ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 79

DOE Review Release 03/14/2012

Document Title:

Y-12 National Security Complex – 

Occupational Internal Dose

Document Number: ORAUT-TKBS-0014-5 

Revision: 03

Effective Date: 03/12/2012

Type of Document: TBD

Supersedes: ORAUT-TKBS-0014-5 

Revision 02 PC-1 and

ORAUT-OTIB-0029 

Revision 00

Subject Expert(s): Bryce L. Rich and Melton H. Chew

Approval: Signature on File Approval Date: 03/06/2012

Steven R. Reed, Document Owner

Concurrence: Signature on File Concurrence Date: 03/06/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 03/07/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 03/06/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/12/2012

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 2 of 79

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. Approval:

Signature on File 10/06/2005

William E. Murray, TBD Team Leader

Signature on File 10/04/2005

Judson L. Kenoyer, Task 3 Manager

Signature on File 10/04/2005

Richard E. Toohey, Project Director

Signature on File 10/11/2005

James W. Neton, Associate Director for Science

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. 

Approval:

Signature on File 01/11/2006

William E. Murray, TBD Team Leader

Signature on File 01/10/2006

Judson L. Kenoyer, Task 3 Manager

Signature on File 01/12/2006

Kate Kimpan, Project Director

Signature on File 01/12/2006

James W. Neton, Associate Director for Science

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. 

Approval: 

Signature on File 02/09/2006

William E. Murray, TBD Team Leader
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 3 of 79

Signature on File 02/08/2006

Judson L. Kenoyer, Task 3 Manager

Signature on File 02/07/2006

Kate Kimpan, Project Director

Signature on File 02/14/2006

James W. Neton, Associate Director for Science

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 1943-1947, and coworker 

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

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. Approval: 

Signature on File 10/05/2006

William E. Murray, Document Owner

Richard Merrill Signature on File for 10/05/2006

John M. Byrne, Task 3 Manager

Signature on File 10/05/2006

Edward F. Maher, Task 5 Manager

Signature on File 10/05/2006

Kate Kimpan, Project Director

Brant A. Ulsh Signature on File for 10/10/2006

James W. Neton, Associate Director for Science

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 U232 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. 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 4 of 79

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 Cohorts .......................................................................................... 11

5.1.3.1 March 1, 1943 through December 31, 1947 ................................................. 11

5.1.3.2 January 1, 1948 to December 31, 1957 ....................................................... 11

5.2 Uranium Solubility in the Lung................................................................................................ 12

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

5.2.2 Uranium Alpha Activity as a Function of Enrichment................................................... 15

5.2.3 Temporal Pattern of Uranium Exposures.................................................................... 16

5.2.4 Other Radionuclides of Concern Including Recycled Uranium 

Contaminants ............................................................................................................. 17

5.2.4.1 Unmonitored Radionuclides from Recycled Uranium ................................... 17

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

5.3.1.1 Coverage ..................................................................................................... 23

5.3.1.2 Sample Collection ........................................................................................ 23

5.3.1.3 Minimum Detectable Activities...................................................................... 24

5.3.2 In Vitro Methods for Individual Radionuclides ............................................................. 26

5.3.2.1 In Vitro Bioassay for Uranium....................................................................... 27

5.3.2.2 In Vitro Bioassay for Plutonium .................................................................... 29

5.3.2.3 In Vitro Bioassay for Tritium ......................................................................... 29

5.3.2.4 In Vitro Bioassay for Other Radionuclides .................................................... 30

5.3.3 Fecal Sample Analysis ............................................................................................... 30

5.3.4 Interferences .............................................................................................................. 30

5.4 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting 

Practices ................................................................................................................................ 32

5.4.1 Whole-Body Counting................................................................................................. 32

5.4.2 Chest Counting........................................................................................................... 32

5.4.2.1 Chest Counting for Uranium......................................................................... 33

5.4.2.2 Chest Counting for Other Actinides .............................................................. 34

5.4.2.3 Chest Counting for Other Radionuclides ...................................................... 34

5.5 Attributions and Annotations .................................................................................................. 35

References ........................................................................................................................................ 36

Glossary ............................................................................................................................................ 42

ATTACHMENT A, OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS ................. 47

ATTACHMENT B, INTERNAL DOSIMETRY COWORKER DATA FOR Y-12.................................... 53
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 5 of 79

LIST OF TABLES

TABLE TITLE PAGE

5-1 Enrichment assumptions for dose assessment for analytical techniques................................ 16

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

recycled uranium contaminants through Plant history ............................................................ 18

5-3 Total Y-12 recycled uranium receipts summary...................................................................... 18

5-4 Primary contaminants in VHEU fuels and of concern in internal dose 

reconstruction ........................................................................................................................18

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

levels reported in HEU shipments to Y-12.............................................................................. 20

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 ratios ...................................................................................................................29

5-11 Reported lung counting detection limits for 235U ..................................................................... 33

5-12 Reported lung counting detection limits for 238U ..................................................................... 33

A-1 Uranium urinalysis detection limits ......................................................................................... 51

A-2 Other in vitro detection limits .................................................................................................. 51

A-3 Reported in vivo detection limits for other radionuclides......................................................... 52

B-1 Monthly uranium 24-hour urinary excretion rate analyses, 1952 to 1988................................ 56

B-2 Type F uranium intake periods and rates ............................................................................... 63

B-3 Type M uranium intake periods and rates .............................................................................. 63

B-4 Type S uranium intake periods and rates ............................................................................... 64

B-5 Type F uranium intake periods and rates ............................................................................... 65

B-6 Type M uranium intake periods and rates .............................................................................. 65

B-7 Type S uranium intake periods and rates ............................................................................... 66

LIST OF FIGURES

FIGURE TITLE PAGE

5-1 Historical solubility types meeting the type M definition .......................................................... 13

5-2 Empirical and theoretical uranium excretion after uranium restriction..................................... 14

5-3 Lung deposition and retention of class W, class Q, and type M uranium................................ 15

5-4 Variation of activity in natural thorium after a single purification.............................................. 35

B-1 Assumed uranium intake, 1/1/47 to 12/31/1969, 50th-percentile results, type F..................... 66

B-2 Assumed uranium intake, 1/1/70 to 12/31/1988, 50th-percentile results, type F..................... 67

B-3 Assumed uranium intake, 1/1/47 to 2/28/78, 50th-percentile results, type M .......................... 68

B-4 Assumed uranium intake, 1/1/47 to 4/30/52, 50th-percentile results, type M .......................... 68

B-5 Assumed uranium intake, 8/1/53 to 11/30/53, 50th-percentile results, type M ........................ 69

B-6 Assumed uranium intake, 11/1/56 to 2/28/59, 50th-percentile results, type M ........................ 69

B-7 Assumed uranium intake, 6/1/61 to 12/31/62, 50th-percentile results, type M ........................ 70

B-8 Assumed uranium intake, 10/1/68 to 12/31/72, 50th-percentile results, type M ...................... 70

B-9 Assumed uranium intake, 3/1/78 to 9/30/84, 50th-percentile results, type M .......................... 71

B-10 Assumed uranium intake, 10/1/84 to 12/31/88, 50th-percentile results, type M ...................... 71

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

percentile, 1952 to 1988......................................................................................................... 72
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 6 of 79

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

percentile, 1952 to 1988......................................................................................................... 73

B-13 Assumed uranium intake, 1/1/47 to 2/28/78, 50th-percentile results, type S........................... 74

B-14 Assumed uranium intake, 1/1/47 to 4/30/52, 50th-percentile results, type S........................... 75

B-15 Assumed uranium intake, 8/1/53 to 11/30/53, 50th-percentile results, type S......................... 75

B-16 Assumed uranium intake, 11/1/56 to 2/28/59, 50th-percentile results, type S......................... 76

B-17 Assumed uranium intake, 6/1/61 to 12/31/62, 50th-percentile results, type S......................... 76

B-18 Assumed uranium intake, 10/1/68 to 12/31/72, 50th-percentile results, type S....................... 77

B-19 Assumed uranium intake, 3/1/78 to 9/30/84, 50th-percentile results, type S........................... 77

B-20 Assumed uranium intake, 10/1/84 to 12/31/88, type S ........................................................... 78

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

percentile, 1952 to 1988......................................................................................................... 78

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

percentile, 1952 to 1988......................................................................................................... 79
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 7 of 79

ACRONYMS AND ABBREVIATIONS 

AMAD aerodynamic median activity diameter

Bq becquerel

CEDE committed effective dose equivalent

CER Center for Epidemiologic Research

Ci curie

cpm counts per minute

d day

DF decontamination factor

DHHS U.S. Department of Health and Human Services

dpm disintegrations per minute

DU depleted uranium (less than 0.72% 235U)

EU enriched uranium

FEMC Feed Materials Production Center

g gram

GSD geometric standard deviation

hr hour

HEU highly enriched uranium (generally between 20% and 90% 235U)

HPGe high-purity germanium (detector)

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

keV kilovolt-electron, 1,000 electron volts

kg kilogram

L liter

LC critical level (commonly 5% chance of a false positive)

LD detection level (commonly 5% chance of a false negative)

LEU low enriched uranium (generally less than 20% 235U)

m meter

MDA minimum detectable activity

mg milligram

min minute

mL milliliter

mm millimeter

MPLB maximum permissible lung burden

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 8 of 79

NU natural uranium

ORAU Oak Ridge Associated Universities

pCi picocurie

ppb parts per billion

ppm parts per million

RU recycled uranium 

RWP Radiation Work Permit

R&D research and development

SEC Special Exposure Cohort

SMC Idaho Specific Manufacturing Capability

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

U.S.C. United States Code

VHEU very highly enriched uranium (greater than 90% 235U)

wt % weight percent

yr year

§ section or sections

μCi microcurie

μg microgram

μm micrometer
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 9 of 79

5.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] guidelines established under 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2010):

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

conventional structures

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

reasons

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 10 of 79

5.1.1 Purpose

The purpose of this technical basis document (TBD) is to describe internal dosimetry systems and 

practices at the Y-12 Plant (now the Y-12 National Security Complex). 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 BWXT Y-12, L.L.C. 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 coworker 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. 

However, the calutrons in Building 9204-3 remained operational to permit research and development 

(R&D) to improve the production of the calutron concept and to separate other element isotopes. 

Eight calutrons were still operating as late as 1997. This program produced an inventory of 225 

isotopes from nearly every element on the periodic chart. 

