux test reactor. The process streams included a spectrum of chemical forms of 

uranium, including uranyl nitrate [UO2(NO3)2] solutions, uranium trioxide (UO3), uranium-aluminum 

(U-Al) alloy ingots, uranium scrap, uranium tetrafluoride (UF4), uranium metal, solvent extraction 

raffinates, and a variety of secondary process wastes and residues [Rich et al. 2000; DOE 2003]. 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 17 of 81

 

Table 5-2. Y-12 uranium receipts summary presenting average levels of three predominant RU 

contaminants through plant history, 1953 to 1999.a

Y-12 VHEU receipts summary

Site

U 

(kg)

Pu 

(g)

Pu 

(ppb)

Pu 

(pCi/μg U)

Np 

(g)

Np

(ppm)

Np 

(pCi/μg U)

Tc 

(g)

Tc 

(ppm)

Tc 

(pCi/μg U)

SRS 125,161 0.0455 0.36 5.2E−03 3,600 29 2E−02 14,268 114 1.9

ICPP 25,696 0.00124 0.05 7.2E−04 66 3 2E−03 231 9 0.15

EU totals/avg. 150,857 0.04674 0.3 4.3E−03 3,666 23 1.6E−02 14,499 96 1.64

Y-12 DU receipts summaryb

Site U (kg)

Hanford 1,502

Oak Ridge Gaseous Diffusion Plant 192,836

Paducah Gaseous Diffusion Plant 38,423

DU total 232,761

Grand total 383,618

a. Source: Rich et al. [2000].

b. No contaminant data for DU from these sites.

Table 5-3. Total Y-12 RU receipts summary.a

Shipping site U (kg)

SRS 153,000

ICPP 25,700

Hanford 4,400

West Valley 900

Total 184,000

Total

U 

(kg)

Pu 

(g)

Pu 

(ppb)

Pu 

(pCi/μg U)

Np 

(g)

Np 

(ppm)

Np 

(pCi/μg U)

Tc 

(g)

Tc 

(ppm)

Tc 

(pCi/μg U)

Totals/averages 184,000 0.5 2.7 0.039 54 0.003 0.00021 9,100 49.5 0.85

a. Source: DOE [2003].

The uranium enrichments ranged from DU, NU, low-enriched uranium (LEU - less than 20% 235U), 

highly enriched uranium (HEU - from 20 to approximately 93% 235U), and VHEU (more than 93% 235U) 

[Rich et al. 2000]. The predominant RU and associated contaminants were in HEU and VHEU 

materials. However, all the uranium at Y-12 came from other DOE facilities, which had either 

generated or received RU materials [Rich et al. 2000]. Therefore, nearly all of the uranium in the DOE 

facilities contained RU contaminants to varying degrees through being processed in the same 

equipment, blending with other materials to adjust the degree of enrichment, etc. For example, most 

of the DU received and processed at Y-12 was seldom chemically processed but was received in 

forms from which parts were produced by mechanical processing. However, even these materials 

were received from other plants, such as the Feed Materials Production Center (FMPC) in Fernald, 

Ohio, which also had an RU contaminant inventory. Therefore, the DU contaminant levels at Y-12 

were inferred from the FMPC and Idaho Specific Manufacturing Capability (SMC) projects recycle 

reports [ORAUT 2017; Lewis et al. 2000]. 

The fundamental conclusion from a variety of reports, including a lack of definitive RU contaminant 

bioassay data, is that for dose reconstruction purposes a default level of RU contaminants should be 

derived and applied as a percentage increase to the derived uranium intake for each of the four major 

contaminants. Analytical information derived from a variety of sources enables the calculation or 

interpolation of the levels of the predominant RU contaminants in the uranium materials received, 

processed, and handled at Y-12. One of the primary sources of RU contaminant information was the 

report generated as a consequence of a comprehensive study of RU in DOE facilities in 2000. The 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 18 of 81

 

report for Y-12 was Recycled Uranium Mass Balance Project Y-12 National Security Complex Site 

Report [BWXT Y-12 2000]. There were some recognized inconsistencies in shipping and receipts 

among the DOE facilities in the DOE 2000 RU reports. In an attempt to resolve these inconsistencies,

a summary report was issued by the DOE Office of Security, Recycled Uranium, United States 

Production, Enrichment, and Utilization [DOE 2003]. Although it corrected some inconsistencies, there 

remain significant inconsistencies between the two reports. For reconstruction of RU contaminant 

inventories in the Y-12 facilities and internal dose default analyses, the rationale for resolving those 

inconsistencies is provided herein. One approach to gain an estimate of the average contaminant 

concentrations through the years is to document the total receipts of contaminants and uranium and 

then to derive the average long-term RU contaminant concentrations.

Tables 5-2 and 5-3 list the documented RU contaminants and their primary source in recycled VHEU. 

For completeness, other DU receipts are listed in the BWXT Y-12 [2000] report without documenting 

the associated RU contaminant levels. However, these uranium receipts also contained low-level RU 

contaminants, the concentrations of which are discussed later in this report.

The results of the DOE [2003] report are presented in Table 5-3. As indicated, there are significant 

differences with the results recorded in the BWXT Y-12 [2000] report (summarized in Table 5-2) and 

those in Table 5-3. The objective of DOE [2003] was to resolve the differences in the reported 

shipment amounts from the recycling plants (the “primary sites”) with the reported receipts at the first 

or “primary receiving sites.” In addition, the detail in DOE [2003] related to uranium enrichment and 

specific RU contaminant concentrations from each site was not provided (note the lack of detail in 

Table 5-3). Examination of DOE [2003] indicates the shipments from these primary receiving sites 

(following processing and incorporation of recycled materials into their process streams) to what can 

be called “secondary” receiving sites were not documented. For example, Y-12 received uranium 

materials from gaseous diffusion plants, the FMPC, and others, which also had RU contaminant 

inventories. Thus, the discrepancies in the values reported in BWXT Y-12 [2000] and DOE [2003] 

could result from documenting the primary shipment and receipts in DOE [2003], while BWXT Y-12 

[2000] accounts for secondary shipments as well. However, where large differences occur, the value 

most favorable to the claimant value is used to derive the defaults for dose reconstruction, which are 

generally the results in BWXT Y-12 [2000]. 

Though plutonium, neptunium, and technetium isotopes were analyzed and documented in all the 

facilities as the primary isotopes of concern, there was a significant level of 228Th detected in nearly all 

the facilities at Y-12 and in reported process stream analyses [BWXT Y-12 2000]. To establish the 

source of this isotope and provide a basis for including it in the RU contaminants to be considered in 

internal dose reconstruction, Tables 5-4 and 5-5 demonstrate the production of this isotope in VHEU 

fuels and the relative/effective enrichment during chemical reprocessing. The values in Table 5-4 

were derived from the Rich et al. [2000] RU report at the ICPP. The ORIGIN2 computer program 

results illustrate the production of isotopes during use of VHEU fuels [Wenzel 2000, 2004]. The 

isotopes of interest for this illustration are the uranium isotopes and plutonium, neptunium, 

technetium, and thorium. Thorium-228 is a contaminant that became an isotope of concern at Y-12, 

which was the primary receiving site of recycled VHEU fuel materials [BWXT Y-12 2000]. 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 19 of 81

 

Table 5-4. Primary contaminants (at 3-year decay) in VHEU fuels and of concern in internal dose 

reconstruction.a

Isotope Half-life

Activity

(μCi/μg total U)

Activity

(μCi/g total U)

Percent weight 

g/100 g Ub ppm U

Activity ratio

(isotope/

total U)

Tc-99 2.11E+05 yr 3.8E−04 3.8E+02 2.2E+00 2.2E+04 2.4E+01

Th-227 18.72 d 3.3E−11 3.3E−05 1.1E−13 1.1E−09 Not applicable

Th-228 1.913 yr 1.6E−07 1.6E−01 1.9E−08 1.9E−04 1.0E−02

Th-229 7.34E+03 yr 1.4E−12 1.4E−06 6.7E−10 6.7E−06 Not applicable

Th-230 7.7E+04 yr 1.9E−11 1.9E−05 9.6E−08 9.6E−04 1.2E−06

Th-231 25.52 hr 1.7E−06 1.7E+00 3.2E−10 3.2E−06 1.0E−01

Th-232 1.41E+10 yr 3.5E−15 3.5E−09 3.2E−06 3.2E−02 2.2E−10

Total Th Not applicable Not applicable 1.9E+00 Not applicable Not applicable Not applicable

U-232 6.89E+01 yr 2.7E−07 2.7E−01 1.3E−06 1.3E−02 Not applicable

U-233 1.59E+05 yr 2.6E−10 2.6E−04 2.7E−06 2.7E−02 Not applicable

U-234 2.46E+05 yr 6.6E−07 6.6E−01 1.0E−02 1.0E+02 Not applicable

U-235 7.04E+08 yr 1.7E−06 1.7E+00 7.8E+01 7.8E+05 Not applicable

U-236 2.34E+07 yr 1.3E−05 1.3E+01 2.0E+01 2.0E+05 Not applicable

U-237 6.75 d 3.6E−08 3.6E−02 4.4E−11 4.4E−07 Not applicable

U-238 4.47E+09 yr 5.6E−09 5.6E−03 1.7E+00 1.7E+04 Not applicable

Total U Not applicable Not applicable 15.7E+00 Not applicable Not applicable Not applicable

Np-236 1.15E+05 yr 9.4E−11 9.4E−05 7.2E−07 7.2E−03 Not applicable

Np-237 2.14E+06 yr 9.1E−06 9.1E+00 1.3E+00 1.3E+04 6.0E−01

Np-238 2.117 d 9.2E−10 9.2E−04 3.6E−13 3.6E−09 Not applicable

Np-239 2.355 d 4.3E−08 4.3E−02 1.9E−11 1.9E−07 Not applicable

Total Np Not applicable Not applicable 9.1E+00 Not applicable Not applicable Not applicable

Pu-236 2.851 yr 3.6E−06 3.6E+00 6.8E−07 6.8E−03 0.2E+00

Pu-238 87.70 yr 3.6E−02 3.6E+04 2.1E−01 2.1E+03 2.3E+03

Pu-239 2.41E+04 yr 1.9E−05 1.9E+01 3.1E−02 3.1E+02 1.2E+00

Pu-240 6.54E+03 yr 1.5E−05 1.5E+01 6.4E−03 6.4E+01 1.0E+00

Pu-241 14.4 yr 1.5E−03 1.5E+03 1.4E−03 1.4E+01 1.0E+02

Pu-242 3.73E+05 yr 8.8E−09 8.8E−03 2.3E−04 2.3E+00 5.6E−04

Total Pu Not applicable Not applicable 4.6E+04 Not applicable Not applicable Not applicable

Am-241 4.32E+02 yr 9.6E−06 9.6E+00 2.8E−04 2.8E+00 6.0E−01

a. Sources: Wenzel [2000, p 104 (Zr column)]; Wenzel [2004, pp. 39, 41, 42, 43, 45, 46]. 

b. There were basically three types of fuel elements with varying enrichments (both at the start and after burnup). The 

typical after-burnup enrichments were 78.21% for aluminum-clad, 78.42% for zirconium-clad, and 51.9% for 

stainless-steel fuel elements. Values represent the 78% after-burnup enrichments as representative, with typical 

1%+ for U-234. 

Table 5-5. Comparison of the production of three selected isotopes in VHEU spent fuel with 

levels reported in HEU shipments to Y-12.

Isotope

μCi/g U product 

from SRS

μCi/g U product 

from ICPP

μCi/g U 

in high-burnup 

VHEU reactor fuelb 

Approximate derived 

DFc,d 

Th-228 0.01–0.03 Not applicablea 0.16 10

Np-237 0.01–0.02 1E−03 9.1 1E+03–1E+04

Pu-238 4E−03 3E−04 3.6E+04 1E+07–1E+08

a. There were no analytical data for Th-228 reported in the ICPP Y-12 receipt reports.

b. These values were taken directly from column 4 in Table 5-4.

c. DF = decontamination factor. 

d. The DF is interpolated from Y-12 process stream analytical data.

Comparing the concentrations in the spent fuels with the reported concentrations in the product 

received at Y-12 provides a rough indication of the decontamination factors (DFs) of the chemical 

extraction processes. Reference to these DFs in Table 5-5 and the average contaminant 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 20 of 81

 

concentrations in Table 5-6 indicate that plutonium is the most efficiently removed with an effective DF 

of approximately 1 × 107 to 108

, with a neptunium DF of approximately 1 × 104, while thorium appears 

to follow the chemistry of uranium more closely, resulting in a DF of only 10, which in turn results in a 

relative enrichment of 228Th in the VHEU product received at Y-12. Although 228Th has not been 

considered an RU contaminant of concern at other DOE facilities and was not routinely analyzed, 

there were sample analyses of contaminants in surface smears and wipes in most of the Y-12 plant 

facilities in 1997 and 2001 that included 228Th [Oliver 1997; BWXT Y-12 2001]. These surface 

contamination surveys of a large number of Y-12 facilities indicate a predominance of 228Th as a 

contaminant [BWXT Y-12 2001]. The uranium-to-nonuranium alpha ratios range from 4 to more than 

1 × 104

, which is not useful for dose reconstruction. However, these data do indicate that 228Th is one 

of the primary contaminants at Y-12, which in turn provides a technical base for inclusion of 228Th in 

the list of RU contaminant isotopes.

Table 5-6. Ranges of reported contaminant values in the main processing 

facilities.

a,b

Contaminant Range

Average of 

range

Average of range

minus the high

Raffinate 

valuesc

Pu 0.11–4.5 ppb 0.5 ppb 0.25 ppb 62 ppb

Np 4.7–346 ppm 30.6 ppm 20.8 ppm 2,980 ppm

Tc 0.13–211 ppm 83 ppm 66 ppm 641 ppm

a. Source: BWXT Y-12 [2000, pp. 156–157]. 

b. The highest values reported in the processing facilities were in just 1 of 17 process 

streams: the solvent extraction purification of NU. Uranium levels are still significant in this 

stream, which indicates a marked enrichment of contaminants in relation to U in this 

stream.

c. The raffinate streams are depleted in uranium to a range of 5 ppm U levels in the sludge.

Therefore, the ppm of the RU contaminants in relation to U are not meaningful (i.e., total 

Pu – maximum 30 and average 21 pCi/g net sludge; Np – maximum 12 and average 8 

pCi/g sludge; Tc – maximum 1.2E+4 and average 3.7E+3 pCi/g sludge).

