ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 49

DOE Review Release 11/24/2014

Pantex Plant – Occupational Internal Dose

ORAUT-TKBS-0013-5 Rev. 03

Effective Date: 11/13/2014

Supersedes: Revision 02

Subject Expert(s): Dale D. Thomas III

Document Owner 

Approval: Signature on File Approval Date: 10/29/2014

Dale D. Thomas III, Document Owner

Concurrence: Signature on File Concurrence Date: 11/03/2014

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 11/01/2014

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 10/29/2014

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 11/13/2014

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-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 2 of 49

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

07/27/2006 00 New Technical Basis Document for the Pantex Plant – Occupational 

Internal Dose. First approved issue. Training required as 

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

Signature on File 05/29/2006

Jack J. Fix, Document Owner

Signature on File 04/26/2006

John M. Byrne, Document Owner

Signature on File 05/26/2006

Edward F. Maher, Task 5 Manager

Signature on File 05/31/2006

Kate Kimpan, Project Director

Signature on File 07/27/2006

James W. Neton, Associate Director for Science

02/08/2007 00 PC-1 Page change revision to change the date for the first DU-laden 

hydroshot from 1952 to 1959 on page 31. Updated the text in 

Section 5.2.2.6 on page 28 and updated Table 5-11 on page 32 and 

5-19 on page 48. Added reference on page 49 for the telephone 

conversation that confirmed the actual date. Table 5-1 was moved 

from page 9 to page 10. Results in a decrease in the intake of DU 

and resultant dose for the years 1952 to 1959. Text was modified or 

added in Section 5 and References (pages 9-12, 28, 31-32, 48-49, 

and 53). No sections were deleted. Incorporates internal formal 

review comments and NIOSH formal review comments. This 

revision results in a reduction in assigned dose and no PER is 

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

Initiated by Donald E. Bihl. Approval:

Donald E. Bihl Signature on File for 01/24/2007

Jack J. Fix, Document Owner

Signature on File 01/25/2007

John M. Byrne, Document Owner

Signature on File 01/26/2007

Edward F. Maher, Task 5 Manager

Signature on File 01/31/2007

Kate Kimpan, Project Director

Brant A. Ulsh Signature on File for 02/08/2007

James W. Neton, Associate Director for Science
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 3 of 49

06/22/2007 01 Approved revision initiated to include Attributions and Annotations 

and additional references. No sections were deleted. Minor editorial 

changes were made in Sections 5.2.1, 5.2.1.1, 5.2.1.1.4, 5.2.2.5, and 

5.2.2.6.2. Constitutes a total rewrite of the document. Incorporates 

formal internal and NIOSH 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 Donald 

E. Bihl. 

01/03/2014 02 Revision initiated to add section detailing the Special Exposure 

Cohort classes and limitations for internal dose assessment from 

intakes before 1991. Sections 5.2.1 through 5.2.3 were revised to 

reflect assessment of internal radiation dose only if monitoring data 

are available for all periods after the end of the SEC periods (1991). 

Incorporates formal internal and NIOSH review comments. 

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

determined by the Objective Manager. Initiated by Dale D. Thomas 

III.

11/13/2014 03 Revision initiated to correct unmonitored tritium doses in Table 5-3. 

Incorporates formal internal and NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by Dale 

D. Thomas III.
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 4 of 49

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 6

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

5.1.1 Purpose........................................................................................................................ 9

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

5.1.3 Special Exposure Cohort Petition Information for Pantex.............................................. 9

5.1.3.1 January 1, 1958, through December 31, 1983 ............................................... 9

5.1.3.2 January 1, 1984, through December 31, 1991 ............................................. 10

5.1.4 History of Internal Dosimetry ...................................................................................... 10

5.1.5 Current Internal Dosimetry Practices .......................................................................... 11

5.1.6 Pantex Workers at Other AEC/DOE Facilities............................................................. 14

5.2 Radionuclides with Potential for Internal Dose ....................................................................... 14