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

However, the past operation has led to internal dose concerns about polonium isotopes and alpha 

airborne activity (Livingston and Martin 1952). The 86-inch Cyclotron was later used to create 

neutron-deficient radionuclides as a part of the R&D program noted above. Some of these R&D 

efforts involved plutonium, and workers could have plutonium bioassay results in their records 

(Patterson, West, and McLendon 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). 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 11 of 79

5.1.3 Special Exposure Cohorts

5.1.3.1 March 1, 1943 through December 31, 1947

On August 15, 2008, as provided for under 42 U.S.C. § 7384q(b), the Secretary of HHS designated 

the following class of employees as an addition to the SEC:

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 (73 FR 58966; October 8, 2008). 

As pointed out in the NIOSH evaluation report and the HHS Designation Letter, the class associated 

with 73 FR 58966, which became effective September 14, 2008, was assembled to clarify and correct 

the class wording defined in 70 FR 57602, because of implementation issues associated with the 70 

FR 57602 SEC class definition. 

Through the course of ongoing dose reconstruction and research, NIOSH has determined that, 

because exposure potential may not have been limited to only specific buildings or groups of workers 

at Y-12, it is necessary to expand the SEC class definition to include 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, through December 31, 1947. The designated class 

evaluation recommends a class that is consistent with current NIOSH methods for defining a 

recommended SEC class (NIOSH 2008).

In its evaluation (NIOSH 2008), NIOSH found that it is not possible to reconstruct internal radiation 

doses for Y-12 employees for all potential radiation exposures. However, NIOSH has decided that 

the occupational medical dose and external exposures directly associated with the calutron uranium 

enrichment process can be estimated. Although NIOSH found that it is not possible to reconstruct 

radiation doses completely for the SEC class, NIOSH intends to use limited available external dose 

rate measurements that might be applicable to an individual claim, and any internal and external 

monitoring data that might become available for an individual claim (and can be interpreted using its 

existing dose reconstruction processes or procedures). Therefore, dose reconstructions for 

individuals employed at Y-12 during the period from March 1, 1943, through December 31, 1947, but 

who do not qualify for inclusion in the SEC may be performed using these data as appropriate.

The assessment of NIOSH’s ability to reconstruct doses from 1948 onward, is addressed in the 

following section.

5.1.3.2 January 1, 1948 to December 31, 1957

On October 18, 2011, as provided for under 42 U.S.C. § 7384q(b), the Secretary of HHS designated 

the following class of employees as an addition to the SEC:

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

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

workdays within the parameters established for one or more other classes of employees in the 

SEC (76 FR 72928; November 28, 2011).
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 12 of 79

As pointed out in the NIOSH evaluation report and the HHS Designation Letter, the class associated 

with 76 FR 72928, which became effective November 17, 2011, was assembled to clarify and correct 

the class wording defined in 71 FR 55477, because of implementation issues associated with the 71 

FR 55477 SEC class definition. 

Through the course of ongoing dose reconstruction and research, NIOSH has determined that, due to 

undocumented worker movements across the site, limited claimant-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 has 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 has determined that it is necessary to remove the area-specific and 

monitoring criteria from the class description associated with SEC-00028 thorium and cyclotron 

exposures for the period from January 1, 1948, through December 31, 1957. NIOSH has determined 

that it is also necessary to expand the SEC class definition to include 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 January 1, 1948, through December 31, 1957. The designated class 

evaluation recommends a class that is consistent with current NIOSH methods for defining a 

recommended SEC class (NIOSH 2011). 

In its evaluation (NIOSH 2011), NIOSH found that it is not possible to reconstruct internal radiation 

doses for Y-12 employees with internal exposures to thorium and cyclotron radionuclides. However, 

NIOSH has decided that the occupational medical dose and external exposures can be estimated. 

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

class, NIOSH intends to use any internal and external monitoring data that may become available for 

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

or procedures). Therefore, dose reconstructions for individuals employed at Y-12 during the period 

from January 1, 1948, through December 31, 1957, but who do not qualify for inclusion in the SEC, 

may be performed using these data as appropriate. 

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 2007). 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. Health physicists also recognized early 

that the lung is the primary organ of concern for the less soluble uranium compounds. During the 

greater part of the Plant’s history, control measures for less soluble compounds were guided by the 

lung and metabolic models in International Commission on Radiological Protection (ICRP) 

Publication 2 (ICRP 1959) or by similar predecessor models. These models related lung dose to 

uranium excretion in urine, particularly under chronic exposure conditions. 

Insoluble material described in ICRP Publication 2 (ICRP 1959) would be classified as moderately 

soluble type M material in the framework of ICRP Publication 66 (ICRP 1994). In addition, ICRP has 

recognized that some forms of uranium are less soluble than initially believed. Figure 5-1 shows 

normalized lung retention for ICRP Publication 2 insoluble material, ICRP Publication 30 (ICRP 1979) 

class W particles, class Q material as described in the 1990s (Barber and Forrest 1995), and the 

Publication 66 type M particles. The boxes on the diagram define the modern clearance types F 

(fast), M (moderate), and S (slow) as a function of lung retention and time, as illustrated in Annex D of 

ICRP Publication 71 (ICRP 1995a). The ICRP Publication 2 insoluble model and the class W models 

fit within the range of type M material. 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 13 of 79

Figure 5-1. Historical solubility types meeting the type M definition. 

After in vivo lung counting was begun at Y-12 in 1961, it was recognized that a few workers had 

uranium lung burdens in considerable excess of the amount indicated by urinalysis. West and Scott 

(1969) noted that, “exposure cases noted in 1962 and 1963 showed chest burdens of enriched 

uranium [EU] which exhibited unexpectedly slow decreases after the persons were assigned to nonuranium activities.” The job description of all individuals was consistent (Snapp 2003b; Barber and 

Forrest 1995).

Subsequent investigations suggested that three process combustion ashes from the EU foundry were 

the most likely exposure materials for these cases (Steckel and West 1966, p. 31). The particular 

thermal histories and particle size distributions of these process ashes were considered to be 

important factors leading to the low solubility in simulated lung fluid (Steckel and West 1966, p. 30). 

Particle sizes tended to be relatively larger (0.86 to 1.6 μm), and the solubility tended to be lower, for 

materials processed at higher temperatures.

While the exceptional cases with unusually protracted lung clearance are important, it is more 

important to note that, for the vast majority of individuals, lung clearance took place in approximate 

accordance with the ICRP Publication 2 (ICRP 1959) insoluble model, which fits within the current 

type M framework. 

In 1964, a composite urinary excretion curve was described for 157 workers who had been removed 

from uranium work because of high uranium urinalysis results (Scott 1964). Figure 5-2 shows this 

empirical composite curve together with expected excretion curves for types F, M, and S uranium. 

The latter curves are derived from 5-μm aerodynamic median activity diameter (AMAD) excretion 

curves (Potter 2002). They were normalized to match the Scott curve at 200 days after removal from 

uranium work areas. The empirical excretion curve matches the type M curve much better than the 

type F or S curves. Indeed, the Scott curve fits the type M curve quite well over most of the 420-d 

time range. These results are consistent with the subsequent conclusions of Rucker et al. (2001) that 

class W is the appropriate default for uranium at Bechtel Jacobs facilities, including Y-12. Exceptions 

to this default included high-fired uranium oxides and compounds formed by the slow oxidation of 

uranium metal.

0

0.2

0.4

0.6

0.8

1

0 30 60 90 120 150 180 210 240 270 300 330 360 390 420

Days

Fraction of Initial Alveolar Deposi

ICRP 2 (I)

Class'Q' (8um)

Type M (5 um)

Class W (1 um)

Type S

Type M

F

Deposition
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 14 of 79

Figure 5-2. Empirical and theoretical uranium excretion after uranium 

restriction. 

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 default modeling assumption was class Q (90% Super-W, 10% Y). 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, p. 28). 

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 clearance 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. 669). 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 2007, pp. 19 ff.).

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., Ballenger

et al. 1953; Steckel and West 1966; Barber and Forrest 1995; Snapp 2003a). For different times and 

different processes, reported particle sizes ranged from less than 1 to greater than 10 μm (physical). 

0E+00

1E-05

2E-05

3E-05

4E-05

0 15 30 45 60 75 90 105 120 135 150 165 180 195

Days Post 300 Days Continuous Intake

Urinary Excretion (μCi/d)

Scott 1964

ICRP 68, F

ICRP 68, M

ICRP 68, S
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 15 of 79

Steckel and West (1966) reported a positive correlation between uranium oxide particle size and 

process temperature. Barber and Forrest (1995) used an 8-μm 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.) In terms of lung deposition and retention, the 8-μm AMAD 

class Q material is closer to 5-μm AMAD type M material than to 1-μm AMAD class W material, as 

shown in Figure 5-3.

Figure 5-3. Lung deposition and retention of class W, class Q, and type M 

uranium. (Particle sizes specified are in micrometers AMAD.) (Barber and 

Forrest 1995).

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 (Veinot 2003, 

p. 89). This is the justification for the change to the 5-μm AMAD as the default particle size. Y-12 

uses the newer weighting factors proposed by the ICRP in its 1990 recommendations (ICRP 1991). 

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 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.(Patterson 1958). Early in vivo chest-counting results were reported either 

as micrograms of 235U or milligrams of 238U based on whether an exposure was believed to be from 

EU, natural uranium (NU), or DU (Cofield 1960; Scott and West 1975). Based on these reporting 

0.0001

0.001

0.01

0.1

1

1 10 100 1000

Days

Fraction of Intake in Lung .

Class W, 1 um

Type M, 5 um

Class Q, 8 um
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 16 of 79

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

1950 through 1989 for urinalysis and from 1961 to 1989 for lung counting. 

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

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. 

Before 1950, 93% enrichment.

After 1949, NU. 

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. 

Lung count μg U-235 93% enrichment.

Lung count μg U-238 NU.

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 for the dose assessment and the 234U dose conversion factor applied. This 

practice is slightly conservative for EU; the resultant overestimate is less than 1% of the effective 

dose. For DU, the use of the 234U dose conversion factor leads to an overestimate of 16% for type S 

uranium and 20% for type M uranium (Eckerman and Kerr 1999, p. 16). Note that 234U dose 

conversion factors are not appropriate substitutes for 232

U, as the DCFs for 232U are typically 5-10 

times larger than those for 234U. 

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, Patterson (1958, p. 58) 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” (Veinot 2003, p. 107). 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 plant-wide basis (see 

Table 5-8 in Section 5.2.4.1 for plant-wide RU contaminant defaults).
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 17 of 79

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 2003a):

• 3

H, 90Sr, 99Tc

• 228Th, 232Th

• 232U, 233U, 236U 

• 238Pu, 239Pu, 241Pu

• 237Np, 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 208Tl.

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 2007).