Uranium has always been the dominant contributor to collective internal dose at Y-12. Monitoring for 

other radionuclides has been performed on a limited basis [Snapp 2003b, p. 68]. Both the relatively 

small concentrations and the difficulty of analyses have contributed to the lack of data. The isotopic 

characterization of the large variety of uranium enrichments/contaminants is listed in Table 5-7, which 

is an average representation of six specific generalized enrichment types handled at Y-12. The 

uranium-to-nonuranium alpha ratios in Table 5-7 are in the same general range as those measured in 

surface contamination in the plant facilities after years of operation. From the values in Table 5-6, it is 

evident that the maximum of the ranges in the plant tended to run a factor of approximately 10 above 

the average of the range for both plutonium and neptunium, with technetium at a factor of 3. A factor 

of 10 is also applied to the thorium levels. The defaults in Table 5-8 were chosen according to all the 

information discussed above to represent the maximum values of the documented RU contaminant 

levels in the Y-12 process streams and, thus, represent values that are favorable to claimants and 

that can apply on a plantwide basis. Application of the default level should be based on the indication 

by the bioassay of the likelihood of VHEU/HEU or LEU/NU/DU. The appropriate default ratio should 

be chosen based on the 235U or 238U indicated in the bioassay results. For each μg or pCi of uranium 

intake calculated based on bioassay data in the claimant files, an additional intake of each of the four 

RU contaminants should be added.
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 21 of 81

Table 5-7. Typical uranium enrichment materials at Y-12 with calculated and inferred RU contaminant 

levels.a

High burn-up (~50%) recycled VHEU from SRSb

Uranium 

isotope

Activity 

(Ci/g) Mass fraction

Activity

(pCi/μg total U)

233 9.66E−03 1.000E−04 9.660E−01

234 6.22E−03 1.280E−02 7.962E+01

235 2.16E−06 5.220E−01 1.128E+00

236 6.50E−05 2.920E−01 1.898E+01

238 3.36E−07 1.730E−01 5.796E−02

Totals Not applicable 9.990E−01 1.007E+02

dpm/μg total U Not applicable Not applicable 2.24E+02

RU 

contaminant

Activity

(pCi/μg total U)

Pu-238 6.20E−03

Pu-239 1.00E−04

Np-237 2.00E−03

Tc-99 1.90E+00

Th-228 3.17E−02

Total α 4.00E−02

α dpm/μg total U 8.87E−02

U α:Non U α 2.52E+03

 

Lower burn-up (~25%) recycled VHEU from ICPPb

Uranium 

isotope

Activity 

(Ci/g) Mass fraction

Activity

(pCi/μg total U)

233 9.66E−03 3.000E−06 2.90E−02

234 6.22E−03 1.000E−02 6.22E+01

235 2.16E−06 7.800E−01 1.68E+00

236 6.50E−05 2.000E−01 1.30E+01

238 3.36E−07 1.700E−02 5.71E−03

Totals Not applicable 1.007E+00 7.69E+01

dpm/μg total U Not applicable Not applicable 1.71E+02

RU 

contaminant

Activity

(pCi/μg total U)

Pu-238 2.00E–03

Pu-239 3.30E–05

Np-237 2.00E–03

Tc-99 1.50E–01

Th-228 5.00E–02

Total α 5.40E–02

α dpm/μg total U 1.20E–01

U α:Non U α 1.4E+03

Weapons-grade Oralloy VHEUb,c

Uranium 

isotope

Activity 

(Ci/g) Mass fraction

Activity

(pCi/μg total U)

233 9.66E–03 0.000E+00 0.00E+00

234 6.22E–03 1.100E–02 6.84E+01

235 2.16E–06 9.300E–01 2.01E+00

236 6.50E–05 7.500E–03 4.88E–01

238 3.36E–07 6.000E–02 2.02E–02

Totals Not applicable 1.009E+00 7.09E+01

dpm/μg total U Not applicable Not applicable 1.57E+02

RU 

contaminant

Activity

(pCi/μg total U)

Pu-238 2.00E−02

Pu-239 2.00E−02

Np-237 6.00E−02

Tc-99 5.00E−02

Th-228 5.00E−02

Total α 1.50E−01

α dpm/μg total U 3.33E−01

U α:non-U α 4.73E+02

Recycled NUd

Uranium 

isotope

Activity 

(Ci/g) Mass fraction

Activity

(pCi/μg total U)

233 9.66E−03 0.000E+00 0.00E+00

234 6.22E−03 5.400E−05 3.36E−01

235 2.16E−06 7.200E−03 1.56E−02

236 6.50E−05 0.000E+00 0.00E+00

238 3.36E−07 9.927E−01 3.34E−01

Totals Not applicable 9.999E−01 6.85E−01

dpm/μg total U Not applicable Not applicable 1.52E+00

RU 

contaminant

Activity

(pCi/μg total U)

Pu-238 0.00E+00

Pu-239 3.00E−06

Np-237 2.40E−05

Tc-99 2.00E−04

Th-228 3.00E−05

Total α 5.70E−05

α dpm/μg total U 1.27E−04

U α:non-U α 1.20E+04
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 22 of 81

Recycled LEU (2%)d

Uranium 

isotope

Activity 

(Ci/g) Mass fraction

Activity

(pCi/μg total U)

233 9.66E−03 0.000E+00 0.00E+00

234 6.22E−03 2.000E−04 1.24E+00

235 2.16E−06 2.000E−02 4.32E−02

236 6.50E−05 0.000E+00 0.00E+00

238 3.36E−07 9.798E−01 3.29E−01

Totals Not applicable 1.000E+00 1.62E+00

dpm/μg total U Not applicable Not applicable 3.59E+00

RU 

contaminant

Activity

(pCi/μg total U)

Pu-238 0.00E+00

Pu-239 2.00E−04

Np-237 2.00E−04

Tc-99 9.20E−02

Th-228 3.00E−04

Total α 7.00E−04

α dpm/μg total U 1.55E−03

U α:non-U α 2.31E+03

Recycled DUd

Uranium 

isotope

Activity 

(Ci/g) Mass fraction

Activity

(pCi/μg total U)

233 9.66E−03 0.000E+00 0.00E+00

234 6.22E−03 1.000E−05 6.22E−02

235 2.16E−06 2.000E−03 4.32E−03

236 6.50E−05 3.000E−06 1.95E−04

238 3.36E−07 9.980E−01 3.35E−01

Totals Not applicable 1.000E+00 4.02E−01

dpm/μg total U Not applicable Not applicable 8.93E−01

RU 

contaminant

Activity

(pCi/μg total U)

Pu-238 0.00E+00

Pu-239 3.70E−06

Np-237 2.70E−05

Tc-99 4.00E−04

Th-228 4.00E−05

Total α 7.07E−05

α dpm/μg total U 1.57E−04

U α:non-U α 5.69E+03

a. Values derived from a number of sources and should be considered as largely average values.

b. The weapons-grade Oralloy VHEU and other burn-up example VHEU were derived from Y-12 data and ICPP reports.

c. The Oralloy RU contaminant data were presented in the reference reports in very general terms (i.e., <0.1 μCi/g U for

total Np and Pu and <0.05 μCi/g for “any other radionuclide”).

d. The recycled DU, recycled NU, and recycled LEU values were derived from FMPC data because this site supplied and

received uranium to and from all DOE sites, including Y-12. In addition, the recycled DU information was verified by the

analytical data reported by the Idaho SMC facility in Lewis et al. [2000].

Table 5-8. Recommended defaults based on upper levels of expected ranges.

a

Isotope

VHEU & Oralloy 

(pCi/μg U)

LEU, NU, & 

DU (pCi/μg U)

VHEU & Oralloy 

(pCi/pCi total U)b

LEU, NU, & DU

(pCi/pCi total U)b

Tc-99 5.7 0.3 5.7E–2 0.15

Th-228c 0.5 0.003 5E–03 1.5E–03

Np-237 0.6 0.003 6E–03 1.5E–03

Pu-238d 0.2 0 2E–03 0

Pu-239d 0.2 (Oralloy) 0.003 2E–03 (Oralloy) 1.5E–03

Total non-U 1.3 0.01 Not applicable Not applicable

Total U 75 2 Not applicable Not applicable

a. For Table 5-8 maximum values, take the highest Table 5-7 value (the average) and multiply by 10 for

Pu, Th, and Np and by 3 for Tc. See page 26 of this TBD.

b. Reference to Table 5-7 indicates activity in pCi per gram of total U for VHEU ranges from 71 to 101;

100 is used for simplicity. For LEU, NU, and DU, the pCi per gram of total U ranges from 0.9 to 1.6;

2 pCi/g total U is used for simplicity.

c. The Th-228 default is justified in concept due to the demonstrated existence of general plantwide

contamination and further indicated by the demonstrated production in VHEU fuels, which is further

validated in the process stream analyses documented in BWXT Y-12 [2000].

d. VHEU was used in the metal form at Y-12 and called Oralloy. Plutonium analyses for Oralloy were

not isotope-specific. For Oralloy VHEU only, Pu-239 is the chosen isotope because the total

quantities of recycled VHEU (in which Pu-238 was the predominant plutonium isotope) returned to

the gaseous diffusion plants was miniscule compared to the LEU returns [BWXT Y-12 2000].

From a practical and simplified approach, when the results in the claimant files are given in mass units 

of uranium per L or lung count, the default levels in pCi per gram of total U in Table 5-8 can be used 

to add the RU contaminant activity to the intake. However, if the bioassay units are in uranium alpha 

activity per unit urine volume and/or U isotope activity units, the default levels in pCi/pCi total U in that 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 23 of 81

 

table can be used. Again, these values are maximum values expected, while average values are 

basically a factor of 10 less than the values for plutonium, neptunium, and thorium and a factor of 3 

less than those for technetium. If a best estimate dose reconstruction is needed, average values (a 

factor of 10 less than the values in Table 5-8 for plutonium, neptunium, and thorium and a factor of 3 

less than those for technetium) should be used.

In reference to Table 5-8 with relationship to the plutonium contaminant, the recommended isotope 

239Pu should be assumed for Oralloy only, because Oralloy at Y-12 did not come directly from the 

primary RU generating sites (i.e., 239Pu came more predominantly from the gaseous diffusion plants

rather than SRS or ICPP by a factor of 1,000 as indicated in Table 5-7) [BWXT Y-12 2000]. In most 

instances the enrichment of uranium is unknown or varies, therefore, the assumption of Oralloy can 

be used as a default for the majority of dose reconstructions as a favorable to claimant assumption.

The existence of air contamination and personnel exposures from 232Th operations is not addressed in 

this section. This section deals only with RU contaminants.

5.3 IN VITRO URINE ANALYSIS

5.3.1 Coverage

Uranium enrichment activities began in the fall of 1943. A uranium urinalysis program based on 

fluorometry was used in conjunction with medical examinations to monitor for kidney damage from 

exposure to soluble uranium compounds [Sterner and Riley 1946]. 

Insufficient information exists on radiological activities for dose reconstruction during the period from 

March 1943 through December 1947, including the development of beneficial radiological isotopes, 

development and testing of a neutron monitor, maintenance and use of a large 226Ra sealed source, 

and thorium extraction.

Beginning from 1948, the occupational environmental internal dose from uranium can be calculated 

using data in ORAUT-TKBS-0014-4, Y-12 National Security Complex – Occupational Environmental 

Dose [ORAUT 2006] for unmonitored workers. Additionally, uranium co-exposure intakes are 

available starting in 1948 for unmonitored workers as discussed in Attachment B. 

The earliest fluorometry data was dated 1948. In 1950, workers in production areas were placed on a 

uranium fluorometry urinalysis program for estimating internal exposure. The program was expanded 

to include certain maintenance workers in 1954 [McLendon 1960, p. 3]. As of 2003, Y-12 requires 

workers with a potential for internal exposures in excess of 100 mrem/yr committed effective dose 

equivalent (CEDE) to participate in the bioassay program [Snapp 2003b, p. 27]. Until September 

1989, routine urinalysis focused on EU, DU, NU, tritium, and plutonium. Beginning in October 1989, 

uranium results were no longer classified as EU, DU, or NU. Rather, they were reported as isotopic 

results based on alpha spectrometry analysis [Souleyrette 2003]. Analyses for other radionuclides 

were performed on an as-needed basis.

5.3.2 Sample Collection

Sample Volumes

For most of the plant’s history, the primary urine collection method was a spot sample submitted 

Monday morning before entering the work area; that is, routine samples were submitted after a 

minimum of a 48-hour absence from the work area. (The July 1 to December 31, 1951, Health 

Physics Progress Report [Ballenger 1952, p. 28] stated that Friday evening samples would be 

discontinued in favor of Monday morning samples.) The fraction of the daily void volume was 

estimated on the basis of the time between the sample void and the previous void. This rate method
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 24 of 81

 

of estimating daily void volumes was used explicitly in the calculation of the daily radionuclide 

excretion.

For example, the following formula was given for calculating EU excretion by electrodeposition and 

gross alpha counting [McRee et al. 1965, p. 24]: 

dpm/d AP Eff Vol TI =× ÷ × ÷ 8 ( ) ( ) (5-1)

where

8 = a constant incorporating time, count, and volume constants, including the 20-mL 

electroplating aliquot volume. (There is no information in McRee et al. [1965] on what 

the constant includes or what the daily excretion was.) 

AP = average number of counts in 30 minutes on a plate (disk)

Eff = uranium recovery efficiency (%)

Vol = volume of total void (mL)

TI = time between sample void and previous void (hr)

The term (Vol ÷ TI) also appears in the corresponding equation for fluorometric determinations of NU 

and DU [McRee et al.1965, p. 26]. The use of the rate method to estimate daily urinary excretion (and 

hence, radionuclide elimination) contributed to the uncertainty associated with any given 

measurement; the corresponding detection level (LD as defined in Currie [1968]) is discussed below.

Since 1989, routine samples have been collected over a 24-hour period, typically while the worker 

was on a scheduled break from the workplace. Many workers have elected to submit a simulated

24-hour sample. This sample is obtained by collecting the last void in the evening (before retiring), 

any urine excreted during the night, and the first void the following morning. This procedure is 

repeated for two consecutive nights [Snapp 2003b, p. 74]. 

Routine Urine Sample Frequency

Urine samples were collected monthly in 1950, and weekly collection for some workers began in 

1951. By 1963, health physics personnel were basing the frequency of participation in the urinalysis 

program for each department on the most recent urinalysis results of that department [McLendon 

1963]. The 1963 edition of the Y-12 Radiation Safety Manual [McLendon 1963] stated that the 

frequency of participation schedules was reviewed monthly and adjusted semiannually to meet the 

following criterion: “Sample at the frequency necessary to assure, with at least 95% confidence, that 

95% of the individuals in a department have a quarterly average below the plant action limit.” This 

criterion was used within the limitations of a maximum frequency of once per week and a minimum 

frequency of once per quarter. Since the late 1980s, most personnel have been on a quarterly 

frequency [Ashley et al. 1992, p. 30]. 

As of 2003, samples are scheduled based on the RWP use rather than a default frequency. Only if an 

RWP was used will a person be scheduled for a sample. If the RWP specified a fecal sample, a 

paired set of urine and fecal samples is scheduled approximately 60 days after entry date. If the RWP 

specified urine only, a urine sample is scheduled approximately 90 days after entry date [Snapp

2003b, p. 26]. 

5.3.3 Minimum Detectable Activities

Normal and Depleted Uranium in Urine (1945 to 1989) and

Enriched Uranium (1948 to 1950) by Fluorometry

In fluorometry, the visible radiation emitted from a doped sodium-fluoride bead illuminated by 

ultraviolet radiation is measured. Poisson counting statistics used in nuclear particle counting do not 
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apply to this procedure. The detection limits were historically determined by testing the performance 

of a particular configuration against standards of known content. Early Health Physics progress 

reports give the minimum detectable limit as 5 ppb, which is a minimum detectable activity (MDA) of 

7 μg/d for a nominal 1.4 L/d urine excretion rate [Struxness 1949a]. In 1952, the instrument sensitivity 

and sample volumes indicated an MDA of 7 μg/d for 1.4 L/d of urine excreted [Struxness 1953, p. 28]. 

The activity was calculated from the fluorometric mass reading for NU using a specific activity of 

1.55 dpm/μg. For EU (93 wt% 235U), a specific activity of 150 dpm/μg was used. The corresponding 

limits of detection are 11 dpm/d for NU and 1,100 dpm/d for EU (93 wt% 235U). At times, procedures 

called for the direct conversion of fluorometer current into disintegrations per minute, implicitly omitting 

the mass calculation. Whether the results were expressed in terms of mass or activity, the method 

remained essentially stable until replaced by long-duration alpha spectroscopy in 1989. Therefore, 

from 1945 through 1989, the MDA was 7 μg/d or 11 dpm/d for NU. From 1948 to 1950, the MDA was 

7 μg/d or 1,100 dpm/d for EU (93 wt% 235U). Additional variation resulted from individual urine 

excretion volumes. Given the limitations of the rate method of estimating daily urine volumes, 

uncertainty in the excretion volume is likely to contribute significantly to the uncertainty associated 

with the detection limit of a single measurement.

Enriched Uranium in Urine by Gross Alpha Counting (1950 to September 1989)

Determination of the MDAs for this method is complicated by the standard counting method used.

From the mid-1950s to 1989, each sample consisted of two silver disks, each containing uranium 

electroplated from separate 20-mL raw urine aliquots drawn from the container submitted by the 

worker. Each disk was counted twice (on two different proportional counters) for 30 minutes per count.

If the two results from a single disk did not agree within tabulated limits, a third count was made [UCC 

1966, p. 9] and the two most concordant counts were used. If the average results of the two disks 

from the same sample did not agree within specified limits, two more plates were prepared, volume 

permitting [UCC 1966, p. 9]. The expression of potentially censored data of this sort in terms of formal 

detection limits is not straightforward, and no detailed analysis of the statistics of this process has 

been found. However, an approximate indicator of the detection limit can be determined from 

published values, if issues about statistical independence are set aside.