5.2.1 Tritium ........................................................................................................................ 14

5.2.1.1 Internal Assessment for Tritium During Routine Operations ......................... 15

5.2.1.1.1 Dose to Uptake ......................................................................... 16

5.2.1.2 Unmonitored Workers, 1956 to 1971............................................................ 16

5.2.1.3 Unmonitored Workers, 1972 to the Present.................................................. 16

5.2.1.4 Tritium Release Event in 1989 ..................................................................... 17

5.2.1.5 Assignment of Tritium Doses for Production Technicians (Assembly 

Operators), RSTs, and Quality Assurance Technicians, 1956 to 

Present ........................................................................................................ 17

5.2.2 Uranium...................................................................................................................... 17

5.2.2.1 Background.................................................................................................. 17

5.2.2.2 Uranium Reporting Levels or Minimum Detectable Activities........................ 19

5.2.2.3 Assessment of Uranium Intake and Internal Dose........................................ 19

5.2.2.4 Thorium........................................................................................................ 20

5.2.2.5 Assessment of Thorium Intake and Internal Dose ........................................ 21

5.2.3 Plutonium ................................................................................................................... 21

5.2.3.1 General Information ..................................................................................... 21

5.2.3.2 Assessment of Plutonium Intake and Internal Dose After 1983 .................... 21

5.3 Assessing Occupational Internal Dose from Elevated Radon................................................. 22

5.3.1 Dose from Radon-222 Progeny .................................................................................. 23

5.3.2 Radon Monitoring at DOE Facilities............................................................................ 23

5.3.3 Underground Buildings ............................................................................................... 23

5.3.4 Radon Concentrations................................................................................................ 24

5.3.5 Working Level-Months................................................................................................ 25

5.4 Attributions and Annotations .................................................................................................. 26

References ........................................................................................................................................ 33

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

ATTACHMENT A, EXAMPLES OF PANTEX SITE BIOASSAY DATA FILES ................................... 39

ATTACHMENT B, RESULTS OF 1990 RADON MEASUREMENTS................................................. 40

ATTACHMENT C, ASSESSING TRITIUM INTAKES ........................................................................ 46
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 5 of 49

LIST OF TABLES

TABLE TITLE PAGE

5-1 Number of workers with bioassay monitoring ......................................................................... 12

5-2 Job titles and descriptions of work with possibility for occupational intake.............................. 13

5-3 Internal dose for missed and unmonitored tritium intakes....................................................... 16

5-4 Uranium dose to workers ....................................................................................................... 19

5-5 History of uranium urinalysis .................................................................................................. 20

5-6 Number of workers on plutonium bioassay, 1991 to 2002 ...................................................... 22

5-7 Activity composition of weapons-grade plutonium mixtures from Hanford .............................. 22

5-8 Summary statistics of 1990 radon measurements.................................................................. 24

5-9 Summary of 1969 radon measurements in Cells 1 to 6 .......................................................... 25

B-1 Results of 137 radon measurements in 1990 ......................................................................... 40

C-1 Tritium urinalysis screening levels, 1991................................................................................ 48

C-2 Potentially missed intake and dose from monthly sampling for chronic intake of tritium, 

1989 to Present......................................................................................................................49

LIST OF FIGURES

FIGURE TITLE PAGE

5-1 Absolute differences between duplicate radon measurements............................................... 24
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 6 of 49

ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

CEDE committed effective dose equivalent

CEP Controls for Environmental Pollution

CFR Code of Federal Regulations

Ci curie

d day

DAC derived air concentration

DHHS U.S. Department of Health and Human Services

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DORMS Dosimetry Records Management System

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

g gram

GSD geometric standard deviation

HE,50 committed effective dose equivalent

HE high explosive

HERS Historical Exposure Records System

hr hour

HTO tritiated water vapor

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

L liter

m meter

mCi millicurie

MDA minimum detectable activity

MeV megaelectron-volt, 1 million electron-volts

mL milliliter

mrem millirem

NDE nondestructive examination

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH-Office of Compensation Analysis and Support Claims Tracking System