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 (FEMC) in Fernald, 

Ohio, which also had an RU contaminant inventory. Therefore, the DU contaminant levels at Y-12 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 18 of 79

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

uranium contaminants through Plant history (1953 – 1999).

a

Site U (kg)

Pu Np Tc

g ppb pCi/μg U g ppm pCi/μg U g ppm pCi/μg U

Y-12 VHEU Receipts Summary – Rich et al. (2000)

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 Total 150,857 0.04674 0.3 4.3E-03 3,666 23 1.6E-02 14,499 96 1.64

Y-12 DU Receipts Summary - Rich et al. (2000)

Hanford 1,502 No data

K-25 192,836

PGDP 38,423

DU total 232,761

Grand total 383,618

a. SRS = Savannah River Site; ICPP = Idaho Chemical Processing Plant; K-25 = Oak Ridge Gaseous Diffusion Plant; 

PGDP = Paducah Gaseous Diffusion Plant.

Table 5-3. Total Y-12 recycled uranium receipts summary (DOE 2007).

Shipping site

U 

(kg)

Pu Np Tc

g ppb pCi/μg U g ppm pCi/μg U g ppm pCi/μg U

SRS 153,000

ICPP 25,700

Hanford 4,400

West Valley 900

Total 184,000 0.5 2.7 0.039 54 0.003 0.00021 9,100 49.5 0.85

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

reconstruction (Wenzel 2000, 2004). 

Isotope Half-life

Activity

(μCi/μg total U)

Activity

(μCi/g total U )

Percent weight

g/100 g Ua ppm U

Activity ratio

(isotope/U)

Tc-99 2.11E+05 yr 3.8E-04 3.8E+02 2.2E+00 2.2E+04 24

Th-227 18.72 d 3.3E-11 3.3E-05 1.1E-13 1.1E-09

Th-228 1.913 yr 1.6E-07 1.6E-01 1.9E-08 1.9E-04 0.01

Th-229 7.34E+03 yr 1.4E-12 1.4E-06 6.7E-10 6.7E-06

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 0.1

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

Total 1.9

U-232 6.89E+01 yr 2.7E-07 2.7E-01 1.3E-06 1.3E-02

U-233 1.59E+05 yr 2.6E-10 2.6E-04 2.7E-06 2.7E-02

U-234 2.46E+05 yr 6.6E-07 6.6E-01 1.06E-02 1.0E+04

U-235 7.04E+08 yr 1.7E-06 1.7E-00 7.8E+01 7.8E+05

U-236 2.34E+07 yr 1.3E-05 1.3E+01 2.0E+01 2.0E+05

U-237 6.75 d 3.6E-08 3.6E-02 4.4E-11 4.41E-07

U-238 4.47E+09 yr 5.6E-09 5.6E-03 1.7E+00 1.7E+04

Total 15.7

Np-236 1.15E+05 yr 9.4E-11 9.4E-05 7.2E-07 7.2E-03

Np-237 2.14E+06 yr 9.1E-06 9.1E+00 1.3E+00 1.3E+04 0.6

Np-238 2.117 d 9.2E-10 9.2E-04 3.6E-13 3.6E-09

Np-239 2.355 yr 4.3E-08 4.3E-02 1.9E-11 1.9E-07

Total 9.1

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

Pu-238 87.74 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 1.2

Pu-240 6.54E+03 yr 1.5E-05 1.5E+01 6.4E-03 64 1.0
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 19 of 79

Isotope Half-life

Activity

(μCi/μg total U)

Activity

(μCi/g total U )

Percent weight

g/100 g Ua ppm U

Activity ratio

(isotope/U)

Pu-241 14.4 yr 1.5E-03 1.5E+03 1.4E-03 14 1.0E+02

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

Total 4.6E+04

Am-241 4.32E+02 yr 9.6E-06 9.6E+00 2.8E-04 2.8 0.6

a. 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.

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

reports (ORAUT 2004; 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 

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, FMPC, and others, which also had RU contaminant 

inventories. Thus, the discrepancies in the current 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 clamant value is used to derive the defaults for dose reconstruction, which 

are generally the results in BWXT Y-12 (2000).
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 20 of 79

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 for 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).

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 fuel

Approximate derived 

DF

Th-228 0.01–0.03 NAa 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. The decontamination 

factor (DF) is interpolated from Y-12 process stream analytical data. These values were taken directly from 

column 4 in Table 5-4.

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

received at Y-12 provides a rough indication of the DFs of the chemical extraction processes. 

Reference to these DFs in Table 5-5 and the average contaminant 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 plant 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 (summary of tables on pp. B-3 and B-4 of BWXT Y-12 2000).

a 

Contaminant Range Ave. of range

Ave. of range 

minus the high

Raffinate 

valuesb

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. The highest values reported in the processing facilities were in just one 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. 

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

Therefore, the parts per million 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).
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 21 of 79

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 2003a). 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 consistent with 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 plant-wide basis. The 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 microgram or picocurie 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.

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

levels.a

Uranium 

isotope

Activity 

(Ci/g) Mass fraction

Activity 

(pCi/μg total U) RU cont.

Activity 

(pCi/μg total U)

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

233 9.66E-03 1.00E-04 9.660E-01 Pu-238 6.20E-03

234 6.22E-03 1.28E-02 7.962E+01 Pu-239 1.00E-04

235 2.16E-06 5.22E-01 1.128E+00 Np-237 2.00E-03

236 6.50E-05 2.92E-01 1.898E+01 Tc-99 1.90E+00

238 3.36E-07 1.73E-01 5.796E-02 Th-228 3.17E-02

Total 9.99E-01 1.007E+02 Total alpha 4.00E-02

dpm/μg total U 2.24E+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

233 9.66E-03 3.000E-06 2.90E-02 Pu-238 2.00E-03

234 6.22E-03 1.000E-02 6.22E+01 Pu-239 3.30E-05

235 2.16E-06 7.800E-01 1.68E+00 Np-237 2.00E-03

236 6.50E-05 2.000E-01 1.30E+01 Tc-99 1.50E-01

238 3.36E-07 1.700E-02 5.71E-03 Th-228 5.00E-02

Total 1.007E+00 7.69E+01 Total α 5.40E-02

dpm/μg total U 1.71E+02 α dpm/μg total U 1.20E-01

U α:Non U α 1.4E+03

Weapons-grade Oralloy VHEUb, c

233 9.66E-03 0.000E+00 0.00E+00 Pu-238 2.00E-02

234 6.22E-03 1.100E-02 6.84E+01 Pu-239 2.00E-02

235 2.16E-06 9.300E-01 2.01E+00 Np-237 6.00E-02

236 6.50E-05 7.500E-03 4.88E-01 Tc-99 5.00E-02

238 3.36E-07 6.000E-02 2.02E-02 Th-228 5.00E-02

Total 1.009E+00 7.09E+01 Total α 1.50E-01

dpm/μg total U 1.57E+02 α dpm/μg total U 3.33E-01

U α:Non U α 4.73E+02

Recycled NUd

233 9.66E-03 0.000E+00 0.00E+00 Pu-238 0.00E+00

234 6.22E-03 5.400E-05 3.36E-01 Pu-239 3.00E-06

235 2.16E-06 7.200E-03 1.56E-02 Np-237 2.40E-05

236 6.50E-05 0.000E+00 0.00E+00 Tc-99 2.00E-04

238 3.36E-07 9.927E-01 3.34E-01 Th-228 3.00E-05
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 22 of 79

Uranium 

isotope

Activity 

(Ci/g) Mass fraction

Activity 

(pCi/μg total U) RU cont.

Activity 

(pCi/μg total U)

Total 6.85E-01 Total alpha 5.70E-05

dpm/μg total U 1.52E+00 α dpm/μg total U 1.27E-04

U α:Non U α 1.20E+04

Recycled LEU (2%)d

233 9.66E-03 0.000E+00 0.00E+00 Pu-238 0.00E+00

234 6.22E-03 2.000E-04 1.24E+00 Pu-239 2.00E-04

235 2.16E-06 2.000E-02 4.32E-02 Np-237 2.00E-04

236 6.50E-05 0.000E+00 0.00E+00 Tc-99 9.20E-02

238 3.36E-07 9.798E-01 3.29E-01 Th-228 3.00E-04

Total 1.62E+00 Total alpha 7.00E-04

dpm/μg total U 3.59E+00 α dpm/μg total U 1.55E-03

U α:Non U α 2.31E+03

Recycled DUd

233 9.66E-03 0.000E+00 0.00E+00 Pu-238 0.00E+00

234 6.22E-03 1.000E-05 6.22E-02 Pu-239 3.70E-06

235 2.16E-06 2.000E-03 4.32E-03 Np-237 2.70E-05

236 6.50E-05 3.000E-06 1.95E-04 Tc-99 4.00E-04

238 3.36E-07 9.980E-01 3.35E-01 Th-228 4.00E-05

Total 4.02E-01 Total alpha 7.07E-05

dpm/μg total U 8.93E-01 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 was 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. 

Isotope

Default levels (pCi/μg U) Default levels (pCi/pCi total U)a

VHEU & Oralloy LEU, NU, & DU VHEU & Oralloy LEU, NU, & DU

Tc-99 4 0.3 4E-02 0.15

Th-228b 0.5 0.003 5E-03 1.5E-03

Np-237 0.6 0.003 6E-03 1.5E-03

Pu-238c 0.2 2E-03

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

Total non-U 1.3 0.01

Total U 75 2

a. Reference to Table 5-7 indicates activity in picocuries per gram of total U for VHEU ranges from 71 to 

101; 100 is used for simplicity. For LEU, NU, and DU, the picocuries per gram of total U ranges from 

0.9 to 1.6; 2 pCi/g total U is used for simplicity.

b. The Th-228 default is justified in concept due to the demonstrated existence of general plant-wide 

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).

c. 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. In Vitro Minimum 

Detectable Activities, Counting Methods, and Reporting Protocols

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

of uranium per liter or lung count, the default levels in picocuries 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 table can be used. Again, these values are maximum values expected, while 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 23 of 79

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.

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., it was enriched at the gaseous diffusion plants, where 239Pu was 

dominant in comparison with 238Pu) (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

5.3.1.1 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). 

There is not sufficient information 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 the Y-12 National Security Complex – Occupational Environmental Dose (ORAUT 

2005). 

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). Y-12 currently 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 (Veinot 2003, p. 23). 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.

5.3.1.2 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. 25), 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

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

excretion. 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 24 of 79

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

gross alpha counting (McRee, West, and McLendon 1965, p. 25): 

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, West, and McLendon 

(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, West, and McLendon 1965, p. 27). 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 is discussed below.

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

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 2003a, p. 70).

Routine Urine Sample Frequency

Urine samples were collected monthly in 1950, and weekly collection for some employees 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. 22). 

The current practice is to schedule samples 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 2003a).