The background count rate for the proportional counters was reported as 0.12 cpm in 1963 

[McLendon 1963, p. 42]. Counters were not used if the background was greater than 5 counts in 

30 minutes [Hamrick 1958, p. 9]. For two disks counted as described above, assuming a well-known 

background and alpha of 0.05, the LD is 0.13 cpm. For 1,400 mL/d urine output, a nominal 0.5 

cpm/dpm counting efficiency, and an average uranium recovery of 40% as reported by AEC [1959], 

the LD would be 46 dpm/d. With an increase in average uranium recovery to 73% [McRee et al. 1965, 

p. 32], LD would be reduced to 26 dpm/d. In reviewing historical gross alpha urinalysis data, Barber 

and Forrest [1995] reported a decision level of approximately 20 dpm/d. Average uranium recoveries 

between 1958 and 1965 have not been identified, and it is not clear that the statistical convention 

used by Barber and Forrest was identical to that used here. It is provisionally assumed that LD was 

47 dpm/d through 1964 and 25 dpm/d starting in 1965. However, given the limitations of the rate 

method of estimating daily urine volumes, uncertainty in the excretion volume is expected to add 

substantially to the uncertainty associated with the detection limit of a single measurement. Because 

recoveries were based on batch rather than individual measurements, uncertainties in recovery would 

add to the uncertainty of the detection limit of a single measurement.

Uranium in Urine by Alpha Spectrometry (October 1989 to Present)

The alpha spectrometry system employs large urine aliquots, chemical separation of uranium, 

sample-specific tracers to determine recovery, and long counting times on low-background detectors.

The resultant LD is approximately 0.15 dpm/d [Souleyrette et al. 2002, p. 19]. The LD varies with the 

uranium recovery of a particular sample, and is reported with the sample result.
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Tritium in Urine by Gas Counting

Throughout most of the plant’s history (the earliest recorded data were from 1957), tritium was 

counted by reacting urine with calcium to evolve hydrogen gas, which was counted using a vibrating 

reed electrometer. The detection limit for this method is 1 μCi/L (1.4 μCi/d) [Hursh 1958]. This MDA is 

far below the plant action level of 0.25 μCi/mL (350 μCi/d) [Patterson et al. 1957, p. 18)]. Results are 

reported in μCi per mL. 

Tritium in Urine by Liquid Scintillation

A few mL of distillate from a urine sample are placed in liquid scintillation fluid for counting [Hursh 

1958]. Assuming typical method efficiencies, the MDA is 2,000 dpm/d.

Plutonium in Urine by Gross Alpha Counting

The plutonium in urine procedure used before 1988 involved chemical separation from a 24-hour void 

or a simulated 24-hour sample before gross alpha counting. The detection limit has not been 

identified, but was certainly far below the corresponding limit for uranium in urine, which was based on 

20-mL aliquots. If the counting method was identical (same count time of 60 minutes), chemical 

recovery (40% to 73%), and counting efficiency (0.5 cpm/dpm) to the early uranium-counting method, 

the MDA for a 24-hour plutonium sample (less 40 mL) would range from about 0.7 to 1.3 dpm/sample

(inferred from uranium MDA). 

Plutonium in Urine by Alpha Spectrometry (1988 to Present)

The a priori LD for both 238Pu and 239,240Pu is 0.025 dpm/sample, or 0.025 dpm/d assuming a full 

24-hour void volume [Snapp 2003b, Table 4.2, p. 96]. Actual sample-specific MDAs are included in 

the analysis reports.

Other Actinides in Urine and Feces

Table 5-9 lists the analytical laboratory LD values used since about 1988.

Table 5-9. LD values.

a

Radionuclide

LD

(dpm/sample)

Am-241 0.05

Pu-238 0.025

Pu-239 0.025

Th-228 0.15

Th-232 0.07

Np-237 0.1b

a. Source: Snapp [2003b, p. 111]. 

b. Not applicable for fecal.

Uranium in Feces

Fecal sampling was started in 1999 when the presence of a less soluble component was identified as 

a result of the stand down [Veinot 2003, p. 29]. In 2002, of 2,800 participants in the bioassay program, 

700 Y-12 workers participated in the fecal-sampling program [Veinot and Souleyrette 2003, p. 12]. 

Participation in the fecal-sampling program is based on the potential for significant exposures to 

insoluble uranium (i.e., workers who are expected to have an internal exposure to type S uranium and 

there is a potential to exceed 100 mrem) [Veinot and Souleyrette 2003, p. 20]. The Analytical 

Chemistry division cites the value of 0.15 dpm/sample as a representative MDA for fecal analysis 

[Souleyrette et al. 2002, p. 19]. Samples usually consist of a single void. The results in the claim files 

are assumed to be in units of activity per sample unless otherwise indicated [Veinot and Souleyrette 

2003, p. 12]. 
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5.3.4 In Vitro Bioassay Methods for Individual Radionuclides

The following sections discuss the in vitro methods for specific radionuclides in urine. (Section 5.3.5

discusses analysis of fecal samples.)

5.3.4.1 Uranium

Overview

The in vitro bioassay program for uranium at Y-12 can be divided into four eras:

1. From 1943 to 1947, limited monitoring occurred, for which the data cannot be retrieved 

[NIOSH 2011]. 

2. From 1948 to 1950, fluorometric analyses of urine and blood occurred as part of general 

medical surveillance to prevent kidney damage from exposure to soluble uranium compounds.

3. From 1950 to 1989, collection of routine samples based on uranium exposure potential was 

initiated. During this third phase, fluorometric analyses were performed on samples submitted 

by workers in NU and DU areas. Electrodeposition of uranium, followed by gross alpha 

counting, was used for samples submitted by workers in EU areas. The primary goal for the 

EU analyses was to control lung doses from insoluble compounds under the assumption that 

such materials cleared from the lung with a 120-day half-life. During this lengthy era there 

were modifications in procedures, but the basic approach remained the same. Fecal sampling 

was used for some follow-up investigations.

4. After October 1989, routine 24-hour urine samples were collected. The uranium was extracted 

by chemical separation and ion exchange, and the extract was counted by alpha spectroscopy 

for an extended period (1,000 minutes per sample). Since 2000, fecal samples have been 

submitted routinely for individuals working with largely insoluble forms of uranium.

Uranium Analysis by Fluorometry, 1948 to 1989

Uranium processing began in the fall of 1943. Y-12 technical reports describe the use of fluorometry 

for uranium detection as early as 1944 [Van Wazer et al. 1944]. By 1946, fluorometric analyses for 

uranium in urine and blood were in general use to supplement clinical surveillance for soluble uranium 

exposures [Sterner and Riley 1946]. Fluorometric uranium urinalysis data from the Tennessee 

Eastman Corporation era (1943 to 1947) were discussed in a 1981 mortality study of plant personnel 

[Polednak and Frome 1981]. 

Early in the Union Carbide management of Y-12, the fluorometric method was reassessed. Extraction 

methods were used starting in 1949 and 1950 [Struxness 1949b, p. 9; Emerson et al. 1951, p. 16]. By 

the first half of 1952, a technique involving small (0.2-mL) aliquots of raw urine had been put in place 

[Struxness 1953, p. 31]. However, after this change in technique was instituted, the urinary uranium 

concentrations increased greatly. This raises the concern that the technique used before May 1952 

could have underestimated the urinary uranium concentrations.

Fluorometric urinalysis continued to be used for NU and DU until 1989, when alpha spectroscopy 

began to be used. Fluorometry yielded results in mass units (e.g., μg per L), but results were often 

converted to disintegrations per minute per day with the assumed specific activity of 1.55 dpm/μg for 

NU. The fluorometric technique had a sensitivity of about 5 ppb (5 μg/L or 7 μg/d) (e.g., Struxness

[1949a, p. 9]).
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Uranium Analysis by Electrodeposition and Alpha Counting, 1950 to 1989

The primary objective of this procedure was to prevent EU lung burdens in excess of the prevailing 

limit (0.017 μCi) corresponding to an ICRP Publication 2 [ICRP 1960] calculated lung dose of 

15 rem/yr (using a quality factor of 10). Samples from workers in EU areas were analyzed by 

electrodeposition of uranium onto silver discs, which were then counted for gross alpha activity in a 

proportional counter. Development of the gross alpha counting method was reported to be underway 

in early 1949 [Struxness 1949b, p. 9]. At first, samples were subjected to an acid digestion step before 

electrodeposition. In 1951, methodological problems leading to underestimates of a factor of 2.5 were 

noted and corrected [Ballenger 1952, p. 50]. The information available was not sufficient to determine 

if the records themselves were corrected. In the mid-1950s, the acid digestion step was discontinued 

altogether due to reported contamination problems, as well as for logistical considerations [AEC 1959, 

p. 66]. 

From the mid-1950s, two 20-mL aliquots of raw urine from each sample were placed in 

electrodeposition cells, which produced two disks per sample. Each disk was counted once on one 

proportional counter for 30 minutes and a second time on a different counter. The results were then 

averaged. The background count rate was reported as 0.12 cpm in 1963 [McLendon 1963, p. 42]. The 

expected net count rate for a person excreting at the action level of 70 dpm/d was 0.2 cpm, given a 

urine output of 1.4 L/d, a nominal 0.5-cpm/dpm counting efficiency, and a uranium recovery of 40%.

Because each disk was counted twice for 30 minutes each, the expected net count for the two counts 

on each disk under these conditions was 12 counts, over a background of 7.2 counts.

Recoveries were estimated on the basis of spiked samples that accompanied each counting run. The 

typical spike contained relatively little activity. Recovery rates from raw urine tended to be around 40% 

in the late 1950s [AEC 1959], but they improved over time. In 1965, the average recovery for 

electroplating was 73% [McRee et al. 1965, p. 32]. Method limitations are discussed in a number of 

reports, including Johnson et al. [1959] and UCC [1959]. Recovery rates were found to vary with pH, 

drying procedure, and uranium content, with rates decreasing for uranium concentrations below the 

plant action level. The precision of any one sample was acknowledged to be relatively low: 

Further, realizing the over-all lack of precision in any one sample result, we normally 

recommend restriction only on the basis of the 13-week or quarterly cumulative internal 

dose as indicated by from 2 to 13 samples [AEC 1959, p. 69].

The precision of an individual result in this type of procedure is poor and for this reason 

little significance is attached to individual results, particularly those below the “Plant 

Action Limit (PAL)” level [UCC 1966, p.4].

These limitations notwithstanding, the salient point is that when in vivo counting became routine in the 

early 1960s, very few additional workers were found to require restriction from uranium areas (e.g., 

Scott [1963, p. 9]). The in vivo monitoring frequency was determined by urinalysis results and ranged 

from monthly to once each 18 months [Scott 1963, p. 4]. Almost all the restrictions that did take place 

would have occurred on the basis of the urinalysis program alone. In those cases in which restriction 

was based on in vivo analysis alone, lung retention times were often observed to be considerably 

longer than the assumed 120-day lung half-life on which the urinalysis program was based. In other 

words, the problem was largely due to the inapplicability of the lung model for some materials and 

individuals, rather than a failure of the urinalysis program.

Uranium-233 by Gross Alpha Counting

In early 1962, Y-12 undertook the fabrication of metallic 233U from a nitrate solution (233U uranyl 

nitrate) in Building 9205 [West and Roberts 1962]. In recognition of the high specific activity of this 

material, and the potential for gamma radiation from the progeny of contaminant 232U, operations were 

doubly contained. Pilot runs were made with less hazardous materials to test the protective measures.
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As part of these precautions, eight workers submitted urine samples for analysis by electrodeposition 

and gross alpha counting as described above, with the same recoveries and detection efficiencies.

Ratios of 233U to 232U varied with operation and time. Table 5-10 lists ratios for various exposure 

scenarios. In addition to absorption types M and S, exposure to type F material is considered likely in 

this process. Uranium-232 has the larger DCF.

Table 5-10. 233U to 232U activity ratios.

a

Process Activity ratio

Reduction of UF4 1.3

Casting of metal-crucible skull 15–32

Machining final uranium component 160

a. Source: West and Roberts [1962]. 

Due to uncertainty in the process, assumptions that are favorable to claimants should be made about 

solubility and uranium activity ratios.

Uranium Analysis by Alpha Spectrometry, 1989 to Present

Up to 2 L of urine, along with a 232U tracer, are treated with nitric acid (HNO3) and hydrogen peroxide 

(H2O2). The sample is wet-ashed, and the uranium is coprecipitated with calcium oxalate (CaC2O2). 

After dissolving the precipitate in hydrochloric acid (HCl), the uranium is further separated by ion 

exchange chromatography. The uranium is eluted from the column with a solution of dilute HCl to 

which titanous chloride (TiCl3) has been added to reduce actinides that could be in an elevated 

oxidation state. The final fraction of the eluate is first treated with ascorbic acid to reduce the presence 

of any ferric iron and next with hydrofluoric acid. Next, the uranium isotopes are coprecipitated on 

neodymium fluoride (NdF3). The NdF3 is caught on a 0.1-mm filter, rinsed, dried, and mounted on a 

planchet for alpha spectrometry [Snapp 2003b, p. 68]. Typical recoveries are about 85%. Samples are 

typically counted for 16 hours on a passivated implanted planar silicon detector.

As investigation levels decreased over time, the contribution of natural background to worker uranium 

excretion assumed greater importance. Since 1989, Y-12 has corrected the measured uranium 

excretion values for background uranium interference in the performance of dose assessments. 

However, the uranium results reported in response to NIOSH requests for dose records have not 

been corrected for dietary uranium. Any result, other than a baseline sample, that is positive is 

assumed to be due to an occupational exposure for the purposes of dose reconstruction so no 

correction is necessary.

5.3.4.2 Plutonium

Plutonium in Urine by Gross Alpha Counting

Certain workers in the Special Testing Department were tested for plutonium intake by urinalysis.

Because large urine volumes were required, these workers submitted 24-hour samples through the 

use of take-home kits. The sampling frequency was monthly [Patterson et al. 1957, p. 43]. After two 

20-mL aliquots were removed for uranium analysis by electrodeposition, the balance of the sample 

was treated by chemical separation, precipitation, and evaporation onto a stainless-steel planchet for 

alpha counting. Health physics personnel converted the reported activity rate (disintegrations per

minute per 24-hr void) to mrem per day to the critical organ.

Uranium, Plutonium, and Americium by Alpha Spectrometry, October 1989 to Present

Up to 2 L of urine, along with 232U, 242Pu, and 243Am tracers, are treated with HNO3 and H2O2. The 

sample is wet-ashed, and the uranium, plutonium, and americium are coprecipitated with CaC2O2. 

After dissolving the precipitate in HCl, the uranium, plutonium, and americium are further separated by 

the use of two ion exchange columns and an additional oxalate precipitation. The uranium, plutonium, 

and americium in the final fractions are coprecipitated with NdF3. The NdF3 is caught on a 0.1-mm 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 30 of 81

 

filter, rinsed, dried, and mounted on a planchet for alpha spectrometry [Snapp 2003b, p. 69]. The 

same procedure is used for isotopic thorium.

5.3.4.3 Tritium

As of 1957, personnel engaged in processing materials with a potential for tritium contamination 

submitted three urine samples per month. Samples were submitted at the same stations used for 

uranium and plutonium samples. The samples were reacted to evolve hydrogen gas, which was then 

counted for beta activity using a vibrating reed electrometer. Results were reported in μCi per mL 

[Patterson et al. 1957, pp. 43-44]. 

Tritium is measured by liquid scintillation counting. An aliquot of the urine sample is distilled in a 

heating mantle. A 5-mL portion of the collected distillate is mixed with liquid scintillation cocktail. The 

beta activity of the tritium is then measured by liquid scintillation counting [Snapp 2003b, p. 71]. 

Exposure should be assumed to be due to tritiated water (HTO) vapor. 