ORAU Oak Ridge Associated Universities

PAEC potential alpha energy concentration

PAEE potential alpha energy exposure

pCi picocurie
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 7 of 49

POC probability of causation

RST Radiation Safety Technician

SEC Special Exposure Cohort

SNM special nuclear material

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

U.S.C. United States Code

wk week

WL working level

WLM working level-month

yr year

μCi microcurie

μg microgram

§ section or sections
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 8 of 49

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 historical 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) restrict the “performance of duty” 

referred to in 42 U.S.C. § 7384n(b) 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-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 9 of 49

5.1.1 Purpose

This technical basis document (TBD) addresses intakes of radionuclides associated with weapons 

operations as well as radon exposures, which might have been enhanced due to the unique cell 

design at the Pantex Plant for limiting the consequences of accidents.

5.1.2 Scope

Activities at the Pantex Plant resulted in a limited potential for airborne contamination in bays and 

cells. The principal function in the bays is the assembly and disassembly of nuclear explosives, 

particularly the mechanical portion of operations that includes electrical components and tritium 

reservoirs. Physics package assembly and disassembly, which involved bare high explosives (HE) 

and sealed pits, occurred in the cells. There are 13 cells for assembly and disassembly at the Pantex 

Plant. Operations with radioactive components began in these cells in 1956. Cell 1 is no longer in 

use because of an accidental tritium gas release in 1989 [1].

Internal Dosimetry Technical Basis & Quality Assurance Document (BWXT Pantex 2001) implies that 

particle size measurements could have been made for specific incidents but had not been performed 

(at that time) for routine airborne contamination conditions. Data were found for particle size 

measurements assessed for uranium and thorium during the 1990s. It appears that measurements 

were used to determine appropriate radiation protection measures but not for dose assessment. The 

dose reconstructor should use the default 5-μm activity median aerodynamic diameter assumption 

(with the exception of tritium and radon progeny) unless data on specific particle size are available in 

the records and are representative of the intakes being considered (ORAUT 2014a).

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 5.4.

5.1.3 Special Exposure Cohort Petition Information for Pantex 

5.1.3.1 January 1, 1958, through December 31, 1983

On December 21, 2011, the Secretary of the U.S. Department of Health and Human Services (DHHS) 

designated the following class of employees as an addition to the Special Exposure Cohort (SEC) 

(DHHS 2011):

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

contractors and subcontractors who worked at the Pantex Plant in Amarillo, Texas, 

during the period from January 1, 1958 through December 31, 1983, for a number of 

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

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

one or more other classes of employees included in the SEC.

As stated in (DHHS 2011), DHHS finds that it lacks sufficient personnel or area monitoring data, 

source term data, and operational information to support reconstructing internal dose from intakes of 

uranium with sufficient accuracy from January 1, 1958 through December 31, 1983 at the Pantex 

Plant in Amarillo, Texas. Reconstruction of thorium intakes with sufficient accuracy is not feasible for 

all workers during the same period since the proposed method for estimating those intakes depend on 

the reconstruction of uranium intakes. However, reconstruction of doses from radon is feasible based 

on workplace measurements. Plutonium and thorium intakes can be reconstructed for individuals 

who have specific monitoring results for those radionuclides. Tritium doses can be reconstructed 

based on tritium bioassay results from monitored workers. Although DHHS found that it is not 

possible to completely reconstruct internal radiation doses for the proposed class, NIOSH can use 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 10 of 49

any internal monitoring data that might 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 Pantex Plant, during the period from January 1, 1958 

through December 31, 1983, but who do not qualify for inclusion in the SEC, can be performed using 

these data as appropriate to support a partial dose reconstruction.