5.3.1.3 Minimum Detectable Activities

Normal and Depleted Uranium in Urine by Fluorometry (1945 to 1989) and Enriched Uranium 

(1948 to 1950)

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 

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 of 7 μg/d 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 25 of 79

for a nominal 1.4 L/d urine excretion rate (Struxness 1949a, CCCC 1951). In 1952, the instrument 

sensitivity and sample volumes indicated a minimum detectable activity (MDA) of 7 μg/d for 1.4 L/d of 

urine excreted (Ballenger et al. 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. 6) 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. 6). 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. 38). Counters were not used if the background was greater than 5 counts in 30 

minutes (Hamrick 1958, p. 7). For a disk counted for 120 minutes (in fact, two disks were counted for

60 minutes each), assuming a well-known background and alpha of 0.05, the detection level LD

(Currie 1968) 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 Patterson (1958), the LD would be 46 dpm/d. 

With an increase in average uranium recovery to 73% (McRee, West, and McLendon 1965 p. 33), LD

would be reduced to 25 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 46 dpm/d 

before 1965 and 25 dpm/d after 1965 (Currie 1968). 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 present 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, Snapp, and Veinot 2002, p. 15). The LD

varies with the uranium recovery of a particular sample, and is reported with the sample result. 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 26 of 79

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, West, and McLendon 1957, 

p. 13). Results are reported in microcuries per milliliter. 

Tritium in Urine by Liquid Scintillation 

The current system is to place 5 mL of middle distillate from a urine sample in liquid scintillation fluid 

for counting. Assuming typical method efficiencies, the MDA is 2,000 dpm/d (Hursh 1958).

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. Actual sample-specific MDAs are included in the analysis reports (Snapp

2003a, Table 4.2, p. 92).

Other Actinides in Urine and Feces

Table 5-9 lists the analytical laboratory LD values used since about 1988 (Snapp 2003a, p. 107).

Table 5-9. LD values. 

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 0.1

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. 25). In 2002, of 2,800 participants in the bioassay 

program, 700 Y-12 workers participated in the fecal-sampling program (Souleyrette, Snapp, and 

Veinot 2002, p. 7). 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 2003, p. 80)]. At present, the Analytical 

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

(Souleyrette, Snapp, and Veinot 2002, p. 15). 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. 7). 

5.3.2 In Vitro Methods for Individual Radionuclides

The following sections discuss the in vitro methods for specific radionuclides in urine. (Section 5.2.3 

discusses analysis of fecal samples.)
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 27 of 79

5.3.2.1 In Vitro Bioassay for 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 min/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, Reiss, and Young 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. 7; CCCC 1951, p. 14). By the 

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

(Ballenger et al. 1953, p. 28). However, after this change in technique was instituted, the urinary 

uranium concentrations increased greatly, as listed in Table 5-7. This raises the concern that the 

technique used before May 1952 could have underestimated the urinary uranium concentrations. 

(See Section 5.4.2 for a discussion of missed doses.)

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., micrograms per liter), 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 an industry standard sensitivity of about 5 ppb 

(5 μg/L) (e.g., Struxness 1949a, p. 7).

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 1959) 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 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 28 of 79

in early 1949 (Struxness 1949b, p. 7). At first, samples were subjected to an acid digestion step 

before electrodeposition. In 1951, methodological problems leading to underestimates of a factor of 2 

were noted and corrected (Ballenger 1952, p. 53). 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 (Patterson 1958, p. 34). 

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. 38). 

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-dpm/cpm 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 (Patterson 1958), but they improved over time: In 1965, the average recovery 

for electroplating was 73% (McRee, West, and McLendon 1965, p. 33). Method limitations are 

discussed in a number of reports, including Johnson et al. (1959) and UCNC (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 (Patterson 1958, p. 57).

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).

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. 6). The in vivo monitoring frequency was determined by urinalysis results and 

ranged from monthly to once each 18 months (Scott 1963, p. 1). 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. 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 dose conversion factor. 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 29 of 79

Table 5-10. 233U to 232U ratios. 

Process

U-233:U-232 

activity ratio

Reduction of UF4 1.3

Casting of metal-crucible skull 15–32

Machining final uranium component 160

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 (HF). 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 (Veinot 2003). Typical recoveries are about 85%. 

Samples are typically counted for 16 hours on a passivated implanted planar silicon detector.

5.3.2.2 In Vitro Bioassay for Plutonium

Plutonium in Urine by Gross Alpha Counting

Certain workers in the Special Testing Department were tested for plutonium intake by urinalysis 

(McRee, West, and McLendon 1965, p. 35). Because large urine volumes were required, these 

employees submitted 24-hour samples through the use of take-home kits. The sampling frequency 

was monthly (Patterson, West, and McLendon 1957, p. 38). 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-hour void) to 

millirem 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 

filter, rinsed, dried, and mounted on a planchet for alpha spectrometry (Veinot 2003). The same 

procedure is used for isotopic thorium.

5.3.2.3 In Vitro Bioassay for 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 in an ionization chamber. Results were reported in microcuries per milliliter 

(Patterson, West, and McLendon 1957, pp. 38–39).

Tritium is currently measured by liquid scintillation counting. An aliquot of the urine sample is distilled 

in a heating mantle, and the middle fraction of the distillate is collected. A 5-mL portion of the 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 30 of 79

collected distillate is mixed with liquid scintillation cocktail. The beta activity of the tritium is then 

measured by liquid scintillation counting (Veinot 2003). Exposure should be assumed to be due to 

HTO. 

5.3.2.4 In Vitro Bioassay for Other Radionuclides

Alpha spectrometry procedures described in Section 3.2 of Snapp (2003a) 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. 

5.3.3 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, Snapp, and Veinot 2002, p. 6). 

If a work area has been determined to have predominantly insoluble airborne uranium, it is current 

practice for workers with a moderate to high exposure potential (CEDE greater than 100 mrem) 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, Snapp, and Veinot 2002, p. 14) 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 2003a, p. 72). These 

normalized results or actual individual daily excretion results can be used to reconstruct dose.

5.3.4 Interferences

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.

Y-12 used the following methods to correct sample results reported after 1989 for dietary levels of 

uranium. This section presents information about Plant practices to correct excretion sample results 

for dietary uranium. The range of background uranium values expected to be observed in urine 

samples is 0.14 to 0.57 dpm, and the range of background uranium values expected to be observed 

in the fecal samples is 2.32 to 2.75 dpm (Veinot 2003, p. 75). Therefore, at Y-12, the reference point 

is established at the midpoint of the quoted ranges of values for urine (0.35 dpm/d) and fecal samples 

(2.5 dpm/sample). In addition, the ratios between 234U and 238U can provide information to help 

characterize the sample. Natural background uranium typically has a 234U-to238U ratio of 

approximately 1:1. EU also has a characteristic ratio, typically greater than 3:1. DU, on the other 

hand, typically has a characteristic ratio of less than 0.9:1 (Snapp 2003a).
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 31 of 79

Based on the above information, the background uranium corrections used at Y-12 are as follows:

For bioassay sample results less than the detection limit for the analytical technique, no action is 

required. Corrections for samples from the early days when MDAs were significantly larger are not 

required.

To adjust results for background, the 234U-to238U ratios are reviewed for each sample to determine if 

the sample is consistent with EU or DU. Conservative ratios are applied in this determination: All 

samples with 234U-to238U ratios greater than 1:1 are considered to be consistent with EU, and all 

samples with 234U-to238U ratios less than 1:1 are considered to be consistent with DU.

The background contribution is subtracted from EU samples as follows: 

a. Assume that the ratio of 234U to 238U is 1:1 for dietary uranium. 

b. Determine if the 238U result exceeds its corresponding critical level value. 

c. If the 238U result is greater than the critical level, subtract two times that 238U result from the 

total uranium result. 

d. If the 238U result is less than its corresponding critical level value, do not subtract a dietary 

component.

For DU samples (e.g., those with a 234U-to238U ratio less than 1:1), the following steps are used to 

determine the dietary correction:

a. The following equations are solved to determine an equation for the occupational 238U content 

and the occupational 234U content in a DU sample:

diet

diet

1

234

238

U

U = (5-2)

occ

occ

0.2

234

238

U

U = (5-3)

actual diet occ

238 238 238 U UU = + (5-4)

actual diet occ

234 234 234 U UU = + (5-5)

b. Using Equation 5-5, and substituting the 238Udiet for the 234Udiet based on Equation 5-2 and 

substituting 0.2 (238Uocc) based on Equation 5-3 for the 234Uocc, the following relationship 

results: 

actual diet 0.2( ) occ

234 238 238 UU U = + (5-6)

Because the 238Udiet can be replaced with 238Uactual – 238Uocc derived from Equation 5-4, 

Equation 5-5 can be solved to determine the occupational quantity of 238U. The resultant 

equation for determining the occupational quantity of 238U is:
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 32 of 79

actual actual

occ 0.8

234 238 U U 238U − = − (5-7)

c. Once the occupational 238U quantity is known, the occupational quantity of 234U can be derived 

using Equation 5-3:

occ 0.2( ) occ

234 238 U U = (5-8)

d. Ensure that both the 234U and 238U results exceed their corresponding critical level values. If 

both results are not greater than their corresponding critical level values, a background 

determination cannot be made. In this case, it is assumed that all the activity observed in the 

total uranium sample result will contribute to occupational dose.

e. Use either Equation 5-7 or 5-8 to determine the occupational 238U component or the 

occupational 234U component, respectively. For ease of reference, the 238U component was 

chosen for the remaining steps.

f. Subtract the occupational component of the sample from the corresponding observed 

component to determine the dietary contributor:

actual occ diet

238 238 238 U UU − = (5-9)

g. Once the dietary component has been calculated, subtract twice this dietary value from the 

total uranium result to remove the background:

tot 2( ) diet totalocc

238 U UU − = (5-10)

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 2003a). The primary in vivo 

detection method was chest counting, as described below.

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 (1959, 1960). A 9- by 4-in. sodium iodide (thallium 

doped) [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 high-purity germanium 
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 33 of 79

(HPGe) counting system was begun. This system began routine operation in a new counting room in 

June 1985 (Souleyrette 2003). 

5.4.2.1 Chest Counting for 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 micrograms of 235U for EU and milligrams of 238U for NU or 

DU. The original limits of detection for 235U and 238U were reported as 130 μg and 13.5 mg, 

respectively (Cofield 1959). 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 of 

activity (nanocurie 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. 