5.3.4.4 Other Radionuclides

Alpha spectrometry procedures described in Section 3.2 of Snapp [2003b] have been in use since 

1989 and include methods for neptunium, plutonium, americium, and thorium. In addition, liquid 

scintillation methods for strontium and technetium were sometimes performed.

5.3.5 Fecal Sample Analysis

Fecal samples have long been used in follow-up investigations when urinalysis or in vivo 

measurements indicate the likelihood of a substantial intake. West and Scott [1966] describe such an 

investigation in the 1960s in which fecal, urine, and in vivo measurements were used jointly to 

investigate clearance. Fecal sampling was discontinued several times in the past, but was reinstated 

in 1998 due to changes in workplace exposure conditions. At first, fecal sampling was used for a 

limited number of workers in Building 9212 but was expanded to include Building 9215 in 1999 and 

other areas of the plant in 2000 and 2001 [Souleyrette et al. 2002, p. 10]. 

If a work area has been determined to have predominantly insoluble airborne uranium, as of 2002, 

workers with a moderate to high exposure potential (CEDE greater than 100 mrem) were to submit 

both urine and fecal samples. The relative elimination by urinary and fecal pathways is used to 

determine the solubility mixtures for each individual. For combined urine and fecal monitoring, the 

routine sampling interval is 53 days [Souleyrette et al. 2002, p. 18] with 24-hour fecal collections 

preferred. When 24-hour fecal samples are not available, the results can be normalized using the ratio 

of the Reference Man excretion rate of 135 g/d to the mass of the submitted sample. The Y-12 

practice was to normalize results routinely to Reference Man [Snapp 2003b, p. 79]. These normalized 

results or actual individual daily excretion results can be used to reconstruct dose.

5.4 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND REPORTING 

PRACTICES

5.4.1 Whole Body Counting

Whole body counting was not routinely practiced at Y-12 [Snapp 2003b]. The primary in vivo detection 

method was chest counting, as described below.
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5.4.2 Chest Counting

The Y-12 in vivo chest-counting facility was developed in the late 1950s and was put into routine use 

in 1961. The original facility is described in Cofield [1959a, 1959b]. A 9- by 4-in. sodium-drifted iodide 

[NaI(Tl)] crystal was placed over the chest of a subject reclining in a cot in a shielded room. Subjects 

showered, shampooed, and changed into a clean garment before counting. They were surveyed for 

remaining surface contamination before entering the counting room. The normal counting time was 

20 minutes.

Incremental improvements in the system were made over time. A second 9- by 4-in. NaI(Tl) crystal 

was placed under the subject’s back in 1963 or 1964. This position provided greater counting 

efficiency than the position over the chest. A second advantage was that skin contamination seldom 

occurred on the back. The ratio of the count rates from the two crystals could be used to identify 

surface contamination missed by the survey meter. In 1965, two 5-in. NaI(Tl) crystals were added, 

positioned at the sides under the arms. In 1985, initial development of a low-energy germanium 

counting system was begun [Souleyrette 2003, p. 12]. This system began routine operation in a new 

counting room in June 1992 [NIOSH 2021, p. 11].

5.4.2.1 Uranium

With the NaI scintillation system used for much of the plant’s history, the amount of uranium deposited 

in the lungs was inferred from the worker’s spectrum by use of a prediction equation. This included 

the subject’s weight and constants derived from the spectra of control subjects. Measured lung 

burdens were originally expressed in μg of 235U for EU and mg of 238U for NU or DU. The original limits 

of detection for 235U and 238U were reported as 130 μg [Cofield 1959b, pp. 8, 36] and 14.4 mg, 

respectively. Improvements in hardware and data reduction procedures led to lower detection limits, 

as indicated in Tables 5-11 and 5-12. Limits are given in terms of mass units and activity (nCi and 

dpm), with the published value in regular type and the converted value in italics. When known, the 

reporting convention used to define the detection limit is indicated in the tables. The alternative 

definitions listed by King and Barclay [1983] illustrate the importance of this factor.

Table 5-11. Reported lung counting detection limits for 235U.a

Year Detectors

U-235

(μg)

U-235

(nCi)

U-235

(dpm)

Reporting

convention Reference

1959 One 9-in. NaI 130 0.28b 620b Unknown Cofield 1959b

1963 Two 9-in. NaI 96 0.21b 470b Unknown Scott and West 1967

1965 Two 9-in., 2 5-in. NaI 72 0.16b 360b Unknown Scott and West 1967

1975 Two 9-in., 2 5-in. NaI 70 0.15b 330b Unknown Scott and West 1975

1983 Two 9-in., 2 5-in. NaI 62 0.13b 290b 1.96 σBKG King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 68.5 0.15b 330b LC King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 137 0.30b 670b LD King and Barclay 1983

1990 Two 9-in., 2 5-in. NaI 70 0.15b 330b (d) Barber and Forrest 1995

06/1992 High-purity 

germanium

46c 0.10 220c (e) Souleyrette et al. 2002; 

NIOSH 2021

a. LC = critical level; LD = detection level as defined by Currie [1968] with α = β = 0.05, paired observations.

b. Value converted from published value in μg. 

c. Value converted from published value in nCi. 

d. Type I error = 5%.

e. Individual MDA or LC supplied with analytical results.

Due to the uncertainty about actual methods for determining the lung counting detection limits, for 

dose reconstruction purposes and based on review of the Y-12 data, the 235U and 238U lung count 

detection thresholds are assumed to be 130 μg and 14.4 mg, respectively, through 1990. Beginning in 

1991, the critical level (LC) or MDA is supplied with the individual results. The sample-specific MDA is 

reported through 1999 and should be used for calculations in the Integrated Modules for Bioassay 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 32 of 81

 

Analysis (IMBA). The sample specific LC or LD is reported beginning in 2000. Multiply this value by 2 

for the MDA for IMBA calculations.

Table 5-12. Reported lung counting detection limits for 238U.a

Year Detectors

U-238

(mg)

U-238

(nCi)

U-238

(dpm)

Reporting

convention Reference

1965 Two 9-in., 2 5-in. NaI 6.5 2.2b 4,900b Unknown Scott and West 1975

1983 Two 9-in., 2 5-in. NaI 5 1.7b 3,800b 1.96 σBKG King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 7.2 2.4b 5,300b LC King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 14.4 4.8b 11,000b LD King and Barclay 1983

06/1992 High-purity 

germanium

4.5c 1.5 3,300c (d) Souleyrette et al. 2002; 

NIOSH 2021

a. LC = critical level as defined by Currie [1968] with α = β = 0.05, paired observations. 

b. Value converted from published value in mg. 

c. Value converted from published value in nCi. 

d. Individual MDA or critical level supplied with analytical results.

No description of the conversion count rate to activity to mass has been found. The analyte reported 

was based on the area in which the person worked. Individuals working in NU or DU areas had results 

reported as 238U, and workers in enriched areas had results reported as 235U. Assumptions that are 

favorable to claimants should be based on conversions of 93% enrichment for 235U (Oralloy VHEU) 

and natural isotopic abundances for 238U.

5.4.2.2 Other Actinides

Thorium

Thorium-232, 228Th, and 228Ra exposures can be assessed using 228Ac and 212Pb chest counting data

which are available starting in August of 1979. Before August of 1979 Y-12 thorium chest count 

measurements are presented in mass units (total mg Th), and such measurements cannot be used to 

assess thorium exposure with sufficient accuracy (see Section 5.1.3).

Because IMBA is not designed to be used with multiple-nuclide chains, it cannot be used for an exact 

assessment when progeny are used to determine the parent intake. However, an approximation can 

be made that can be used for many cases. ORAUT-OTIB-0076, Guiding Reconstruction of Intakes of 

Thorium Resulting from Nuclear Weapons Programs [ORAUT 2014], derives the basis and provides 

examples of how to evaluate thorium chest count data using IMBA. Assumptions about the process 

history of the material are made so the derived intakes are favorable to the claimant. In most cases, 

the 212Pb activity is a more reliable indicator of a thorium intake, so it is used for the thorium for dose 

reconstruction. Triple-separated thorium is assumed because it results in intakes that are favorable to 

the claimant. For thorium, absorption types M and S should be evaluated, and the type most favorable 

to claimants should be used to assign dose.

Because IMBA does not correctly calculate 232Th and 228Th doses, the following steps apply only to 

the intake assessment. Alternate methods must be used for the dose assessment. Given a 212Pb 

chest count result for a chronic intake longer than 1 year or a chest count more than 30 days after an 

acute intake:

1. Evaluate the chest burden using the 228Th biokinetic model in IMBA (i.e., use the 212Pb result to 

model the 228Th intake in IMBA). See Table 5-13 for detection limit information.

2. Multiply the intake rate obtained in step 1 by a factor of 1.1 and assign it as the intake rate of 

228Th.

3. Divide the 228Th intake rate by 0.19 to obtain the 232Th intake rate.
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If the conditions above (chronic intake longer than 1 year or chest count more than 30 days after an 

acute intake) are not met and an over- or underestimate cannot be performed, contact the Principal

Internal Dosimetry Scientist for a best estimate.

Reported 228Ac chest count results can be used to calculate a 228Ra intake. Determine the 228Ra intake 

by assuming a 228Ra chest count equal to the 228Ac measured value. Assume a type M material 

because that is the only material type to which radium is assigned.

The calculated 212Pb and 228Ac detection limits for 1979 through 1991 are in Table 5-13. The reported 

212Pb and 228Ac results should be adjusted by adding the adjustment factor according to the values in 

the table below before comparing it to 2 times the 212Pb or 228Ac detection limit (MDA = 2LD). The MDA 

for post-1994 228Ac lung counts (if a sample-specific value is not supplied) is 1.0 nCi

[Snapp 1995, p. 61]. 

Table 5-13. Calculated lung counting detection limits for 212Pb and 228Ac, 1979 to 

1991.

a,b

Year

Pb-212 

adjustmentc

Pb-212 DL 

(nCi)

Ac-228 

adjustmentc

Ac-228 DL 

(nCi)

1979 0.115 0.163 −0.012 0.232

1980 0.094 0.116 −0.039 0.098

1981 0.002 0.287 −0.028 0.145

1982 −0.034 0.209 −0.046 0.180

1983 −0.059 0.224 −0.096 0.187

1984 −0.052 0.227 −0.100 0.137

1985 −0.097 0.319 −0.112 0.218

1986 −0.004 0.192 −0.048 0.159

1987 0.017 0.319 −0.057 0.148

1988 0.212 0.223 −0.043 0.247

1989 0.251 0.237 −0.029 0.402

1990 0.266 0.234 −0.019 0.346

1991 0.127 0.294 −0.012 0.115

a. Source: ORAUT [2022]. 

b. DL = decision level; called “detection threshold” at Y-12.

c. Pb-adjustment must be added to the reported Pb-212 chest burden before comparing it to 2 

times the Pb DL. The same approach must be applied to the reported Ac results and the Ac

DL.

For the years 1992 to 1995, the MDA for 212Pb derived in DCAS-RPRT-008, Evaluation of the Pb-212 

Detection Limit for a Lung Count at the Y-12 Facility from 1992 through 1995 [NIOSH 2020], is 

presented as a function of the chest wall thickness (CWT) of the individual being counted. This 

equation, given at the top of page 17 in the report, simplifies to

MDA = 0.148 exp (0.321CWT) (5-2)

where the MDA is given in units of nCi and the CWT in cm. The plot in Figure 10 of the report is 

reproduced as Figure 5-1 below using this equation [ORAUT 2022]. 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 34 of 81

 

Figure 5-1. 212Pb MDAs at various CWTs for a 30-minute count [ORAUT 

2022]. 

From January to June 1992, lung counts for Y-12 personnel were performed at the Oak Ridge 

Gaseous Diffusion Plant (K-25) [NIOSH 2020]. The provided 232Th sample-specific MDAs should be 

used to calculate a missed dose.

Neptunium

At the time the in vivo system was put into routine service in 1961, the reported detection limit for 

237Np without progeny radiation was 2.7 nCi. For 237Np in full equilibrium with 233Pa, the detection limit 

was reported as 0.255 nCi [Cofield 1961, p.12]. 

Cesium

At Y-12, 137Cs results were quantified (along with 40K) so their effects on the spectra could be taken 

into account and removed from the occupational monitoring results. Because there is no evidence of 

137Cs in the occupational exposure source term at Y-12, no dose of record is associated with these 

results.

5.4.2.3 Other Radionuclides

At the time the in vivo system was put into routine service in 1961, the reported detection limit for 60Co

was 0.66 nCi. For 95Zr in transient equilibrium with 95Nb, the reported detection limit was 1 nCi [Cofield 

1961, p. 12]. Although bremsstrahlung counting could have been done for 99Tc, no information is 

available about the sensitivity of the technique.

5.5 INTERNAL DOSE RECONSTRUCTION GUIDANCE

1. Ensure that internal doses are assigned according to the SEC classes outlined in 

Section 5.1.3 of this document.
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2. It is preferred to use individual bioassay results to reconstruct internal dose if available in the 

claim records. Use the guidance in ORAUT-OTIB-0060 [ORAU 2018] to evaluate the bioassay 

results.

3. The urine and fecal sample-specific detection threshold provided in the Y-12 claim records 

starting in October 1989 should be used to calculate the MDA when using these results in 

IMBA. At Y-12, the decision level (DL) was called the “detection threshold.” Therefore, the 

MDA is DL × 2.

4. For any in vivo results (specifically, lung counts) before 2000, the LD that is reported on the 

“Bioassay Results History Report” is actually the MDA. Beginning in 2000, the LD that is 

reported for in vivo results is the LC of detection [Snapp 2014]. Use the provided samplespecific value as the MDA for results before 2000. Multiply the provided sample-specific value 

by 2 to obtain the MDA for results starting in 2000, where the MDA is the DL × 2). 

5. At Y-12, 137Cs results were quantified (along with 40K) so their effects on the spectra could be 

taken into account and removed from the occupational monitoring results. Because there is no 

evidence of 137Cs in the occupational exposure source term at Y-12, no dose of record is 

associated with these results.

6. Recycled uranium components should be added to uranium intakes starting in 1953. See 

Table 5-8 and the end of Section 5.2.4.1 for more information.

a. The ratios given for VHEU and Oralloy should be assigned unless there is information that 

justifies a different enrichment assumption. The values given for pCi/pCi total U in 

Table 5-8 can be used for the ratios. For a best estimate, divide the technetium by 3 and 

the plutonium, neptunium, and thorium by 10 after applying the ratios. The assumption of 

VHEU/Oralloy is the default assumption and can be used for the majority of cases as a 

claimant-favorable assumption.

b. For cases involving intakes other than VHEU/Oralloy, apply the RU components for LEU, 

NU & DU from Table 5-8. The values given for pCi/pCi total U can be used for the ratios. 

Divide ratios listed in Table 5-8 by 3 for 99Tc and by 10 for plutonium, neptunium, and 

thorium for best-estimate cases.