5.1.3.2 January 1, 1984, through December 31, 1991

On September 30, 2013, the Secretary of DHHS designated the following class of employees as an 

addition to the SEC (DHHS 2013):

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

contractors and subcontractors who worked at the Pantex Plant in Amarillo, Texas, 

during the period from January 1, 1984 through December 31, 1991, for a number of 

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

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

one or more other classes of employees included in the SEC.

As stated in (DHHS 2013), DHHS found it lacks sufficient information to reconstruct internal radiation 

doses adequately for all Pantex Plant employees from intakes of uranium and thorium with sufficient 

accuracy from January 1, 1984, through December 31, 1991, at the Pantex Plant in Amarillo, Texas. 

Specifically, DHHS found that the available monitoring data, as well as available process and source 

term information for the Pantex Plant was inadequate to estimate with sufficient accuracy the internal 

doses from potential exposures to uranium during the period from 1984 through 1990, and to thorium 

from January 1, 1984, through December 31, 1991. However, tritium internal doses can be 

reconstructed for the period based on the available tritium bioassay data. Although DHHS found that 

it is not possible to completely reconstruct internal radiation doses for the proposed class, NIOSH can 

use any internal monitoring data that might 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 Pantex Plant, during the period from January 1, 1984 

through December 31, 1991 but who do not qualify for inclusion in the SEC, can be performed using 

these data as appropriate to support a partial dose reconstruction. 

5.1.4 History of Internal Dosimetry

Between 1952 and 1954, the primary mission at Pantex was to precision-machine HE castings and 

send them to Sandia National Laboratory in Albuquerque, New Mexico, for assembly. From 1956 to 

1958, with the in-flight insertable design, the only nuclear components handled at Pantex were 

depleted uranium (DU) cases and tritium reservoirs; during this time there was no processing of 

nuclear material. Because these DU components were new at the time of assembly, this analysis 

assumed that removable DU oxide contamination on the components was minimal. In similar fashion, 

the potential for significant removable tritium contamination was minimal because the tritium 

reservoirs had to meet rigorous shipping requirements. The only other sources of radiation exposure 

at Pantex during this period were industrial radiography and medical X-rays (ORAUT 2007). 

From 1951 to about 1980, nuclear weapons assembly operations were generally free of 

contamination. Occasional checks for removable contamination usually demonstrated negative 

results, so few precautions were taken in relation to personal protective equipment and clothing. 

There was no evidence of any intakes of radioactive materials by Pantex workers (ORAUT 2007).

There was no routine bioassay program at Pantex before 1972 for uranium, thorium, or plutonium [2]. 

Bioassay was performed for specific events; for instance, bioassay was obtained from workers 

involved in a plutonium contamination event in 1961 and from those involved in decontamination of 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 11 of 49

the facility after the event [3]. A 1967 report that describes an inspection of the radiation protection 

program states that Pantex used air samples and contamination surveys to indicate the need for 

bioassay and did not maintain a routine plutonium or uranium bioassay program (Davis 1967). The 

report further states that Pantex performed about 10 tritium urinalyses a month, and there was no 

indication of personnel exposure. There might have been a small routine tritium program, but the 

research for this analysis found no other information.

The 1991 procedure Analysis of Biological Samples for Uranium, Thorium, and/or Plutonium (MHSMC 

1991a) stated that urinalysis was to be conducted for personnel exposed to 40 derived air 

concentration (DAC)-hr integrated air concentrations as measured by breathing-zone monitors or was 

to be estimated if not specifically monitored. The procedure also stated that, “personnel working in 

potentially contaminated areas shall be entered into the routine bioassay program and shall have a 

routine bioassay for the suspect heavy metal radionuclide performed every 4 to 6 months.” However, 

the routine bioassay program for radionuclides other than tritium was short-lived, occurring mostly in 

1991 and 1992 [4].

The research for this TBD did not reveal the level of air concentrations or other workplace indicators 

that triggered special bioassays before 1991 [5].