Year Detectors

U-235 

(μg)

U-235 

(nCi)

U-235 

(dpm)

Reporting 

convention Reference

1959 One 9-in. NaI 130 0.28 620 Cofield 1959

1963 Two 9-in. NaI 96 0.21 470 Scott and West 1967

1965 Two 9-in., 2 5-in. NaI 72 0.16 360 Scott and West 1967

1975 Two 9-in., 2 5-in. NaI 70 0.15 330 Scott and West 1975

1983 Two 9-in., 2 5-in. NaI 62 0.13 290 1.96 σBKG King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 68.5 0.15 330 LC

a King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 137 0.30 670 LD

a King and Barclay 1983

1990 Two 9-in., 2 5-in. NaI 70 0.15 330 (b) Barber and Forrest 1995

6/1992 HPGe 46 0.10 220 (c) Veinot 2003

a. As defined by Currie (1968) with α = β = 0.05, paired observations.

b. Type I error = 5%.

c. Individual MDA or critical level supplied with analytical results.

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

Year Detectors

U-238

(mg)

U-239 

(nCi)

U-238

(dpm)

Reporting 

convention Reference

1959 One 9-in. NaI 13.5 4.5 10,000 Cofield 1959

1965 Two 9-in., 2 5-in. NaI 6.5 2.2 4,900 Scott and West 1967

1983 Two 9-in., 2 5-in. NaI 5 1.7 3,800 1.96 σBKG King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 7.2 2.4 5,300 LC

a King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 14.4 4.8 11,000 LD

a King and Barclay 1983

1991 HPGe 4.5 1.5 3,300 (b) Veinot 2003

a. As defined by Currie (1968) with α = β = 0.05, paired observations.

b. Individual MDA or critical level 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 13.5 mg, respectively, through 1990. Beginning 

in 1991, the critical level or MDA is supplied with the individual 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 employee 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.
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 34 of 79

5.4.2.2 Chest Counting for Other Actinides

Thorium

The in vivo lung count was the only monitoring technique for monitoring thorium exposure in the body 

during the Plant’s first decades. Thorium lung activity was inferred from 228Ac and/or 212Pb lung 

activity (Souleyrette 2003). Thorium lung counting was conducted from 1958 to 1984 with routine 

lung counts, scheduled at approximately 6-month intervals, starting in 1961 (BWXT, 2005).

The interpretation of thorium spectra is complicated by the tendency of decay chain members to 

become separated during operations, particularly those involving heat. The boiling point of radium is 

below the melting point of thorium. The 228Ac gamma radiation in an in vivo count could be an 

indication of a 228Ra intake or an intake of the entire 232Th chain. The 239-keV gamma line from 212Pb,

a progeny of 228Th, is a more reliable indicator of 232Th, because the thorium isotopes would remain 

together during processing. However, if the 228Ra is lost, 212Pb decreases with the 1.9-year half-life of 

its parent, 228Th, moderated by the ingrowth of new 228Ra with its 5.7-year half-life. This complicated 

disequilibrium pattern operates before and after chain members are taken into the body. Some 

knowledge of the likely state of equilibrium is, therefore, necessary to translate observed activities into 

dose.

The maximum permissible lung burden (MPLB) of 232Th (corresponding to 15 rem/yr to the lung) 

varied markedly with the ratio of 228Th to 232Th. For full equilibrium (ratio = 1), the MPLB was 

calculated to be 2.9 nCi of 232Th. For a ratio of 0.1, the lung burden was 12 nCi (West 1965, p. 22). 

For a ratio of 0.8, an MPLB of 3.2 nCi is indicated. For this ratio, with a 228Ra-to232Th ratio of 0.6, an 

MDA of 0.2 lung burdens, or 0.6 nCi, is indicated (West 1965, p. 26). In mass units, this is 5.5 mg. 

West (1965, p. 18) reported that thorium at Y-12 was processed less than 1 year after purification by 

the supplier and, as a consequence, had only about 10% as much 228Ra as 224Ra. This means the 

maximum dose conversion factor per milligram of 232Th would be less than that for 232Th in full 

equilibrium with its progeny. For calculating intakes of 228Th and 232Th from lung counts, a default 

assumption must be made for the ratio of equilibrium of 228Th to 232Th. This ratio will be used to 

determine the intake of 228Th from the intake of 232Th calculated from the lung count results (assumed 

to be 232Th). As can be seen in Figure 5-4, the activity of 228Th after initial purification declines over 

the first 5 years due to the decay of the original 228Th in the mixture, then slowly builds back up due to 

the 228Th resulting from the decay of 232Th. 

Because the actual time from initial purification is unknown and will change over time, an average of 

80% equilibrium of 228Th to 232Th should be assumed. This is a reasonable assumption because the 

actual equilibrium percentage ranges from 100% at initial purification to below 60% and then slowly 

rises again. The drop to 80% occurs within the first 6 months and continues dropping over the first 

year, so the assumption of 80% is consistent with the assumption the thorium was processed within a 

year of purification. This translates to a ratio of 228Th to 232Th of 0.8.

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).

5.4.2.3 Chest Counting for 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). Although bremsstrahlung counting could have been done for 99Tc, no information is 

available about the sensitivity of the technique.
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 35 of 79

Figure 5-4. Variation of activity in natural thorium after a single purification (West 

1965).

5.5 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

[1] Mahathy, James M. ORAU Team. Health Physicist. December 2004. 

Data extraction and all statistical analyses were performed by Mr. Mahathy and results were 

validated by Richard Sparks.

[2] Brackett, Elizabeth. ORAU Team. Principal Internal Dosimetrist. March 2005. 

Ms. Brackett modeled the Type M and S intakes as described in this section. These fits were 

reviewed by Dave Allen of NIOSH.

[3] Arno, Matthew. ORAU Team. Health Physicist. January 2012. 

Dr. Arno modeled the Type F intakes as described in this section. These fits were reviewed by 

Tom Labone.
Document No. ORAUT-TKBS-0014-5 Revision No. 03 Effective Date: 03/12/2012 Page 36 of 79

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GLOSSARY

absorption type

Categories for materials according to their rate 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 in the 

respiratory tract (slow solubilization). Also called solubility type. 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. 

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 an undesired location including air, soil, buildings, animals, and 

persons.

cyclotron

Particle accelerator capable of large beam currents where the beam is injected in the center of 

a circular magnet. A fixed radio frequency 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 roentgens, rads, reps, or grays.

dose conversion factor

Multiplier for conversion of potential dose to the personal dose equivalent to the organ of 

interest (e.g., liver or colon).

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% in 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.

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 received from radioactive material in the body.

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 utilizing instrumentation that detects 

radiation emitted from the radioactive material in the body.
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low-enriched uranium (LEU)

Uranium enriched to less than 20% 235U by mass.

maximum permissible lung burden (MPLB)

Historical occupational limit for radionuclides in the lung defined as the quantity of the isotope 

that could be present at any given time to deliver 15 rem per year to the lung at the end of a 

50-year period of chronic exposure. At Y-12 during the 1970s the occupational limit for 

plutonium was a quantity of plutonium that could be present in the chest at any given time 

equal to 16 microcuries or 0.25 micrograms of 239Pu.

minimum detectable amount or activity (MDA)

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error).

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements.

natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcuries

per gram). See uranium. 

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.

routine monitoring

Monitoring carried out at regular intervals during normal operations.
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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 (wT) 

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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TABLE OF CONTENTS 

SECTION TITLE PAGE

A.1 Conventions Used in Internal Dose Reports........................................................................... 47

A.1.1 Y-12 Urinalysis Data through September 1989........................................................... 47

A.1.2 Y-12 Urinalysis Data since October 1989 ................................................................... 48

A.1.3 In Vivo Data................................................................................................................ 49

A.2 Uranium Solubility in the Lung................................................................................................ 50

A.3 In Vitro Detection Limits ......................................................................................................... 51

A.4 In Vivo Detection Limits.......................................................................................................... 51

LIST OF TABLES

TABLE TITLE PAGE

A-1 Uranium urinalysis detection limits ......................................................................................... 51

A-2 Other in vitro detection limits .................................................................................................. 51

A-3 Reported in vivo detection limits for other radionuclides......................................................... 52

A.1 CONVENTIONS USED IN INTERNAL DOSE REPORTS 

The following is adapted from Souleyrette (2003). 

A.1.1 Y-12 Urinalysis Data through September 1989

Dept – Department number of employee at the time of monitoring.

Badge/Old SSN – Employee 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 milliliters. 

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.)

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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 since October 1989

Dept - Department number of employee at the time of monitoring. 

Badge - Employee identification number.

Sample Number – Laboratory sample identifier.

Sample Date - Date sample was voided.

Sample Type 

A = Alpha spec 

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.)

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 

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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 employee 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 chest wall thickness 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.

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.

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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 exam)

02 = Thorium (second or later exam)

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 daughter isotopes detected in chest cavity.

U-238 (mg) – 234Th activity detected in chest cavity.

Tc-99 (μCi) - 99Tc activity detected in chest cavity.

Current Dept – Additional instance of employee 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 clearance 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. 669). 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). Based on this study, both depleted and enriched uranium 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 necessity 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 

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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). 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 detection limits LD 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) Convention 

Fluorometry 1950–9/1989 7 μg/d 11 dpm/d

Gross alpha 1950–1964 47 dpm/d LD

Gross alpha 1965–9/1989 26 dpm/d LD

Alpha spectrometry 10/1989–present 0.15 dpm/d LD

Table A-2. Other in vitro detection limits. 

Analyte/solubility type Method Period

Detection limit 

(activity)

Tritium (HTO) Liquid scintillation 10/1988–present 2,000 dpm/da

Isotopic plutonium/M, S Alpha spectrometry 10/1989–present 0.025 dpm/sample

Am-241/M Alpha spectrometry 10/1989–present 0.050 dpm/sample

Th-228/M, S Alpha spectrometry 10/1989–present 0.150 dpm/sample

Th-232/M, S Alpha spectrometry 10/1989–present 0.070 dpm/sample

Np-237/M Alpha spectrometry 10/1989–present 0.100 dpm/sample

a. Estimate.

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. 

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. 

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. 

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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 13.5 mg, respectively, through 1990. Beginning 

in 1991, the critical level or MDA is supplied with the individual results.