7. For unmonitored workers:

a. If co-exposure intakes are to be assigned, refer to the guidance in Section B.6.3 of this 

document. Co-exposure intakes can be assigned starting in 1948. The co-exposure 

intakes can be extended beyond the last year of available intakes, if necessary, using the 

1988 intakes for subsequent years. The maximum values should be used for most 

applications. Average values can be used for workers who had a lower potential for 

exposures to radioactive materials using the guidance in the ORAUT-OTIB-0014 [ORAUT 

2004] and based on the records provided for the worker or for best estimates.

b. Environmental intakes can be assigned starting in 1948 in accordance with ORAUT-TKBS0014-4 [ORAUT 2006] and ORAUT-OTIB-0014, Technical Information Bulletin:

Assignment of Environmental Internal Doses for Employees Not Exposed to Airborne 

Radionuclides in the Workplace [ORAUT 2004]. The environmental intakes can be 

extended beyond the last year of available intakes, if necessary, using the 2002 intakes for 

subsequent years.
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 36 of 81

 

8. Thorium guidance:

a. If the worker has in vitro bioassay results (specifically urine and/or fecal results) for 

thorium, then it can be inferred that the worker was exposed to thorium. These results 

would be shown on the “Bioassay Results History Report” [Snapp 2014]. Although some 

bioassays are labeled as dietary, treat all results greater than the MDA as indicating 

potential occupational intakes and evaluate the bioassay results using the guidance in 

ORAUT-OTIB-0060 [ORAUT 2018]. 

b. Before August 1979, available thorium-monitoring data are presented in mass units of total 

thorium. As discussed in the SEC-00251 evaluation report [NIOSH 2018], NIOSH does not 

have the necessary information to use these data for reconstructing internal dose due to 

228Th, 232Th, and 228Ra. For this reason, these data cannot be used for internal dose 

reconstruction for either thorium or thorium progeny.

c. For August 1979 to December 31, 1991, 232Th, 228Th, and 228Ra exposures can be 

assessed using 228Ac and 212Pb chest counting data in accordance with Section 5.4.2.2 to 

evaluate individual bioassay results. The 228Ac and 212Pb results are not always reported 

from the site. Therefore, if the records include a thorium lung count reported in mass units 

after July 1979, or if they include data from “Y-12 Lung Counts 1958-1991” [Y-12 lung, no 

date a] via a “Personnel Exposure” file in the claim files, it is easier to search the “Y-12 

Lung Counts Spreadsheet” [Y-12 lung, no date b] by the worker’s Social Security Number 

to determine if 228Ac and 212Pb results are available.

d. If the ANALYSIS TYPE code associated with a lung count result is listed as either 3, 6 or 7, 

then it can be inferred that the worker was exposed to thorium [Snapp 2022, p. 4].

e. For 1992 to 1995, use the equation in Section 5.4.2.2 to calculate the 212Pb MDA for the 

missed dose calculation. Additionally, for the period from January to June 1992, use the 

sample-specific 232Th MDAs to calculate a missed dose.

f. Beginning in 1992 and continuing through 2004, lung count results showing thorium 

(typically represented by 228Ac on the “Bioassay Results History Report”) should only be 

considered as screening results for thorium and are not necessarily indicative of potential 

for exposure to thorium. These screenings were performed on all workers who were lung 

counted regardless of work with thorium [Snapp 2014].

g. Additional information about work locations involving Y-12 Plant thorium operations and 

thorium work history are in ORAUT-TKBS-0014-2, Y-12 National Security Complex – Site 

Description [ORAUT 2007b]. 
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GLOSSARY

absorption type

Categories for materials according to their rates of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively insoluble in the 

respiratory tract (slow solubilization). See inhalation class. 

action level

A radiation or activity level that, when reached, results in a prescribed action. For example, 

high contact external dose rates can result in the use of extremity dosimetry; elevated airborne 

concentrations can result in the use of respiratory protection. Action levels are typically higher 

than minimum detectable levels. See minimum detectable level. 

activity median aerodynamic diameter (AMAD)

Diameter of a unit density sphere with the same terminal settling velocity in air as that of the 

aerosol particle whose activity is the median for the entire aerosol.

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.

bioassay

Measurement of amount or concentration of radionuclide material in the body (in vivo 

measurement) or in biological material excreted or removed from the body (in vitro 

measurement) and analyzed for purposes of estimating the quantity of radioactive material in 

the body. Also called radiobioassay. 

bremsstrahlung

Electromagnetic radiation released as a result of inelastic scattering of a moving charged 

particle within the nucleus of an atom. X-rays produced in a typical medical X-ray tube 

frequently originate from inelastic scattering of accelerated electrons in the anode material.

burnup

Depletion of fissionable material in nuclear fuel.

calutron

Accelerator that separates isotopes (e.g., 235U from 238U) according to their masses using 

strong magnetic fields. The name derives from California University cyclotron. 
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chronic exposure

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years). 

See acute exposure. 

class

See inhalation class and absorption type. 

class Q

Mixture of 10% class Y uranium and 90% modified class W uranium (the modification being an 

increase of the class W 50-day compartment to 120 days). An example of class Q material is 

that which occurs as a thin layer of uranium dioxide (UO2) on the surface of uranium metal as 

the metal interacts with oxygen in the air. See inhalation class and absorption type. 

class (type) Q system

Dosimetry system that uses an activity median aerodynamic diameter of 8 micrometers as a 

basis. See class Q. 

committed effective dose equivalent (CEDE)

Sum of the effective dose equivalents to various tissues or organs in the body each multiplied 

by the appropriate tissue weighting factor and committed for a 50-year period following an 

acute intake or the onset of chronic intake. It does not include contributions from external 

dose. See dose. 

contamination

Radioactive material in undesired locations including air, soil, buildings, animals, and persons.

cyclotron

Particle accelerator capable of large beam currents where the beam is injected into the center 

of a circular magnet. A fixed radiofrequency signal applied to two D-shaped electrodes 

accelerates the beam as it passes from one electrode to the other as the potential alternates. 

The radius of the beam increases as the energy increases.

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. As examples, 

spent (used) fuel elements, byproduct tails, residues from uranium isotope separation, and 

some weapons materials contain DU. DU can be blended with highly enriched uranium to 

make reactor fuel or used as a raw material to produce plutonium. Y-12 gives the isotopic 

distribution by mass as 0.001% 234U, 0.199% 235U, and 99.8% 238U.

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 rad, rep, or grays.

dose conversion factor (DCF)

(1) Multiplier for conversion of potential dose to the personal dose equivalent to the organ of 

interest (e.g., liver or colon). (2) In relation to radiography, ratio of dose equivalent in tissue or 

organ to entrance kerma in air at the surface of the person being radiographed.

enriched uranium (EU)

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 enriched to approximately 3.5% 235U; 
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weapons-grade uranium contains greater than 90% 235U. At Y-12, a common isotopic 

distribution by mass is 1% 234U, 93% 235U, and 6% 238U.

enrichment

Isotopic separation process that increases the percentage of a radionuclide in a given amount 

of material above natural levels. For uranium, enrichment increases the amount of 235U in 

relation to 238U. Along with the enriched uranium, this process results in uranium depleted in

235U. See depleted uranium and enriched uranium. 

exposure

In general, the act of being exposed to ionizing radiation. In this document, exposure does not 

refer to the radiological physics concept of charge liberated per unit mass of air. The product 

of exposure time to a radioactive aerosol and the average concentration during exposure, 

divided by the value of the derived air concentration for the radioactive material in question 

(expressed in derived air concentration-hours). See acute exposure and chronic exposure.

gaseous diffusion plant

Facility where uranium hexafluoride (UF6) gas is filtered to enrich the 235U and separate it from 

238U. The process requires enormous amounts of electric power and results in an increase in 

235U enrichment from 1% to about 3%.

half-life

Time in which half of a given quantity of a particular radionuclide disintegrates (decays) into 

another nuclear form. During one half-life, the number of atoms of a particular radionuclide 

decreases by one half. Each radionuclide has a unique half-life ranging from millionths of a 

second to billions of years.

highly enriched uranium (HEU)

Uranium enriched to at least 20% 235U for use as fissile material in nuclear weapons 

components and some reactor fuels. Also called high-enriched uranium. SRS lists the isotopic 

activity fractions as:

Isotope Activity fraction

234U 0.9806

235U 0.0194

238U 0.0000

The specific activity of highly enriched uranium with these fractions is 6,207 picocuries per 

microgram.

inhalation class

Former respiratory tract inhalation classification scheme developed by the International 

Council on Radiological Protection for inhaled material according to its rate of clearance from 

the pulmonary region of the lung. Materials were classified as D (days, half-life less than 

10 days), W (weeks, 10 to 100 days), or Y (years, more than 100 days). See absorption type, 

which superseded this concept.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds.

internal dose

Dose to the organ of interest from sources inside the body (e.g., plutonium or uranium) that 

are inhaled, ingested, absorbed, or injected through a wound).
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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

Measurements of radioactive material in the human body using instrumentation that detects 

radiation emitted from the radioactive material in the body.

low-enriched uranium (LEU)

Uranium enriched to less than 20% 235U by mass.

minimum detectable amount or activity (MDA)

The smallest amount of an analyte in a sample that can be detected with a probability β of 

nondetection while accepting a probability α of erroneously deciding that a positive (nonzero) 

quantity of analyte is present in an appropriate blank sample.

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements.

natural uranium (NU)

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcuries 

per gram). See uranium. 

nephrotoxicity

Poisoning of the kidney.

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.

quality factor

Principal modifying factor (which depends on the collision stopping power for charged 

particles) that is employed to derive dose equivalent from absorbed dose. The quality factor 

multiplied by the absorbed dose yields the dose equivalent. See dose and weighting factor. 

recycled uranium (RU)

Uranium first irradiated in a reactor, then recovered through chemical separation and 

purification. RU contains minor amounts of transuranic material (e.g., plutonium and 

neptunium) and fission products (e.g., technetium) or uranium products (e.g., 236U) after 

purification. SRS lists the isotopic activity fractions as:

Isotope Activity fraction

234U 0.8489

235U 0.0120

236U 0.1388

238U 0.0003
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The specific activity of recycled uranium with these fractions is 5,296 picocuries per 

microgram.

routine monitoring

Monitoring carried out at regular intervals during normal operations.

sensitivity

Minimum amount of contaminant (or dose) than can be repeatedly measured by (or calculated 

from) a particular analysis. See minimum detectable amount. 

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 depleted uranium, enriched uranium, highly enriched 

uranium, low-enriched uranium, recycled uranium, and natural uranium. 

very highly enriched uranium (VHEU)

Uranium enriched to 90% or more of 235U.

weighting factor

The ratio of the stochastic risk arising from tissue T to the total risk when the whole body is 

irradiated uniformly.
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ATTACHMENT A 

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Conventions Used in Internal Dose Reports........................................................................... 50

A.1.1 Y-12 Urinalysis Data, 1943 to September 1989 .......................................................... 50

A.1.2 Y-12 Urinalysis Data, October 1989 to Present .......................................................... 50

A.1.3 In Vivo Data................................................................................................................ 51

A.2 Uranium Solubility in the Lung................................................................................................ 53

A.3 In Vitro Detection Limits ......................................................................................................... 53

A.4 In Vivo Detection Limits.......................................................................................................... 54

LIST OF TABLES

TABLE TITLE PAGE

A-1 Uranium urinalysis detection limits ......................................................................................... 53

A-2 Other in vitro detection limits .................................................................................................. 54

A-3 Reported in vivo detection limits for other radionuclides......................................................... 54
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 50 of 81

ATTACHMENT A 

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS (continued)

A.1 CONVENTIONS USED IN INTERNAL DOSE REPORTS

The following is adapted from Souleyrette [2003]. 

A.1.1 Y-12 Urinalysis Data, 1943 to September 1989

Dept – Department number of employee at the time of monitoring.

Badge/Old Social Security Number – Worker identification number.

Void Date – Date sample was voided.

Program Code

1 = EU urinalysis by gross alpha

2 = NU or DU by fluorophotometry (used for EU before 1950)

6 = control sample

9 = (might have been used for a limited time to indicate experimental analyses methods)

Volume – volume of sample in mL. 

Time – Time between last void and sample void. Used to normalize urine result to 24-hour void based 

on individual employee daily excretion rate. (A zero time in conjunction with a large void volume and 

program code 9 probably indicates an experimental measurement technique.)

Uranium Result – The normalized urine result in disintegrations per minute per day.

Background Value – Background was stored for later results. The column Uranium Result (dpm) is 

background corrected already.

A.1.2 Y-12 Urinalysis Data, October 1989 to Present

Dept – Department number of worker at the time of monitoring.

Badge – Worker identification number.

Sample Number – Laboratory sample identifier.

Sample Date – Date sample was voided.

Sample Type

A = alpha spectroscopy

F = fecal

U = urine

S = start or stop date

(Note: For urinalysis, A is expected because all urinalyses have been done this way since October 

1989. The other entries are for database flexibility.)
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ATTACHMENT A 

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS (continued)

Reason – Reason for scheduling a sample:

1S = acute exposure

2S = follow-up sample to an acute exposure

3 = routine (chronic) exposure

2R = follow-up to routine (chronic) exposure

4 = background sample

5 = baseline sample

6 = random sample

8 = start date of exposure*

9 = stop date of exposure*

0 = acute exposure

*(These should appear only in separate records introducing a series related to a chronic intake study).

Retention Class – Assigned retention class: ICRP Publication 30 [ICRP 1979] classes D, W, and Y.

[Class Q indicates 90% class Super-W (120-day lung retention) and 10% class Y, 8 μm AMAD.]

Volume (mL) or Weight (g) – Sample volume (urine) or weight (fecal).

Total Uranium Result (dpm) – This is the normalized total uranium urine result in disintegrations per 

minute per day.

Type Assimilation

H = inhalation

I = ingestion

J = injection

Incident Date – Date of documented radiological exposure incident, if any.

Incident Time – Time of documented radiological exposure incident, if any.

A.1.3 In Vivo Data

Dept – Department number of worker at the time of monitoring.

Run Date – Date subject counted.

Run No. – Laboratory identifier of counting session.

Weight – Subject weight in pounds.

Chest Thickness – Subject CWT in inches (could be in tenths of inches for some years).

Surface Contamination Code – A six-digit code indicating if surface contamination was detected.

• If the first position is a 1, there was no surface contamination found, and the remaining five 

digits should be zeros.

• If the first digit is a 2 or 3, there was surface contamination found and the remaining five digits 

give the survey meter reading of the contamination.
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 52 of 81

ATTACHMENT A 

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS (continued)

Type Analysis – Code indicating the activity regions analyzed. There are two distinct sets of codes 

depending on the date of the count.

From October 1962 to 1971, the codes are:

0 = Background run

1 = Uranium

2 = New hires/control counts

3 = Thorium

4 = Neptunium

5 = Miscellaneous

6 = Second thorium count

Since 1972, the codes are:

1 = 235U 

3 = Thorium

4 = 238U and 235U 

6 = Thorium and 235U (first thorium count)

7 = 235U and thorium (second thorium count)

Material Type – This code provides additional details for cases in which Type Analysis = 0, 1, 2, or 3.

For counts after 1971, where Type Analysis is 0:

00 = Background run

01 = Sensitivity check

For Type Analysis of 1 (any year) or Type Analysis of 2 (after 1971):

01 = Unknown

02 = DU

03 = DU

07 = Normal uranium

13 = EU

19 = EU

For Type Analysis of 3 (any year):

00 = Thorium (first examination)

02 = Thorium (second or later examination)

U-235 (μg) – 235U activity detected in chest cavity. If the count is invalid, an “&” appears as the first 

character.

Thorium (mg) – 232Th activity calculated from the progeny isotopes detected in chest cavity.

U-238 (mg) – 234Th activity detected in chest cavity.

Tc-99 (μCi) – 99Tc activity detected in chest cavity. 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 53 of 81

ATTACHMENT A 

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS (continued)

Current Dept – Additional instance of worker department number used historically for reporting 

purposes.

A.2 URANIUM SOLUBILITY IN THE LUNG

For a workplace as varied as Y-12, no single solubility type or particle size distribution applies to all 

workers. Further, accurate assignment of the uranium lung absorption type to a given bioassay result 

is virtually impossible because of uncertainties about chemical form and limitations of the personnel 

tracking system [Barber and Forrest 1995, p. 10]. Exposure to type M material from 1948 to June 

1998 appears to be the most likely absorption type. After June 1998, exposure to absorption type S 

material is more likely. However, the absorption type can be based on the monitoring data or 

favorable to claimant assumptions. Further, several solubility studies were conducted between 1992 

and 1999 [Veinot 2003, Section 6]. Based on this study, both depleted and EU metals and oxides 

were found to act as type S material. A more conclusive assessment would require examining 

activities in each building. A number of buildings seemed to contain type S material. An important 

observation in this study was that certain areas in buildings could contain varying solubilities of 

uranium. One area might contain a very insoluble form, while an adjacent area might contain a more 

soluble form. Barber and Forrest [1995] showed the need to better characterize the type of material 

found at Y-12. As a result of this study, RWPs were modified for areas where insoluble forms 

dominated. Areas with mixtures of soluble and insoluble forms of uranium were designated for 

additional sampling for better characterization. If a worker entered an area that was determined to 

contain type S material, fecal and urine bioassays were implemented. If the worker entered an area 

that contained a more soluble form of uranium, urine bioassay only was used to estimate intakes 

[Veinot 2003, p. 29]. Thus, the presence of fecal sample results in an individual’s monitoring records 

is a strong indicator that the worker was exposed to insoluble uranium compounds.

A.3 IN VITRO DETECTION LIMITS

Tables A-1 and A-2 summarize information developed in Section 5.3. The tabulated values for 

urinalysis results represent laboratory LDs and do not include uncertainties introduced by sample 

collection or conversion from submitted volumes to daily void volumes. As noted in Section 5.3, LD

values for some historical techniques remain to be identified and will be reported in later revisions of 

this TBD as they become available.

Table A-1. Uranium urinalysis detection limits. 

Method Period

Detection limit 

(mass)

Detection limit 

(activity) (dpm/d) Convention 

Fluorometry 1950–09/1989 7 μg/da 11 Not applicable

Gross alpha 1950–1964 Not applicable 46b LD

Gross alpha 1965–09/1989 Not applicable 25c LD

Alpha spectrometry 10/1989–present Not applicable 0.15d LD

a. Based on 5 ppb sensitivity determination and a nominal 1.4 L/d excretion rate [Struxness 1949a, p. 9].

b. Estimated based on assuming a well-known background and alpha of 0.05, the LD is 0.13 cpm. For 1,400 mL/d urine 

output, a nominal 0.5 cpm/dpm counting efficiency, and an average uranium recovery of 40% as reported by AEC 

[1959], the LD would be 46 dpm/d.

c. Estimated based on assuming a well-known background and alpha of 0.05, the LD is 0.13 cpm. For 1,400 mL/d urine 

output, a nominal 0.5 cpm/dpm counting efficiency and an increase in average uranium recovery to 73% [McRee et al. 

1965, p. 32] the LD would be 25 dpm/d. 

d. Souleyrette et al. [2002, p. 19].
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 54 of 81

ATTACHMENT A 

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS (continued)

Table A-2. Other in vitro detection limits. 

Analyte/solubility type Method Period Detection limit (activity)

Tritium (HTO) vapor Liquid scintillation 10/1988–present 2,000 dpm/da

Isotopic plutonium/M, S Alpha spectrometry 10/1989–present 0.025 dpm/sampleb

Am-241/M Alpha spectrometry 10/1989–present 0.050 dpm/sampleb

Th-228/M, S Alpha spectrometry 10/1989–present 0.150 dpm/sampleb

Th-232/M, S Alpha spectrometry 10/1989–present 0.070 dpm/sampleb

Np-237/M Alpha spectrometry 10/1989–present 0.100 dpm/sampleb

a. Estimate assuming typical method efficiencies. 

b. Source: Snapp [2003b, p. 111, Table 5.3]. 

A.4 IN VIVO DETECTION LIMITS

From the review of Y-12 documents and conversations with present and former Y-12 staff members, 

the vast majority of site operations from 1943 to the present involved uranium in a variety of chemical 

forms and degrees of enrichment. Therefore, the primary internal radiation exposure to Y-12 workers 

was from uranium. However, the internal dosimetry program has included limited monitoring for 137Cs, 

99Tc, thorium, plutonium, 228Ac, and tritium, among other radionuclides; there are difficulties in 

interpreting some of the measurement data that cannot be resolved satisfactorily at this time. At Y-12, 

137Cs results were quantified (along with 40K) so their effects on the spectra could be taken into 

account and removed from the occupational monitoring results. Because there is no evidence of 137Cs

in the occupational exposure source term at Y-12, no dose of record is associated with these results.

Table A-3 summarizes information developed in Section 5.4. Detection limits for some historical 

techniques remain to be identified, and will be reported in later revisions of this TBD as they become 

available.

Table A-3. Reported in vivo detection limits for other radionuclides. 

Radionuclide and 

solubility type

Reported limit 

(nCi)

Reported limit 

(dpm)

Ac-228 1.0a 2,220

Np-237/M 2.7b 5,900

Np-237/M; Pa-233/M, S 0.255b 560

Co-60/M, S 0.66b 1,500

Zr-95/M, S; Nb-95/M, S 1.0b 2,200

a. Source: Snapp [1995, p. 61], post-1994 lung counts if sample specific value is not 

available. For lung counts from 1979 to 1991, see Table 5-13.

b. Source: Cofield [1961, p. 12]. 

Although internal monitoring has existed from the earliest days, the data from these measurements 

are not available for years before 1950. Guidance for the period from 1948 to 1950 is provided.

For dose reconstruction purposes and based on review of Y-12 data, the 235U and 238U lung count 

detection thresholds are assumed to be 130 μg and 14.4 mg, respectively, through 1990. Beginning in 

1991, the LC or MDA is supplied with the individual results.

For dose reconstruction due to potential thorium intakes, if the monitoring records contain evidence 

that the worker received in vivo chest counts for thorium before 1997, it should be assumed that the 

worker had the potential for occupational thorium intakes. Beginning in May 1961, Y-12 Plant policy 

indicated that workers were to be counted at the in vivo facility before working with thorium to obtain a 

baseline or background spectrum. Thereafter, the routine measurements were performed at 

approximately 6-month intervals as long as the worker was involved with thorium processing. In 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 55 of 81

ATTACHMENT A 

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS (continued)

addition, if an incident occurred or there was reason to believe that an employee received an internal 

thorium exposure, the affected individual was to be scheduled for an immediate in vivo measurement 

[BWXT Y-12 2005; West 1961; McLendon 1963]. 

After the initial transuranic hazard assessment performed in the first quarter of 1997 (which included 

thorium), programmatic steps were implemented to infer when further bioassay sampling was 

indicated to determine if additional analyses were required to determine CEDE from intakes of the 

associated transuranic radionuclides. This approach was based on uranium bioassay results. Thus, if 

the monitoring records indicate increased surveillance for thorium (or other transuranic materials), it is 

reasonable to conclude that the worker might have had the potential to receive intakes in excess of 

the rates inferred based on the Table 5-7 ratios. In such cases, missed or measured intakes should 

be applied based on the available bioassay data.

Oftentimes 228Ac was reported on in vivo chest counts. In such cases after 1997, this should not be 

inferred to represent a potential for occupational intakes of thorium in excess of the associated 

transuranic ratio unless there are other indications that the worker was being monitored with greater 

scrutiny (more frequent in vivo counts, urine or fecal sample submissions, etc.).
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 56 of 81

 

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA

TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Purpose ................................................................................................................................. 58

B.2 Overview................................................................................................................................ 58

B.3 Data ....................................................................................................................................... 58

B.3.1 Selected Bioassay Data.............................................................................................. 58

B.3.2 Analysis...................................................................................................................... 58

B.4 Intake Modeling...................................................................................................................... 63

B.4.1 Assumptions............................................................................................................... 63

B.4.2 Bioassay Data Fitting.................................................................................................. 64

B.5 Uranium Intake Rates by Material Type ................................................................................. 64

B.5.1 Type F........................................................................................................................ 64

B.5.2 Type M ....................................................................................................................... 65

B.5.3 Type S........................................................................................................................ 65

B.6 Assignment of Intakes and Doses .......................................................................................... 66

B.6.1 Intake Rate Summary................................................................................................. 66

B.6.2 Contribution from Contaminants in Recycled Uranium................................................ 67

B.6.3 Dose Assignment ....................................................................................................... 67

LIST OF TABLES

TABLE TITLE PAGE

B-1 Monthly uranium 24-hour urinary excretion rate analyses, 1952 to 1988................................ 59

B-2 Type F uranium intake periods and rates ............................................................................... 65

B-3 Type M uranium intake periods and rates .............................................................................. 65

B-4 Type S uranium intake periods and rates ............................................................................... 66

B-5 Type F uranium intake periods and rates ............................................................................... 66

B-6 Type M uranium intake periods and rates .............................................................................. 67

B-7 Type S uranium intake periods and rates ............................................................................... 67

LIST OF FIGURES

FIGURE TITLE PAGE

B-1 Assumed uranium intake, 01/01/1947 to 12/31/1969, 50th-percentile results, type F ............. 68

B-2 Assumed uranium intake, 01/01/1970 to 12/31/1988, 50th-percentile results, type F ............. 69

B-3 Assumed uranium intake, 01/01/1947 to 02/28/1978, 50th-percentile results, type M ............ 70

B-4 Assumed uranium intake, 01/01/1947 to 04/30/1952, 50th-percentile results, type M ............ 70

B-5 Assumed uranium intake, 08/01/1953 to 11/30/1953, 50th-percentile results, type M ............ 71

B-6 Assumed uranium intake, 11/01/1956 to 02/28/1959, 50th-percentile results, type M ............ 71
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 57 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

B-7 Assumed uranium intake, 06/01/1961 to 12/31/1962, 50th-percentile results, type M ............ 72

B-8 Assumed uranium intake, 10/01/1968 to 12/31/1972, 50th-percentile results, type M ............ 72

B-9 Assumed uranium intake, 03/01/1978 to 09/30/1984, 50th-percentile results, type M ............ 73

B-10 Assumed uranium intake, 10/01/1984 to 12/31/1988, 50th-percentile results, type M ............ 73

B-11 Predicted uranium urinary excretion from eight independent intakes, 50th percentile, 

1952 to 1988..........................................................................................................................74

B-12 Predicted uranium urinary excretion from eight independent intakes, 84th percentile, 

1952 to 1988..........................................................................................................................75

B-13 Assumed uranium intake, 01/01/1947 to 02/28/1978, 50th-percentile results, type S............. 76

B-14 Assumed uranium intake, 01/01/1947 to 04/30/1952, 50th-percentile results, type S............. 76

B-15 Assumed uranium intake, 08/01/1953 to 11/30/1953, 50th-percentile results, type S............. 77

B-16 Assumed uranium intake, 11/01/1956 to 02/28/1959, 50th-percentile results, type S............. 77

B-17 Assumed uranium intake, 06/01/1961 to 12/31/1962, 50th-percentile results, type S............. 78

B-18 Assumed uranium intake, 10/01/1968 to 12/31/1972, 50th-percentile results, type S............. 78

B-19 Assumed uranium intake, 03/01/1978 to 09/30/1984, 50th-percentile results, type S............. 79

B-20 Assumed uranium intake, 10/01/1984 to 12/31/1988, 50th-percentile results, type S............. 79

B-21 Predicted uranium urinary excretion from eight independent intakes, 50th percentile, 

1952 to 1988..........................................................................................................................80

B-22 Predicted uranium urinary excretion from eight independent intakes, 84th percentile, 

1952 to 1988..........................................................................................................................81
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 58 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

B.1 PURPOSE

There are instances of workers who, for a variety of reasons, (1) were not monitored for internal 

exposure during the course of their employment at a DOE facility or (2) whose records of such 

monitoring are incomplete or unavailable. In such cases, data from coworkers can be used to 

approximate an individual’s possible exposure. This TBD details the calculation and assignment of 

intakes based on co-exposure data from the Y-12 Plant.

B.2 OVERVIEW

ORAUT-OTIB-0019, Analysis of Coworker Bioassay Data for Internal Dose Assignment [ORAUT 

2005a] describes the general process for analysis of bioassay data for assignment of doses to 

individuals based on co-exposure results.

Bioassay results were obtained from the Oak Ridge Institute for Science and Education Center for 

Epidemiologic Research (CER) Dosimetry Database, which contains uranium urinalysis records from 

the Y-12 Plant for 1950 to 1988. The CER obtained this database from Y-12 to conduct an 

epidemiology study of site workers. The database results are in units of disintegrations per minute per 

day, although original urinalysis results were reported in terms of either mass or activity 

concentrations, depending on the measurement method. Conversion factors from the original results 

to activity per day are discussed in this TBD. A statistical analysis of these data was performed in 

accordance with ORAUT-OTIB-0019 [ORAUT 2005a]. The resultant values were input to the IMBA 

computer program, and a fit to the data was performed to obtain intake rates for assigning dose 

distributions.

B.3 DATA

B.3.1 Selected Bioassay Data

Data were extracted from “Y12 Urinalysis 1950-1988,” a version of the CER Dosimetry Database 

[ORAUT 2005b, 2022]. Sample dates were taken from the Date field and uranium urinalysis results 

were taken from the DPERM_INT field. The latter field contains uranium bioassay results in units of 

disintegrations per minute per day. Samples labeled Control were excluded from the analysis, as were 

those with a code of 9 in the UseFlag field; the documentation provided with the database indicated 

that the latter meant “do not use data as specified by [Y-12] HP.” The reasons for marking particular 

results are unknown. A review of the database revealed that nearly 20% (85,544 of 479,446) of the 

results were flagged as not to be used. Of those, 82,892 (97%) had results less than or equal to zero.

The excluded results greater than zero were relatively uniformly distributed among all positive results, 

and accounted for less than 1% (2,652 of 299,967).

B.3.2 Analysis

Because of the large number of sample results, the data were analyzed by month. A lognormal 

distribution was assumed, and the 50th and 84th percentiles were calculated for each month using the 

method in ORAUT [2005a]. Because there were fewer results in 1950 and 1951, these years were 

analyzed on an annual basis rather than by month. However, the analysis results were not included in 

the intake modeling because the 50th- and 84th-percentile values were much smaller than those in 

later years and might underestimate the urinary uranium concentrations as discussed in Section 5.3.2. 

January through April 1952 were included in the modeling because the 50th- and 84th-percentile 

values were generally larger than subsequent values. Table B-1 lists the statistical analysis results.
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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Table B-1. Monthly uranium 24-hour urinary excretion rate analyses, 1952 to 1988. 

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

01/15/1952 21.16 64 696

02/14/1952 24.92 84.51 741

03/15/1952 23.86 88.16 918

04/15/1952 25.49 74.15 665

05/15/1952 21.15 51.31 1,103

06/15/1952 18.06 47.31 1,401

07/15/1952 14.13 41.9 1,236

08/15/1952 11.25 33.06 1,258

09/14/1952 13.7 35.97 1,590

10/14/1952 16.71 46.49 1,305

11/14/1952 16.68 46.41 1,133

12/14/1952 14.29 37.68 1,368

01/14/1953 12.61 37.15 1,123

02/13/1953 10.51 32.28 1,065

03/16/1953 9.18 26.4 1,380

04/15/1953 10.34 32.25 1,095

05/16/1953 10.43 33.25 1,079

06/15/1953 9.81 34.9 1,337

07/15/1953 11.13 40.08 830

08/15/1953 9.81 33.3 1,261

09/14/1953 19.53 63.39 1,099

10/15/1953 32.43 110.13 677

11/14/1953 26.65 88.11 1,752

12/15/1953 18.47 57.43 1,614

01/14/1954 19.45 84.37 1,655

02/13/1954 16.06 74.2 1,938

03/16/1954 19.5 74.74 2,233

04/15/1954 22.04 78.72 1,688

05/16/1954 22.85 77.01 1,774

06/15/1954 22.07 73.11 1,801

07/16/1954 22.06 83.01 1,303

08/15/1954 15.39 63.49 2,571

09/15/1954 8.47 28.08 1,064

10/15/1954 11.58 50.14 1,726

11/14/1954 22.02 91.14 118

12/15/1954 14.85 45.87 783

01/14/1955 7.36 25.15 591

02/14/1955 5.62 19.81 639

03/16/1955 13.71 39.83 909

04/16/1955 16.2 50.07 1,840

05/16/1955 13.27 37.94 2,006

06/15/1955 15.71 47.63 1,595

07/16/1955 13.1 36.57 1,573

08/15/1955 13.85 42.78 1,874

09/15/1955 17.1 55.19 1,592

10/15/1955 16.87 51.49 1,990

11/15/1955 12.99 39.56 1,725

12/15/1955 15.34 45.01 1,732

01/15/1956 14.5 47.71 2,121

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

02/14/1956 16.03 48.01 1,853

03/15/1956 14.02 42.59 1,902

04/15/1956 16.45 46.84 1,886

05/15/1956 15.73 47.57 1,980

06/15/1956 13.79 40.67 1,986

07/15/1956 13.82 41.39 2,068

08/15/1956 13.39 41.44 843

09/14/1956 14.11 38.28 1,637

10/14/1956 11.1 32.33 2,022

11/14/1956 4.72 15.14 1,835

12/14/1956 7.3 26.34 1,889

01/14/1957 9.42 28.04 2,413

02/13/1957 10.2 38.81 1,967

03/16/1957 11.04 40.78 1,951

04/15/1957 10.79 33.8 2,302

05/16/1957 10.48 42.11 1,975

06/15/1957 9.44 34.06 2,099

07/15/1957 8.46 28.63 2,471

08/15/1957 10.56 35.75 2,038

09/14/1957 7.51 24.91 2,552

10/15/1957 12.06 53.49 2,278

11/14/1957 13.03 44.83 2,387

12/15/1957 11.29 37.96 2,878

01/14/1958 8.36 30.87 2,708

02/13/1958 13.64 53.23 2,208

03/16/1958 16.24 59.24 2,219

04/15/1958 13.3 52.29 2,498

05/16/1958 12.71 56.35 2,396

06/15/1958 13.7 77.06 2,665

07/16/1958 17.44 79.66 2,332

08/15/1958 19.45 65.95 2,206

09/15/1958 19.19 86.52 2,669

10/15/1958 15.21 58.52 2,115

11/14/1958 21.01 73.93 2,258

12/15/1958 22.76 80.61 3,071

01/14/1959 20.74 77.72 2,949

02/14/1959 24.9 90.03 2,798

03/16/1959 18.71 60.78 2,424

04/16/1959 19.72 66.84 3,126

05/16/1959 17.44 56.57 2,725

06/15/1959 12.41 45.55 2,878

07/16/1959 16.69 57.44 2,520

08/15/1959 12.79 40.15 3,001

09/15/1959 13.66 43.44 2,580

10/15/1959 11.53 40.87 2,983

11/15/1959 14.86 53.07 3,317

12/15/1959 16.04 56.6 2,785

01/15/1960 13.87 43.69 3,395

02/14/1960 13.36 46 3,666
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 60 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