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

Radionuclide/solubility

type

Reported limit 

(nCi)

Reported limit 

(dpm)

Citation 

year

Th-232/M, S 0.6a 1,300 1965

Np-237/M 2.7b 5,900 1961

Np-237/M; Pa-233/M, S 0.255 b 560 1961

Co-60/M, S 0.66 b 1,500 1961

Zr-95/M, S; Nb-95/M, S 1.0 b 2,200 1961

a. West (1965, p. 26)

b. Cofield (1961) 

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 

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, 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). 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

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TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Purpose ................................................................................................................................. 55

B.2 Overview................................................................................................................................ 55

B.3 Data ....................................................................................................................................... 55

B.3.1 Selected Bioassay Data.............................................................................................. 55

B.3.2 Analysis...................................................................................................................... 55

B.4 Intake Modeling...................................................................................................................... 61

B.4.1 Assumptions............................................................................................................... 61

B.4.2 Bioassay Data Fitting.................................................................................................. 62

B.5 Uranium Intake Rates by Material Type ................................................................................. 62

B.5.1 Type F........................................................................................................................ 63

B.5.2 Type M ....................................................................................................................... 63

B.5.3 Type S........................................................................................................................ 64

B.6 Assignment of Intakes and Doses .......................................................................................... 64

B.6.1 Intake Rate Summary................................................................................................. 64

B.6.2 Contribution from Contaminants in Recycled Uranium................................................ 64

B.6.3 Dose Assignment ....................................................................................................... 64

LIST OF TABLES

TABLE TITLE PAGE

B-1 Monthly uranium 24-hour urinary excretion rate analyses, 1952 to 1988................................ 56

B-2 Type F uranium intake periods and rates ............................................................................... 63

B-3 Type M uranium intake periods and rates .............................................................................. 63

B-4 Type S uranium intake periods and rates ............................................................................... 64

B-5 Type F uranium intake periods and rates ............................................................................... 65

B-6 Type M uranium intake periods and rates .............................................................................. 65

B-7 Type S uranium intake periods and rates ............................................................................... 66

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LIST OF FIGURES 

FIGURE TITLE PAGE

B-1 Assumed uranium intake, 1/1/47 to 12/31/1969, 50th-percentile results, type F..................... 66

B-2 Assumed uranium intake, 1/1/70 to 12/31/1988, 50th-percentile results, type F..................... 67

B-3 Assumed uranium intake, 1/1/47 to 2/28/78, 50th-percentile results, type M .......................... 68

B-4 Assumed uranium intake, 1/1/47 to 4/30/52, 50th-percentile results, type M .......................... 68

B-5 Assumed uranium intake, 8/1/53 to 11/30/53, 50th-percentile results, type M ........................ 69

B-6 Assumed uranium intake, 11/1/56 to 2/28/59, 50th-percentile results, type M ........................ 69

B-7 Assumed uranium intake, 6/1/61 to 12/31/62, 50th-percentile results, type M ........................ 70

B-8 Assumed uranium intake, 10/1/68 to 12/31/72, 50th-percentile results, type M ...................... 70

B-9 Assumed uranium intake, 3/1/78 to 9/30/84, 50th-percentile results, type M .......................... 71

B-10 Assumed uranium intake, 10/1/84 to 12/31/88, 50th-percentile results, type M ...................... 71

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

percentile, 1952 to 1988......................................................................................................... 72

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

percentile, 1952 to 1988......................................................................................................... 73

B-13 Assumed uranium intake, 1/1/47 to 2/28/78, 50th-percentile results, type S........................... 74

B-14 Assumed uranium intake, 1/1/47 to 4/30/52, 50th-percentile results, type S........................... 75

B-15 Assumed uranium intake, 8/1/53 to 11/30/53, 50th-percentile results, type S......................... 75

B-16 Assumed uranium intake, 11/1/56 to 2/28/59, 50th-percentile results, type S......................... 76

B-17 Assumed uranium intake, 6/1/61 to 12/31/62, 50th-percentile results, type S......................... 76

B-18 Assumed uranium intake, 10/1/68 to 12/31/72, 50th-percentile results, type S....................... 77

B-19 Assumed uranium intake, 3/1/78 to 9/30/84, 50th-percentile results, type S........................... 77

B-20 Assumed uranium intake, 10/1/84 to 12/31/88, type S ........................................................... 78

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

percentile, 1952 to 1988......................................................................................................... 78

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

percentile, 1952 to 1988......................................................................................................... 79

ATTACHMENT B

INTERNAL DOSIMETRY COWORKER DATA FOR Y-12

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B.1 PURPOSE

There are instances of energy employees 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 coworker data from the Y-12 Plant.

B.2 OVERVIEW

ORAUT-OTIB-0019, Analysis of Coworker Bioassay Data for Internal Dose Assignment (ORAUT 

2005) describes the general process for analysis of bioassay data for assignment of doses to 

individuals based on coworker 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 2005). The resultant values were input to the 

Integrated Modules for Bioassay Analysis (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 [1]. 

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 (2005). 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 may underestimate the urinary uranium concentrations as discussed in Section 

5.3.2.1. January through April 1952 were included in the modeling because the 50th- and 84thATTACHMENT B

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percentile values were generally larger than subsequent values. Table B-1 lists the statistical analysis 

results.

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

Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

1/15/1952 21.16 64 696 7/16/1970 8.88 38.05 399

2/14/1952 24.92 84.51 741 8/15/1970 8.76 42.72 401

3/15/1952 23.86 88.16 918 9/15/1970 11.59 55.4 236

4/15/1952 25.49 74.15 665 10/15/1970 9.81 34.23 500

5/15/1952 21.15 51.31 1103 11/14/1970 13.1 51.16 460

6/15/1952 18.06 47.31 1401 12/15/1970 17.14 67.22 275

7/15/1952 14.13 41.9 1236 1/14/1971 4.41 18.55 12

8/15/1952 11.25 33.06 1258 2/14/1971 11.3 35.27 523

9/14/1952 13.7 35.97 1590 3/16/1971 12.76 50.18 334

10/14/1952 16.71 46.49 1305 4/16/1971 9.52 39.99 458

11/14/1952 16.68 46.41 1133 5/16/1971 5.87 21.42 557

12/14/1952 14.29 37.68 1368 6/15/1971 10.37 41.16 263

1/14/1953 12.61 37.15 1123 7/16/1971 4.67 19.49 396

2/13/1953 10.51 32.28 1065 8/15/1971 10.74 47.49 557

3/16/1953 9.18 26.4 1380 9/15/1971 13.95 62.06 275

4/15/1953 10.34 32.25 1095 10/15/1971 8.25 32.76 463

5/16/1953 10.43 33.25 1079 11/15/1971 8.24 29.09 504

6/15/1953 9.81 34.9 1337 12/15/1971 11.67 35.91 283

7/15/1953 11.13 40.08 830 1/15/1972 11.54 44.77 289

8/15/1953 9.81 33.3 1261 2/14/1972 12.2 43.92 227

9/14/1953 19.53 63.39 1099 3/15/1972 15.64 56.16 253

10/15/1953 32.43 110.13 677 4/15/1972 12.09 47.06 246

11/14/1953 26.65 88.11 1752 5/15/1972 12.64 50.99 296

12/15/1953 18.47 57.43 1614 6/15/1972 13.68 34.9 155

1/14/1954 19.45 84.37 1655 7/15/1972 18.95 70.38 257

2/13/1954 16.06 74.2 1938 8/15/1972 20.33 67.87 314

3/16/1954 19.5 74.74 2233 9/14/1972 18.14 54.91 234

4/15/1954 22.04 78.72 1688 10/14/1972 15.48 64.45 225

5/16/1954 22.85 77.01 1774 11/14/1972 12 42.03 369

6/15/1954 22.07 73.11 1801 12/14/1972 20.42 61.5 164

7/16/1954 22.06 83.01 1303 1/14/1973 15.63 41.34 347

8/15/1954 15.39 63.49 2571 2/13/1973 15.79 57.05 271

9/15/1954 8.47 28.08 1064 3/16/1973 15.71 49.06 325

10/15/1954 11.58 50.14 1726 4/15/1973 16.2 48.06 268

11/14/1954 22.02 91.14 118 5/16/1973 14.46 47.02 366

12/15/1954 14.85 45.87 783 6/15/1973 15.04 56.06 248

1/14/1955 7.36 25.15 591 7/15/1973 14.38 43.07 326

2/14/1955 5.62 19.81 639 8/15/1973 16.5 52.5 296

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Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

3/16/1955 13.71 39.83 909 9/14/1973 12.97 56.63 509

4/16/1955 16.2 50.07 1840 10/15/1973 12.31 40.96 485

5/16/1955 13.27 37.94 2006 11/14/1973 10.62 29.86 486

6/15/1955 15.71 47.63 1595 12/15/1973 14.2 46.52 393

7/16/1955 13.1 36.57 1573 1/14/1974 12.16 38.96 418

8/15/1955 13.85 42.78 1874 2/13/1974 11.16 34.34 378

9/15/1955 17.1 55.19 1592 3/16/1974 20.33 56.38 287

10/15/1955 16.87 51.49 1990 4/15/1974 12.99 40.27 325

11/15/1955 12.99 39.56 1725 5/16/1974 9.14 22.89 405

12/15/1955 15.34 45.01 1732 6/15/1974 9.75 29.71 222

1/15/1956 14.5 47.71 2121 7/16/1974 7.54 20.57 270

2/14/1956 16.03 48.01 1853 8/15/1974 8.44 24.96 330

3/15/1956 14.02 42.59 1902 9/15/1974 8.77 24.28 264

4/15/1956 16.45 46.84 1886 10/15/1974 14.19 137.02 298

5/15/1956 15.73 47.57 1980 11/14/1974 9.16 53.77 368

6/15/1956 13.79 40.67 1986 12/15/1974 8.11 35.07 250

7/15/1956 13.82 41.39 2068 1/14/1975 8.93 37.19 252

8/15/1956 13.39 41.44 843 2/14/1975 11.49 44.05 360

9/14/1956 14.11 38.28 1637 3/16/1975 16.92 57.92 291

10/14/1956 11.1 32.33 2022 4/16/1975 10.37 46.11 319

11/14/1956 4.72 15.14 1835 5/16/1975 11.74 39.05 297

12/14/1956 7.3 26.34 1889 6/15/1975 10.66 33.4 186

1/14/1957 9.42 28.04 2413 7/16/1975 11.04 38.53 188

2/13/1957 10.2 38.81 1967 8/15/1975 11.07 32.85 338

3/16/1957 11.04 40.78 1951 9/15/1975 8.9 31.58 171

4/15/1957 10.79 33.8 2302 10/15/1975 11.92 36.58 186

5/16/1957 10.48 42.11 1975 11/15/1975 9.18 32.82 298

6/15/1957 9.44 34.06 2099 12/15/1975 12.88 40.99 148

7/15/1957 8.46 28.63 2471 1/15/1976 9.97 30.51 196

8/15/1957 10.56 35.75 2038 2/14/1976 10.81 32.89 321

9/14/1957 7.51 24.91 2552 3/15/1976 9.19 28.35 225

10/15/1957 12.06 53.49 2278 4/15/1976 8.65 27.03 211

11/14/1957 13.03 44.83 2387 5/15/1976 8.3 24.08 347

12/15/1957 11.29 37.96 2878 6/15/1976 8.24 23.91 185

1/14/1958 8.36 30.87 2708 7/15/1976 3.82 12.47 194

2/13/1958 13.64 53.23 2208 8/15/1976 6.51 18.64 332

3/16/1958 16.24 59.24 2219 9/14/1976 5.39 14.99 193

4/15/1958 13.3 52.29 2498 10/14/1976 9.5 26.06 228

5/16/1958 12.71 56.35 2396 11/14/1976 13.24 40.32 297

6/15/1958 13.7 77.06 2665 12/14/1976 13.72 43.42 202

ATTACHMENT B

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Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