03/15/1960 9.31 34.17 2,662

04/15/1960 9.75 39.57 3,375

05/15/1960 11.19 38.09 3,796

06/15/1960 9.88 37.26 2,468

07/15/1960 10.51 37.64 2,330

08/15/1960 12.76 51.09 3,115

09/14/1960 9.45 32.2 2,392

10/14/1960 8.93 32.44 3,155

11/14/1960 9.85 35.5 2,486

12/14/1960 7.6 27.11 2,443

01/14/1961 7.53 32.63 2,571

02/13/1961 6.99 27.55 2,409

03/16/1961 9.36 37.96 2,397

04/15/1961 8.94 32.38 754

05/16/1961 7.46 26.83 1,556

06/15/1961 9.17 32.46 532

07/15/1961 11.04 38.61 1,035

08/15/1961 5.8 23.46 1,137

09/14/1961 12.96 54.55 621

10/15/1961 8.55 29.21 1,216

11/14/1961 7.46 28.68 1,222

12/15/1961 16.66 64.1 555

01/14/1962 11.93 48.51 1,303

02/13/1962 14.1 55.59 1,191

03/16/1962 22.09 81.26 830

04/15/1962 16.35 63.01 1,431

05/16/1962 10.69 39.93 1,462

06/15/1962 19.41 71.41 1,045

07/16/1962 13.57 52.64 1,528

08/15/1962 12.66 58.77 1,420

09/15/1962 19.24 72.07 1,049

10/15/1962 15.98 53.81 1,524

11/14/1962 16.91 65.36 1,470

12/15/1962 22.87 97.5 1,134

01/14/1963 16.8 64.65 1,222

02/14/1963 14.35 65.28 1,307

03/16/1963 17.58 70.17 1,203

04/16/1963 13.52 46.46 1,481

05/16/1963 12.56 44 1,661

06/15/1963 14.88 56 1,038

07/16/1963 13.4 36.41 1,321

08/15/1963 13.29 43.99 902

09/15/1963 13.48 44.18 1,247

10/15/1963 11.65 38.88 576

11/15/1963 9.5 26.83 1,002

12/15/1963 7.73 26.04 1,562

01/15/1964 8 28.09 1,390

02/14/1964 7.96 24.14 1,579

03/15/1964 10.27 32.95 1,172

04/15/1964 4.6 15.35 902

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

05/15/1964 3.58 10.53 1,203

06/15/1964 5.7 21.01 599

07/15/1964 5.66 18.62 903

08/15/1964 4.65 14.2 1,088

09/14/1964 4.62 12.61 477

10/14/1964 3.43 9.95 667

11/14/1964 3.26 10.78 1,112

12/14/1964 7.72 21.81 291

01/14/1965 4.7 14.91 545

02/13/1965 3.99 12.84 646

03/16/1965 6.81 23.26 404

04/15/1965 4.59 21.09 312

05/16/1965 6.7 23.6 571

06/15/1965 9.45 31.48 329

07/15/1965 5.64 18.05 364

08/15/1965 6.41 23.41 609

09/14/1965 2.74 11.57 376

10/15/1965 4.91 18.07 416

11/14/1965 3.44 12.05 571

12/15/1965 9.85 37.35 397

01/14/1966 4.44 19.14 469

02/13/1966 4.41 24.7 668

03/16/1966 10.17 42.75 391

04/15/1966 10.75 48.01 441

05/16/1966 5.04 16.99 739

06/15/1966 6.32 26.44 415

07/16/1966 8.7 39.61 359

08/15/1966 5.01 18.31 624

09/15/1966 8.86 40.76 431

10/15/1966 15.88 71.36 392

11/14/1966 8.68 32.4 539

12/15/1966 8.98 40.16 567

01/14/1967 6.79 33.34 339

02/14/1967 5.8 25.69 710

03/16/1967 8.35 49.25 426

04/16/1967 10.81 56.82 364

05/16/1967 3.25 18.88 593

06/15/1967 6.51 25.46 224

07/16/1967 5.49 35.82 390

08/15/1967 7.3 40.93 585

09/15/1967 4.42 19.34 401

10/15/1967 5.21 20.92 412

11/15/1967 5.01 20.71 514

12/15/1967 6.06 33.53 405

01/15/1968 6.5 28.54 393

02/14/1968 10.15 41.09 573

03/15/1968 8.1 39.56 476

04/15/1968 7.07 43.78 411

05/15/1968 11.07 53.86 458

06/15/1968 7.05 29.75 574
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 61 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

07/15/1968 5.2 23.25 395

08/15/1968 5.86 23.48 533

09/14/1968 5.48 26.46 421

10/14/1968 3.86 16.16 391

11/14/1968 5.9 23.37 516

12/14/1968 9.71 32.61 533

01/14/1969 8.43 41.12 220

02/13/1969 5.02 20.69 731

03/16/1969 5.65 27.79 655

04/15/1969 7.61 36.08 477

05/16/1969 5.22 24.92 593

06/15/1969 8.26 30.14 546

07/15/1969 7.37 29.64 375

08/15/1969 6.62 28.37 630

09/14/1969 4.7 17.88 475

10/15/1969 6.6 25.36 514

11/14/1969 7.06 22.98 541

12/15/1969 7.71 30.81 401

01/14/1970 9.84 40.81 336

02/13/1970 8.33 28.27 559

03/16/1970 7.85 29.62 316

04/15/1970 6.99 27.21 378

05/16/1970 4.93 18.99 103

06/15/1970 11.17 44.73 137

07/16/1970 8.88 38.05 399

08/15/1970 8.76 42.72 401

09/15/1970 11.59 55.4 236

10/15/1970 9.81 34.23 500

11/14/1970 13.1 51.16 460

12/15/1970 17.14 67.22 275

01/14/1971 4.41 18.55 12

02/14/1971 11.3 35.27 523

03/16/1971 12.76 50.18 334

04/16/1971 9.52 39.99 458

05/16/1971 5.87 21.42 557

06/15/1971 10.37 41.16 263

07/16/1971 4.67 19.49 396

08/15/1971 10.74 47.49 557

09/15/1971 13.95 62.06 275

10/15/1971 8.25 32.76 463

11/15/1971 8.24 29.09 504

12/15/1971 11.67 35.91 283

01/15/1972 11.54 44.77 289

02/14/1972 12.2 43.92 227

03/15/1972 15.64 56.16 253

04/15/1972 12.09 47.06 246

05/15/1972 12.64 50.99 296

06/15/1972 13.68 34.9 155

07/15/1972 18.95 70.38 257

08/15/1972 20.33 67.87 314

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

09/14/1972 18.14 54.91 234

10/14/1972 15.48 64.45 225

11/14/1972 12 42.03 369

12/14/1972 20.42 61.5 164

01/14/1973 15.63 41.34 347

02/13/1973 15.79 57.05 271

03/16/1973 15.71 49.06 325

04/15/1973 16.2 48.06 268

05/16/1973 14.46 47.02 366

06/15/1973 15.04 56.06 248

07/15/1973 14.38 43.07 326

08/15/1973 16.5 52.5 296

09/14/1973 12.97 56.63 509

10/15/1973 12.31 40.96 485

11/14/1973 10.62 29.86 486

12/15/1973 14.2 46.52 393

01/14/1974 12.16 38.96 418

02/13/1974 11.16 34.34 378

03/16/1974 20.33 56.38 287

04/15/1974 12.99 40.27 325

05/16/1974 9.14 22.89 405

06/15/1974 9.75 29.71 222

07/16/1974 7.54 20.57 270

08/15/1974 8.44 24.96 330

09/15/1974 8.77 24.28 264

10/15/1974 14.19 137.02 298

11/14/1974 9.16 53.77 368

12/15/1974 8.11 35.07 250

01/14/1975 8.93 37.19 252

02/14/1975 11.49 44.05 360

03/16/1975 16.92 57.92 291

04/16/1975 10.37 46.11 319

05/16/1975 11.74 39.05 297

06/15/1975 10.66 33.4 186

07/16/1975 11.04 38.53 188

08/15/1975 11.07 32.85 338

09/15/1975 8.9 31.58 171

10/15/1975 11.92 36.58 186

11/15/1975 9.18 32.82 298

12/15/1975 12.88 40.99 148

01/15/1976 9.97 30.51 196

02/14/1976 10.81 32.89 321

03/15/1976 9.19 28.35 225

04/15/1976 8.65 27.03 211

05/15/1976 8.3 24.08 347

06/15/1976 8.24 23.91 185

07/15/1976 3.82 12.47 194

08/15/1976 6.51 18.64 332

09/14/1976 5.39 14.99 193

10/14/1976 9.5 26.06 228
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 62 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

11/14/1976 13.24 40.32 297

12/14/1976 13.72 43.42 202

01/14/1977 13.97 44.66 169

02/13/1977 13.41 41.7 268

03/16/1977 12.91 40.62 245

04/15/1977 8.86 24.33 191

05/16/1977 8.39 22.81 302

06/15/1977 9.62 31.33 180

07/15/1977 7.46 28.02 190

08/15/1977 9.29 25.59 342

09/14/1977 11.22 54.94 207

10/15/1977 10.86 37.43 244

11/14/1977 10.75 47.78 341

12/15/1977 9.31 35.7 151

01/14/1978 9.28 35.13 184

02/13/1978 9.82 29.72 293

03/16/1978 4.34 13.13 223

04/15/1978 8.07 26.41 196

05/16/1978 5.53 18.15 291

06/15/1978 4.73 14.79 164

07/16/1978 6.58 20.43 185

08/15/1978 5.1 24.98 362

09/15/1978 7.02 22.71 198

10/15/1978 7.51 22.85 205

11/14/1978 6.81 24.14 390

12/15/1978 4.93 15.71 155

01/14/1979 5.48 22.99 180

02/14/1979 5.73 24.87 398

03/16/1979 7.5 55.07 219

04/16/1979 4.12 20.11 213

05/16/1979 5.45 26.64 357

06/15/1979 6.54 24.51 198

07/16/1979 6.15 26.52 205

08/15/1979 2.81 8.47 407

09/15/1979 7.52 23.8 180

10/15/1979 6.69 21.68 205

11/15/1979 5.68 16.34 360

12/15/1979 10.15 28.45 219

01/15/1980 5.43 19.51 238

02/14/1980 5.39 18.48 462

03/15/1980 9.96 31.5 206

04/15/1980 7.03 28.92 273

05/15/1980 6.08 23.84 473

06/15/1980 7.52 24.86 215

07/15/1980 4.43 13.44 231

08/15/1980 4.33 14.91 488

09/14/1980 5.59 16.59 237

10/14/1980 8.8 27.7 253

11/14/1980 3.7 14.66 468

12/14/1980 6.65 26.31 255

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

01/14/1981 6.78 27.85 305

02/13/1981 5.54 25.15 462

03/16/1981 6.2 20.98 325

04/15/1981 4.81 19.11 269

05/16/1981 3.66 15.89 535

06/15/1981 5.97 21.83 251

07/15/1981 6.47 31.68 26

08/15/1981 4.89 16.79 575

09/14/1981 6.13 22.23 315

10/15/1981 5.14 21.72 314

11/14/1981 4.35 18.28 605

12/15/1981 5.73 29.53 294

01/14/1982 4.53 18.34 351

02/13/1982 4.13 15.44 540

03/16/1982 5.01 18.71 482

04/15/1982 5.72 32.11 323

05/16/1982 6.89 24.15 584

06/15/1982 5.47 20.5 366

07/16/1982 4.66 17.11 320

08/15/1982 6.48 31.75 581

09/15/1982 2.92 10.86 351

10/15/1982 6.66 25.77 332

11/14/1982 3.24 12.97 637

12/15/1982 2.94 13.34 308

01/14/1983 3.67 15.42 348

02/14/1983 5.93 23.87 634

03/16/1983 5.51 26.4 316

04/16/1983 4.51 17.99 296

05/16/1983 4.83 19.89 633

06/15/1983 2.99 12.46 288

07/16/1983 2.74 9.92 271

08/15/1983 5.17 19.39 690

10/15/1983 3.36 16.08 279

11/15/1983 2.69 14.1 706

12/15/1983 2.08 8.98 325

01/15/1984 2.72 11 326

02/14/1984 2.59 14.01 653

03/15/1984 4.2 20.09 329

04/15/1984 4.14 21.85 360

05/15/1984 3.51 15.94 637

06/15/1984 3.42 15.07 335

07/15/1984 3.49 15.65 352

08/15/1984 2.96 17.55 697

09/14/1984 3.97 20.81 356

10/14/1984 6.23 26.17 367

11/14/1984 2.68 8.13 644

12/14/1984 4.62 20.08 348

01/14/1985 2.93 12 397

02/13/1985 1.93 8.23 605

03/16/1985 4.45 31.7 425
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 63 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

04/15/1985 1.67 8.6 391

05/16/1985 1.66 8.97 625

06/14/1985 0.76 5.03 318

07/15/1985 1.91 8.85 383

08/14/1985 1.97 7.66 569

09/14/1985 2.76 11.77 369

10/14/1985 1.82 6.05 419

11/13/1985 1.53 6.04 544

12/14/1985 2.71 11.36 379

01/13/1986 2.79 11.56 472

02/13/1986 1.1 5.61 593

03/14/1986 2.41 13.98 388

04/15/1986 1.35 5.24 497

05/15/1986 1.12 5.13 555

06/15/1986 1.63 7.52 401

07/15/1986 2.42 11.67 408

08/14/1986 2.54 10.13 572

09/14/1986 1.79 8.35 352

10/14/1986 0.95 4.22 550

11/14/1986 2.6 11.31 626

12/14/1986 1.17 6.57 323

01/14/1987 0.8 8.03 437

Effective 

sample 

date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

02/13/1987 9.62 57.02 555

03/15/1987 5.04 18.9 345

04/15/1987 2.5 13.62 404

05/15/1987 2.43 11.68 525

06/15/1987 2.2 19.91 328

07/15/1987 8.16 46.96 34

08/15/1987 13.59 70.44 24

09/14/1987 4.01 19.3 54

10/15/1987 0.44 2.13 474

11/14/1987 1.78 14.61 494

12/14/1987 2.13 11.13 327

01/14/1988 2.19 14.84 338

02/13/1988 2.37 10.64 550

03/15/1988 4.62 24.43 371

04/14/1988 4.03 26.36 345

05/15/1988 1.52 7.09 606

06/14/1988 2.05 13.32 363

07/14/1988 4.03 19.34 416

08/14/1988 2.96 17.83 563

09/14/1988 2.38 14.04 432

10/14/1988 0.75 3.7 434

11/13/1988 2.18 10.51 103

12/14/1988 3.69 13.94 152

B.4 INTAKE MODELING

B.4.1 Assumptions

All results were assumed to be representative of a full day (24 hours) of urinary excretion. Each result 

in the intake calculation was assumed to have a normal distribution, and a uniform absolute error of 1 

was applied to all results to weight all results equally. Because of the nature of work at Y-12, a chronic 

exposure pattern best approximates the true exposure conditions for most workers with a potential for 

intakes. Intakes were assumed to be by inhalation using a default breathing rate of 1.2 m3

/hr and a 

5-μm AMAD particle size distribution.