7/16/1958 17.44 79.66 2332 1/14/1977 13.97 44.66 169

8/15/1958 19.45 65.95 2206 2/13/1977 13.41 41.7 268

9/15/1958 19.19 86.52 2669 3/16/1977 12.91 40.62 245

10/15/1958 15.21 58.52 2115 4/15/1977 8.86 24.33 191

11/14/1958 21.01 73.93 2258 5/16/1977 8.39 22.81 302

12/15/1958 22.76 80.61 3071 6/15/1977 9.62 31.33 180

1/14/1959 20.74 77.72 2949 7/15/1977 7.46 28.02 190

2/14/1959 24.9 90.03 2798 8/15/1977 9.29 25.59 342

3/16/1959 18.71 60.78 2424 9/14/1977 11.22 54.94 207

4/16/1959 19.72 66.84 3126 10/15/1977 10.86 37.43 244

5/16/1959 17.44 56.57 2725 11/14/1977 10.75 47.78 341

6/15/1959 12.41 45.55 2878 12/15/1977 9.31 35.7 151

7/16/1959 16.69 57.44 2520 1/14/1978 9.28 35.13 184

8/15/1959 12.79 40.15 3001 2/13/1978 9.82 29.72 293

9/15/1959 13.66 43.44 2580 3/16/1978 4.34 13.13 223

10/15/1959 11.53 40.87 2983 4/15/1978 8.07 26.41 196

11/15/1959 14.86 53.07 3317 5/16/1978 5.53 18.15 291

12/15/1959 16.04 56.6 2785 6/15/1978 4.73 14.79 164

1/15/1960 13.87 43.69 3395 7/16/1978 6.58 20.43 185

2/14/1960 13.36 46 3666 8/15/1978 5.1 24.98 362

3/15/1960 9.31 34.17 2662 9/15/1978 7.02 22.71 198

4/15/1960 9.75 39.57 3375 10/15/1978 7.51 22.85 205

5/15/1960 11.19 38.09 3796 11/14/1978 6.81 24.14 390

6/15/1960 9.88 37.26 2468 12/15/1978 4.93 15.71 155

7/15/1960 10.51 37.64 2330 1/14/1979 5.48 22.99 180

8/15/1960 12.76 51.09 3115 2/14/1979 5.73 24.87 398

9/14/1960 9.45 32.2 2392 3/16/1979 7.5 55.07 219

10/14/1960 8.93 32.44 3155 4/16/1979 4.12 20.11 213

11/14/1960 9.85 35.5 2486 5/16/1979 5.45 26.64 357

12/14/1960 7.6 27.11 2443 6/15/1979 6.54 24.51 198

1/14/1961 7.53 32.63 2571 7/16/1979 6.15 26.52 205

2/13/1961 6.99 27.55 2409 8/15/1979 2.81 8.47 407

3/16/1961 9.36 37.96 2397 9/15/1979 7.52 23.8 180

4/15/1961 8.94 32.38 754 10/15/1979 6.69 21.68 205

5/16/1961 7.46 26.83 1556 11/15/1979 5.68 16.34 360

6/15/1961 9.17 32.46 532 12/15/1979 10.15 28.45 219

7/15/1961 11.04 38.61 1035 1/15/1980 5.43 19.51 238

8/15/1961 5.8 23.46 1137 2/14/1980 5.39 18.48 462

9/14/1961 12.96 54.55 621 3/15/1980 9.96 31.5 206

10/15/1961 8.55 29.21 1216 4/15/1980 7.03 28.92 273

ATTACHMENT B

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Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

11/14/1961 7.46 28.68 1222 5/15/1980 6.08 23.84 473

12/15/1961 16.66 64.1 555 6/15/1980 7.52 24.86 215

1/14/1962 11.93 48.51 1303 7/15/1980 4.43 13.44 231

2/13/1962 14.1 55.59 1191 8/15/1980 4.33 14.91 488

3/16/1962 22.09 81.26 830 9/14/1980 5.59 16.59 237

4/15/1962 16.35 63.01 1431 10/14/1980 8.8 27.7 253

5/16/1962 10.69 39.93 1462 11/14/1980 3.7 14.66 468

6/15/1962 19.41 71.41 1045 12/14/1980 6.65 26.31 255

7/16/1962 13.57 52.64 1528 1/14/1981 6.78 27.85 305

8/15/1962 12.66 58.77 1420 2/13/1981 5.54 25.15 462

9/15/1962 19.24 72.07 1049 3/16/1981 6.2 20.98 325

10/15/1962 15.98 53.81 1524 4/15/1981 4.81 19.11 269

11/14/1962 16.91 65.36 1470 5/16/1981 3.66 15.89 535

12/15/1962 22.87 97.5 1134 6/15/1981 5.97 21.83 251

1/14/1963 16.8 64.65 1222 7/15/1981 6.47 31.68 26

2/14/1963 14.35 65.28 1307 8/15/1981 4.89 16.79 575

3/16/1963 17.58 70.17 1203 9/14/1981 6.13 22.23 315

4/16/1963 13.52 46.46 1481 10/15/1981 5.14 21.72 314

5/16/1963 12.56 44 1661 11/14/1981 4.35 18.28 605

6/15/1963 14.88 56 1038 12/15/1981 5.73 29.53 294

7/16/1963 13.4 36.41 1321 1/14/1982 4.53 18.34 351

8/15/1963 13.29 43.99 902 2/13/1982 4.13 15.44 540

9/15/1963 13.48 44.18 1247 3/16/1982 5.01 18.71 482

10/15/1963 11.65 38.88 576 4/15/1982 5.72 32.11 323

11/15/1963 9.5 26.83 1002 5/16/1982 6.89 24.15 584

12/15/1963 7.73 26.04 1562 6/15/1982 5.47 20.5 366

1/15/1964 8 28.09 1390 7/16/1982 4.66 17.11 320

2/14/1964 7.96 24.14 1579 8/15/1982 6.48 31.75 581

3/15/1964 10.27 32.95 1172 9/15/1982 2.92 10.86 351

4/15/1964 4.6 15.35 902 10/15/1982 6.66 25.77 332

5/15/1964 3.58 10.53 1203 11/14/1982 3.24 12.97 637

6/15/1964 5.7 21.01 599 12/15/1982 2.94 13.34 308

7/15/1964 5.66 18.62 903 1/14/1983 3.67 15.42 348

8/15/1964 4.65 14.2 1088 2/14/1983 5.93 23.87 634

9/14/1964 4.62 12.61 477 3/16/1983 5.51 26.4 316

10/14/1964 3.43 9.95 667 4/16/1983 4.51 17.99 296

11/14/1964 3.26 10.78 1112 5/16/1983 4.83 19.89 633

12/14/1964 7.72 21.81 291 6/15/1983 2.99 12.46 288

1/14/1965 4.7 14.91 545 7/16/1983 2.74 9.92 271

2/13/1965 3.99 12.84 646 8/15/1983 5.17 19.39 690

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Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

3/16/1965 6.81 23.26 404 10/15/1983 3.36 16.08 279

4/15/1965 4.59 21.09 312 11/15/1983 2.69 14.1 706

5/16/1965 6.7 23.6 571 12/15/1983 2.08 8.98 325

6/15/1965 9.45 31.48 329 1/15/1984 2.72 11 326

7/15/1965 5.64 18.05 364 2/14/1984 2.59 14.01 653

8/15/1965 6.41 23.41 609 3/15/1984 4.2 20.09 329

9/14/1965 2.74 11.57 376 4/15/1984 4.14 21.85 360

10/15/1965 4.91 18.07 416 5/15/1984 3.51 15.94 637

11/14/1965 3.44 12.05 571 6/15/1984 3.42 15.07 335

12/15/1965 9.85 37.35 397 7/15/1984 3.49 15.65 352

1/14/1966 4.44 19.14 469 8/15/1984 2.96 17.55 697

2/13/1966 4.41 24.7 668 9/14/1984 3.97 20.81 356

3/16/1966 10.17 42.75 391 10/14/1984 6.23 26.17 367

4/15/1966 10.75 48.01 441 11/14/1984 2.68 8.13 644

5/16/1966 5.04 16.99 739 12/14/1984 4.62 20.08 348

6/15/1966 6.32 26.44 415 1/14/1985 2.93 12 397

7/16/1966 8.7 39.61 359 2/13/1985 1.93 8.23 605

8/15/1966 5.01 18.31 624 3/16/1985 4.45 31.7 425

9/15/1966 8.86 40.76 431 4/15/1985 1.67 8.6 391

10/15/1966 15.88 71.36 392 5/16/1985 1.66 8.97 625

11/14/1966 8.68 32.4 539 6/14/1985 0.76 5.03 318

12/15/1966 8.98 40.16 567 7/15/1985 1.91 8.85 383

1/14/1967 6.79 33.34 339 8/14/1985 1.97 7.66 569

2/14/1967 5.8 25.69 710 9/14/1985 2.76 11.77 369

3/16/1967 8.35 49.25 426 10/14/1985 1.82 6.05 419

4/16/1967 10.81 56.82 364 11/13/1985 1.53 6.04 544

5/16/1967 3.25 18.88 593 12/14/1985 2.71 11.36 379

6/15/1967 6.51 25.46 224 1/13/1986 2.79 11.56 472

7/16/1967 5.49 35.82 390 2/13/1986 1.1 5.61 593

8/15/1967 7.3 40.93 585 3/14/1986 2.41 13.98 388

9/15/1967 4.42 19.34 401 4/15/1986 1.35 5.24 497

10/15/1967 5.21 20.92 412 5/15/1986 1.12 5.13 555

11/15/1967 5.01 20.71 514 6/15/1986 1.63 7.52 401

12/15/1967 6.06 33.53 405 7/15/1986 2.42 11.67 408

1/15/1968 6.5 28.54 393 8/14/1986 2.54 10.13 572

2/14/1968 10.15 41.09 573 9/14/1986 1.79 8.35 352

3/15/1968 8.1 39.56 476 10/14/1986 0.95 4.22 550

4/15/1968 7.07 43.78 411 11/14/1986 2.6 11.31 626

5/15/1968 11.07 53.86 458 12/14/1986 1.17 6.57 323

6/15/1968 7.05 29.75 574 1/14/1987 0.8 8.03 437

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Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

Effective 

sample date

50th 

percentile 

(dpm/d)

84th 

percentile 

(dpm/d)

Number of 

samples

7/15/1968 5.2 23.25 395 2/13/1987 9.62 57.02 555

8/15/1968 5.86 23.48 533 3/15/1987 5.04 18.9 345

9/14/1968 5.48 26.46 421 4/15/1987 2.5 13.62 404

10/14/1968 3.86 16.16 391 5/15/1987 2.43 11.68 525

11/14/1968 5.9 23.37 516 6/15/1987 2.2 19.91 328

12/14/1968 9.71 32.61 533 7/15/1987 8.16 46.96 34

1/14/1969 8.43 41.12 220 8/15/1987 13.59 70.44 24

2/13/1969 5.02 20.69 731 9/14/1987 4.01 19.3 54

3/16/1969 5.65 27.79 655 10/15/1987 0.44 2.13 474

4/15/1969 7.61 36.08 477 11/14/1987 1.78 14.61 494

5/16/1969 5.22 24.92 593 12/14/1987 2.13 11.13 327

6/15/1969 8.26 30.14 546 1/14/1988 2.19 14.84 338

7/15/1969 7.37 29.64 375 2/13/1988 2.37 10.64 550

8/15/1969 6.62 28.37 630 3/15/1988 4.62 24.43 371

9/14/1969 4.7 17.88 475 4/14/1988 4.03 26.36 345

10/15/1969 6.6 25.36 514 5/15/1988 1.52 7.09 606

11/14/1969 7.06 22.98 541 6/14/1988 2.05 13.32 363

12/15/1969 7.71 30.81 401 7/14/1988 4.03 19.34 416

1/14/1970 9.84 40.81 336 8/14/1988 2.96 17.83 563

2/13/1970 8.33 28.27 559 9/14/1988 2.38 14.04 432

3/16/1970 7.85 29.62 316 10/14/1988 0.75 3.7 434

4/15/1970 6.99 27.21 378 11/13/1988 2.18 10.51 103

5/16/1970 4.93 18.99 103 12/14/1988 3.69 13.94 152

6/15/1970 11.17 44.73 137

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 dose conversion factors) for 234U that are 

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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 dose conversion factor 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 than in later years, so 

this period was not included in the modeling. The ORAUT-TKBS-0014-2, Y-12 National Security 

Complex – Site Description (ORAUT 2007) 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. 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 [2,

3]. 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 time 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. 

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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.

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

Start Stop

Uranium intake rate (dpm/d)

50th percentile 84th percentile GSD

1/1/1947 5/31/1952 85.35 264.1 3.09

6/1/1952 8/31/1953 43.89 131.3 2.99

9/1/1953 10/31/1953 104.3 339.5 3.26

11/1/1953 12/31/1954 66.43 254.1 3.83

1/1/1955 12/31/1956 47.03 145.2 3.09

1/1/1957 12/31/1957 36.07 145.6 4.04

1/1/1958 12/31/1959 59.48 207.8 3.49

1/1/1960 12/31/1961 34.68 134 3.86

1/1/1962 12/31/1963 53.27 171.9 3.23

1/1/1964 12/31/1969 22.97 97.28 4.24

1/1/1970 12/31/1971 35.8 115.6 3.23

1/1/1972 12/31/1975 47.51 168.4 3.54

1/1/1976 12/31/1977 35.09 121.7 3.47

1/1/1978 12/31/1983 19.68 85.61 4.35

1/1/1984 12/31/1988 9.098 51.26 5.63

7/1/1987 8/31/1987 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 geometric standard deviations (GSDs).

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

Start Stop

Uranium intake rate (dpm/d)

50th percentile 84th percentile GSD

1/1/1947 2/28/1978 169.34 598.93 3.54

1/1/1947 4/30/1952 354.59 1,154.9 3.26

8/1/1953 11/30/1953 547.22 1,963.1 3.59

11/1/1956 2/28/1959 226.90 825.31 3.64

6/1/1961 12/31/1962 248.68 866.81 3.49

10/1/1968 12/31/1972 160.67 601.60 3.74

3/1/1978 9/30/1984 80.03 355.93 4.45

10/1/1984 12/31/1988 44.203 223.85 5.06

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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.

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

Start Stop

Uranium intake rate (dpm/d)

50th percentile 84th percentile GSD

1/1/1947 2/28/1978 1,844.4 6,544.2 3.55

1/1/1947 4/30/1952 5,210.4 16,970 3.26

8/1/1953 11/30/1953 17,983 64,559 3.59

11/1/1956 2/28/1959 5,694.6 20,560 3.61

6/1/1961 12/31/1962 6,849.4 23,632 3.45

10/1/1968 12/31/1972 3,290.3 12,415 3.77

3/1/1978 9/30/1984 1,280.3 5,802.9 4.53

10/1/1984 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 Table 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.

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 recycled uranium, 

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

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 

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on 50th-percentile intake rates are assigned as lognormal distribution with the applicable GSD. There 

are situations when the 95th percentile of the coworker 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. 

Table B-5. Type F uranium intake periods and rates, dpm/day. 

Start Stop

50th 

percentile GSD

95th

percentile

1/1/1947 5/31/1952 85.4 3.54 683

6/1/1952 8/31/1953 43.9 3.54 351

9/1/1953 10/31/1953 104 3.54 834

11/1/1953 12/31/1954 66.4 4.24 715

1/1/1955 12/31/1956 47.0 3.54 376

1/1/1957 12/31/1957 36.1 4.24 388

1/1/1958 12/31/1959 59.5 3.54 476

1/1/1960 12/31/1961 34.7 4.24 373

1/1/1962 12/31/1963 53.3 3.54 426

1/1/1964 12/31/1969 23.0 4.24 247

1/1/1970 12/31/1971 35.8 3.54 286

1/1/1972 12/31/1975 47.5 3.54 380

1/1/1976 12/31/1977 35.1 3.54 281

1/1/1978 12/31/1983 19.7 4.35 221

1/1/1984 6/30/1987 9.10 5.63 156

7/1/1987 8/31/1987 43.5 3.00 265

9/1/1987 12/31/1988 9.10 3.54 156

Table B-6. Type M uranium intake periods and rates, dpm/day . 

Start Stop

50th 

percentile GSD

95th

percentile

1/1/1947 4/30/1952 524 3.74 4,588

5/1/1952 7/31/1953 169 3.74 1,483

8/1/1953 11/30/1953 717 3.74 6,275

12/1/1953 10/31/1956 169 3.74 1,483

11/1/1956 2/28/1959 396 3.74 3,470

3/1/1959 5/31/1961 169 3.74 1,483

6/1/1961 12/31/1962 418 3.74 3,661

1/1/1963 9/30/1968 169 3.74 1,483

10/1/1968 12/31/1972 330 3.74 2,890

1/1/1973 2/28/1978 169 3.74 1,483

3/1/1978 9/30/1984 80.0 4.45 933

10/1/1984 12/31/1988 44.2 5.06 636

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Table B-7. Type S uranium intake periods and rates, dpm/day . 

Start Stop

50th 

percentile GSD

95th

percentile

1/1/1947 4/30/1952 7,054 3.77 62,595

5/1/1952 7/31/1953 1,844 3.77 16,366

8/1/1953 11/30/1953 19,827 3.77 175,929

12/1/1953 10/31/1956 1,844 3.77 16,366

11/1/1956 2/28/1959 7,539 3.77 66,895

3/1/1959 5/31/1961 1,844 3.77 16,366

6/1/1961 12/31/1962 8,693 3.77 77,138

1/1/1963 9/30/1968 1,844 3.77 16,366

10/1/1968 12/31/1972 5,134 3.77 45,559

1/1/1973 2/28/1978 1,844 3.77 16,366

3/1/1978 9/30/1984 1,280 4.53 15,367

10/1/1984 12/31/1988 885 4.91 12,125

Recycled uranium contaminants, when appropriate for the period, also factor into this comparison.

The lognormal distribution is selected in the Interactive RadioEpidemiological Program (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, 1/1/47 to 12/31/1969, 50th-percentile 

results, type F. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-2. Assumed uranium intake, 1/1/70 to 12/31/1988, 50th-percentile 

results, type F. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-3. Assumed uranium intake, 1/1/47 to 2/28/78, 50th-percentile 

results, type M. 

Figure B-4. Assumed uranium intake, 1/1/47 to 4/30/52, 50th-percentile results,

type M. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-5. Assumed uranium intake, 8/1/53 to 11/30/53, 50th-percentile

results, type M. 

Figure B-6. Assumed uranium intake, 11/1/56 to 2/28/59, 50th-percentile

results, type M. 

O monthly 

bioassay value

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-7. Assumed uranium intake, 6/1/61 to 12/31/62, 50th-percentile 

results, type M. 

Figure B-8. Assumed uranium intake, 10/1/68 to 12/31/72, 50th-percentile 

results, type M. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-9. Assumed uranium intake, 3/1/78 to 9/30/84, 50th-percentile 

results, type M. 

Figure B-10. Assumed uranium intake, 10/1/84 to 12/31/88, 50th-percentile

results, type M. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-11. Predicted (type M, 5 μm) uranium urinary excretion from eight

independent intakes, 50th percentile, 1952 to 1988. 

O monthly 

bioassay value 

(50th percentile)

___ predicted 

excretion

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Figure B-12. Predicted (type M, 5 μm) uranium urinary excretion from eight

independent intakes, 84th percentile, 1952 to 1988. 

O monthly 

bioassay value 

(84th percentile)

___ predicted 

excretion

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Figure B-13. Assumed uranium intake, 1/1/47 to 2/28/78, 50th-percentile

results, type S. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-14. Assumed uranium intake, 1/1/47 to 4/30/52, 50th-percentile

results, type S. 

Figure B-15. Assumed uranium intake, 8/1/53 to 11/30/53, 50th-percentile

results, type S. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-16. Assumed uranium intake, 11/1/56 to 2/28/59, 50th-percentile 

results, type S.

Figure B-17. Assumed uranium intake, 6/1/61 to 12/31/62, 50th-percentile

results, type S. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-18. Assumed uranium intake, 10/1/68 to 12/31/72, 50th-percentile 

results, type S. 

Figure B-19. Assumed uranium intake, 3/1/78 to 9/30/84, 50th-percentile

results, type S. 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit 

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-20. Assumed uranium intake, 10/1/84 to 12/31/88, type S. 

Figure B-21. Predicted (type S, 5 μm) uranium urinary excretion from eight

independent intakes, 50th percentile, 1952 to 1988. 

O monthly 

bioassay value 

(50th percentile)

___ predicted 

excretion

O monthly 

bioassay value 

(50th percentile)

O not used in fit

___ fit

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Figure B-22. Predicted (type S, 5 μm) uranium urinary excretion from eight

independent intakes, 84th percentile, 1952 to 1988. 

O monthly 

bioassay value 

(84th percentile)

___ predicted 

excretion

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