The database reported all results as “uranium.” Because a variety of enrichments have been present 

at the Y-12 site, 234U was assumed for the IMBA intake modeling. This does not affect the fitting of the 

data for intake determination (i.e., the same total intakes would be obtained for any assumed 

enrichment) because all uranium isotopes behave the same biokinetically and the isotopes that were 

considered in this analysis have long half-lives in relation to the assumed intake period. ICRP 

Publication 68 provides dose coefficients (also referred to as DCFs) for 234U that are 7% to 31% larger 

than those for 235U, 236U, and 238U [ICRP 1995b]. Because of the isotopic compositions of the source 

terms, use of the 234U DCF overestimates doses.

Although there are no bioassay results from before 1950 in the database, the first intake period was 

assumed to begin on January 1, 1947. Before 1947, the calutron was in operation. There are no 

bioassay measurements for the period, and conditions were likely quite different from those in later 

years, so this period was not included in the modeling. The ORAUT-TKBS-0014-2, Y-12 National 
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 64 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Security Complex – Site Description states, “Y-12 was shut down in December 1946 and employment 

was cut drastically,” in reference to the calutron and associated uranium isotope separation programs

[ORAUT 2007b]. Primary operations from 1947 to 1951 consisted of salvage, recovery, and recycling 

operations with uranium preparation and machining beginning in 1949. It was assumed, therefore, 

that exposure conditions beginning in 1947 would have been similar to those in the early 1950s.

B.4.2 Bioassay Data Fitting

The IMBA computer program was used to fit the bioassay results to a series of inhalation intakes 

[ORAUT 2022]. IMBA allows the input of 200 urine sample results, which is insufficient to include all 

Y-12 monthly results. However, this can be expanded to 400 results by using the user-defined

bioassay function and applying all the urine parameters to this function. Data from January 1952

through December 1988 were fit as a series of chronic intakes.

The initial intake assumptions were based on periods that coincided with major operations on the site.

Operations from 1947 to 1951 were very specific and, therefore, were modeled as one intake period.

However, the bioassay data had some distinct patterns, so the intake dates were adjusted to obtain a 

better approximation of the data. There appeared to be low-level chronic intakes of uranium 

throughout long periods, with briefer, larger intakes superimposed on them. To model this pattern, 

three long-term chronic exposures were assumed to cover 1947 through 1988. For the type M and S 

fits, five shorter chronic exposures were modeled on top of the early period to account for the 

intermittent rises in the urine results. Type F uranium was fit as a series of chronic exposures 

corresponding to the patterns in the urine results.

Because the uranium isotopes at Y-12 have very long half-lives and the material is retained in the 

body for long periods for solubility types M and S, excretion results are not independent. For example, 

an intake in the early 1950s could contribute to urinary excretion in the 1980s and later. To avoid 

potential underestimation of intakes for people who worked at Y-12 for relatively short periods, each 

intake was fit independently using only the bioassay results from the single intake period. This is likely 

to result in overestimation of intakes, particularly for assumed type S exposures that extend through 

multiple assumed intake periods. Type F intake periods were fit simultaneously, subject to the 

limitations of IMBA, which required that the intake modeling be split into two periods.

B.5 URANIUM INTAKE RATES BY MATERIAL TYPE

All the bioassay results were entered into IMBA and are displayed in the figures at the end of this 

attachment. However, as described in Section B.4.2, each period was fit separately for solubility types 

M and S and into two periods for type F, so only the results during the intake period were selected for 

use in fitting each period. Excluded results are shown in red in the figures.

B.5.1 Type F

Uranium urine results were fit assuming type F material. Figures B-1 and B-2 show the individual fits 

to the 50th-percentile values. The same intake periods were applied to the 84th-percentile values 

because the values followed a similar pattern; results of the individual fits are not shown here but they 

fit reasonably well. Table B-2 summarizes the intake periods and corresponding intake rates for the 

50th and 84th percentiles with their geometric standard deviations (GSDs).

The GSDs were determined by dividing the 84th-percentile intake rates by the 50th-percentile intake 

rates.
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 65 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Table B-2. Type F uranium intake periods and rates (dpm/d). 

Start Stop 50th percentile 84th percentile GSD

01/01/1947 05/31/1952 85.35 264.1 3.09

06/01/1952 08/31/1953 43.89 131.3 2.99

09/01/1953 10/31/1953 104.3 339.5 3.26

11/01/1953 12/31/1954 66.43 254.1 3.83

01/01/1955 12/31/1956 47.03 145.2 3.09

01/01/1957 12/31/1957 36.07 145.6 4.04

01/01/1958 12/31/1959 59.48 207.8 3.49

01/01/1960 12/31/1961 34.68 134 3.86

01/01/1962 12/31/1963 53.27 171.9 3.23

01/01/1964 12/31/1969 22.97 97.28 4.24

01/01/1970 12/31/1971 35.8 115.6 3.23

01/01/1972 12/31/1975 47.51 168.4 3.54

01/01/1976 12/31/1977 35.09 121.7 3.47

01/01/1978 12/31/1983 19.68 85.61 4.35

01/01/1984 12/31/1988 9.098 51.26 5.63

07/01/1987 08/31/1987 43.498 102.61 2.36

B.5.2 Type M

Uranium urine results were fit assuming type M material. Figures B-3 to B-10 show the individual fits 

to the 50th-percentile values. The same intake periods were applied to the 84th-percentile values 

because the values followed a similar pattern; results of the individual fits are not shown here but they 

fit reasonably well. Table B-3 summarizes the intake periods and corresponding intake rates for the 

50th and 84th percentiles with their GSDs.

Table B-3. Type M uranium intake periods and rates (dpm/d). 

Start Stop 50th percentile 84th percentile GSD

01/01/1947 02/28/1978 169.34 598.93 3.54

01/01/1947 04/30/1952 354.59 1,154.9 3.26

08/01/1953 11/30/1953 547.22 1,963.1 3.59

11/01/1956 02/28/1959 226.90 825.31 3.64

06/01/1961 12/31/1962 248.68 866.81 3.49

10/01/1968 12/31/1972 160.67 601.60 3.74

03/01/1978 09/30/1984 80.03 355.93 4.45

10/01/1984 12/31/1988 44.203 223.85 5.06

Figure B-11 shows a representation of the predicted 50th-percentile values from all intakes, and 

Figure B-12 shows the 84th-percentile predictions. These depict the expected excretion rates from an 

individual who was exposed for all the periods at the 50th- and 84th-percentile intake rates, 

respectively, in Table B-3. The GSDs were determined by dividing the 84th-percentile intake rates by 

the 50th-percentile intake rates.

B.5.3 Type S

The intake periods for the type M fits were applied to the type S material fits. Figures B-13 to B-20 

show the individual fits for the 50th-percentile values. Table B-4 summarizes the intake rates for the 

50th- and 84th-percentile values. The GSDs were determined as noted for the type M intake rates.
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 66 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Table B-4. Type S uranium intake periods and rates (dpm/d). 

Start Stop 50th percentile 84th percentile GSD

01/01/1947 02/28/1978 1,844.4 6,544.2 3.55

01/01/1947 04/30/1952 5,210.4 16,970 3.26

08/01/1953 11/30/1953 17,983 64,559 3.59

11/01/1956 02/28/1959 5,694.6 20,560 3.61

06/01/1961 12/31/1962 6,849.4 23,632 3.45

10/01/1968 12/31/1972 3,290.3 12,415 3.77

03/01/1978 09/30/1984 1,280.3 5,802.9 4.53

10/01/1984 12/31/1988 884.85 4,340.6 4.91

Figures B-21 and B-22 show the predicted 50th- and 84th-percentile excretion rates, respectively, 

from all type S intakes.

B.6 ASSIGNMENT OF INTAKES AND DOSES

B.6.1 Intake Rate Summary

Several intake periods overlapped, so they were combined to make distinct periods with one intake 

rate and associated GSD for each. For 1947 through 1977, similar magnitude GSDs were assigned 

the largest GSD of the similar group for simplicity. Tables B-5 through B-7 summarize the intake 

periods. These are equivalent to the intake periods in Tables B-2 through B-4 but provide a 

chronological layout of the changing intake rates over time, simplify the GSDs, and provide the 

95th-percentile intake rates.

Table B-5. Type F uranium intake periods and rates (dpm/d). 

Start Stop 50th percentile GSD 95th percentile

01/01/1947 05/31/1952 85.4 3.54 683

06/01/1952 08/31/1953 43.9 3.54 351

09/01/1953 10/31/1953 104 3.54 834

11/01/1953 12/31/1954 66.4 4.24 715

01/01/1955 12/31/1956 47.0 3.54 376

01/01/1957 12/31/1957 36.1 4.24 388

01/01/1958 12/31/1959 59.5 3.54 476

01/01/1960 12/31/1961 34.7 4.24 373

01/01/1962 12/31/1963 53.3 3.54 426

01/01/1964 12/31/1969 23.0 4.24 247

01/01/1970 12/31/1971 35.8 3.54 286

01/01/1972 12/31/1975 47.5 3.54 380

01/01/1976 12/31/1977 35.1 3.54 281

01/01/1978 12/31/1983 19.7 4.35 221

01/01/1984 06/30/1987 9.10 5.63 156

07/01/1987 08/31/1987 43.5 3.00 265

09/01/1987 12/31/1988 9.10 3.54 156
Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 67 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Table B-6. Type M uranium intake periods and rates (dpm/d). 

Start Stop 50th percentile GSD 95th percentile

01/01/1947 04/30/1952 524 3.74 4,588

05/01/1952 07/31/1953 169 3.74 1,483

08/01/1953 11/30/1953 717 3.74 6,275

12/01/1953 10/31/1956 169 3.74 1,483

11/01/1956 02/28/1959 396 3.74 3,470

03/01/1959 05/31/1961 169 3.74 1,483

06/01/1961 12/31/1962 418 3.74 3,661

01/01/1963 09/30/1968 169 3.74 1,483

10/01/1968 12/31/1972 330 3.74 2,890

01/01/1973 02/28/1978 169 3.74 1,483

03/01/1978 09/30/1984 80.0 4.45 933

10/01/1984 12/31/1988 44.2 5.06 636

Table B-7. Type S uranium intake periods and rates (dpm/d). 

Start Stop 50th percentile GSD 95th percentile

01/01/1947 04/30/1952 7,054 3.77 62,595

05/01/1952 07/31/1953 1,844 3.77 16,366

08/01/1953 11/30/1953 19,827 3.77 175,929

12/01/1953 10/31/1956 1,844 3.77 16,366

11/01/1956 02/28/1959 7,539 3.77 66,895

03/01/1959 05/31/1961 1,844 3.77 16,366

06/01/1961 12/31/1962 8,693 3.77 77,138

01/01/1963 09/30/1968 1,844 3.77 16,366

10/01/1968 12/31/1972 5,134 3.77 45,559

01/01/1973 02/28/1978 1,844 3.77 16,366

03/01/1978 09/30/1984 1,280 4.53 15,367

10/01/1984 12/31/1988 885 4.91 12,125

B.6.2 Contribution from Contaminants in Recycled Uranium

Throughout the DOE complex, spent fuel from fission reactors has been processed to recover 

uranium for recycling. Because the uranium streams at Y-12 could have contained RU, the dose from 

the added constituents, including plutonium, 237Np, and 99Tc, must be included. See Section 5.2.4 for 

information about intake values in relation to the uranium intake amounts.

B.6.3 Dose Assignment

Doses can be calculated using these intakes starting in 1948. The co-exposure intakes can be 

extended beyond the last year of available intakes, if necessary, using the intakes for 1988 for 

subsequent years. For most cases, individual doses are calculated from the 50th-percentile intake 

rates. Dose reconstructors should select the material type that is the most favorable to claimants.

Doses based on 50th-percentile intake rates are assigned as lognormal distribution with the 

applicable GSD. There are situations when the 95th percentile of the co-exposure distribution and a 

constant distribution are more appropriate than the 50th percentile and lognormal GSDs. For cases 

where the 50th-percentile intake rates are not appropriate, dose reconstructors should use the 

95th-percentile intake rates. The 95th-percentile intakes should be assigned as a constant rather than 

lognormal distribution.

RU contaminants, when appropriate for the period, also factor into this comparison.
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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

The lognormal distribution is selected in IREP, and the calculated dose is entered as Parameter 1 and 

the associated GSD as Parameter 2. The GSD is associated with the intake, so it is applied to all 

annual doses from the intake period.

Figure B-1. Assumed uranium intake, 01/01/1947 to 12/31/1969, 50th-percentile results, type F. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-2. Assumed uranium intake, 01/01/1970 to 12/31/1988, 50th-percentile results, type F. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-3. Assumed uranium intake, 01/01/1947 to 02/28/1978, 50th-percentile results, type M. 

Figure B-4. Assumed uranium intake, 01/01/1947 to 04/30/1952, 50th-percentile results, type M. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-5. Assumed uranium intake, 08/01/1953 to 11/30/1953, 50th-percentile results, type M. 

Figure B-6. Assumed uranium intake, 11/01/1956 to 02/28/1959, 50th-percentile results, type M. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-7. Assumed uranium intake, 06/01/1961 to 12/31/1962, 50th-percentile results, type M. 

Figure B-8. Assumed uranium intake, 10/01/1968 to 12/31/1972, 50th-percentile results, type M. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-9. Assumed uranium intake, 03/01/1978 to 09/30/1984, 50th-percentile results, type M. 

Figure B-10. Assumed uranium intake, 10/01/1984 to 12/31/1988, 50th-percentile results, type M. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-11. Predicted (type M, 5 μm) uranium urinary excretion from eight independent intakes, 

50th percentile, 1952 to 1988. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

___ predicted 

excretion
Figure B-12. Predicted (type M, 5 μm) uranium urinary excretion from eight independent intakes, 

84th percentile, 1952 to 1988. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(84th percentile)

___ predicted 

excretion
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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure B-13. Assumed uranium intake, 01/01/1947 to 02/28/1978, 50th-percentile results, type S. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

Figure B-14. Assumed uranium intake, 01/01/1947 to 04/30/1952, 50th-percentile results, type S. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-15. Assumed uranium intake, 08/01/1953 to 11/30/1953, 50th-percentile results, type S. 

Figure B-16. Assumed uranium intake, 11/01/1956 to 02/28/1959, 50th-percentile results, type S.

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-17. Assumed uranium intake, 06/01/1961 to 12/31/1962, 50th-percentile results, type S. 

Figure B-18. Assumed uranium intake, 10/01/1968 to 12/31/1972, 50th-percentile results, type S. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-19. Assumed uranium intake, 03/01/1978 to 09/30/1984, 50th-percentile results, type S. 

Figure B-20. Assumed uranium intake, 10/01/1984 to 12/31/1988, 50th-percentile results, type S. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit
Figure B-21. Predicted (type S, 5 μm) uranium urinary excretion from eight independent intakes, 50th 

percentile, 1952 to 1988. 

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ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(50th percentile)

___ predicted 

excretion
Figure B-22. Predicted (type S, 5 μm) uranium urinary excretion from eight independent intakes, 

84th percentile, 1952 to 1988. 

Document No. ORAUT-TKBS-0014-5 Revision No. 04 Effective Date: 08/03/2023 Page 81 of 81

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

O monthly 

bioassay value 

(84th percentile)

___ predicted 

excretion