In 1989, Pantex contracted with Delphi Groupe to develop the Historical Exposure Records System 

(HERS), an electronic database that contains the best-available personnel dose data. Original 

personnel dose records were reviewed, discrepancies identified and corrected, and data entered in 

the database. This effort reconstructed and included missing, incomplete, and invalid doses. It 

included records from 1957 to 1983, with the exceptions of March 1976 and December 1979 because 

data were not available for those months (BWXT Pantex 2000). HERS reports are available for 

several contamination events that occurred in 1989, but earlier data were not easy to extract from the 

files. The dose records in the worker or claimant files contain the HERS data, but those data do not 

include bioassay data. As of 2007, a review of all worker files in the NIOSH Office of Compensation 

Analysis and Support Claims Tracking System (NOCTS) revealed that only 10% of the files had any 

bioassay data and most of the data were for tritium results. Only 3% of the files had uranium 

bioassay results and all but one of the results were for samples that were collected since 1986. No 

plutonium or thorium bioassay results were found in the NOCTS files. Table 5-1 provides a historical 

perspective of bioassay practices at Pantex from 1972 to 2002. The table lists the number of workers 

monitored for the radionuclides of interest for each year.

In early 1992, several workers’ bioassay samples were performed by Controls for Environmental 

Pollution (CEP), a commercial analytical laboratory. However, there were technical and legal issues 

associated with any analyses performed by the CEP. For this reason, no worker’s samples results 

which indicated that the analysis was performed by the CEP may be used for the assessment of 

intake and internal dose reconstruction (CEP 1992). 

5.1.5 Current Internal Dosimetry Practices

In the late 1980s and early 1990s, several actions resulted in the current internal dosimetry program. 

First, there were new regulations from DOE [Order 5480.11 (DOE 1988), the RadCon Manual (DOE 

1992), and 10 CFR Part 835]; second, a new contractor came to the site; and third, several workplace 

incidents occurred that demonstrated the need to improve the internal dosimetry program [6].

The Pantex Plant radiation protection program uses engineering and administrative controls to 

prevent intakes. However, because of the quantities of tritium, plutonium, uranium, and thorium that 

have been handled at the Plant, there is the possibility of an accidental intake resulting in 100-mrem 

committed effective dose equivalent (CEDE or HE,50) (BWXT Pantex 2001). 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 12 of 49

Table 5-1. Number of workers with bioassay monitoring (BWXT Pantex 2005). 

Year Tritium Uranium Thorium Plutonium

1972 4 0 0 0

1973 1 0 0 0

1974 0 0 0 0

1975 0 0 0 0

1976 463 0 0 0

1977 466 0 0 0

1978 519 0 0 0

1979 712 0 0 0

1980 14 0 0 0

1981 41 0 0 0

1982 5 0 0 0

1983 0 0 0 0

1984 0 0 0 0

1985 17 0 0 0

1986 626 0 0 0

1987 481 0 0 0

1988 499 0 0 0

1989 212 0 0 0

1990 2,341 46 0 0

1991 1,115 431 0 0

1992 879 239 17 12

1993 1,078 90 0 0

1994 1,104 138 4 3

1995 971 37 90 33

1996 940 69 56 17

1997 933 89 13 18

1998 610 12 1 2

1999 554 13 16 1

2000 535 33 9 8

2001 512 65 16 1

2002 511 57 11 10

2003 441 87 25 9

2004 421 109 15 0

According to BWXT Pantex (2001), the purpose of the current internal dosimetry program is to detect 

intakes equal to or greater than 10 mrem HE,50 based on International Commission on Radiological 

Protection (ICRP) Publication 30 dose calculation methodology (ICRP 1982). To meet the 

requirements of 10 CFR 835.402(c), workers who might be likely to have internal HE,50 doses higher 

than 100 mrem participate in the internal dose evaluation program. Pantex maintains routine 

bioassay monitoring programs for tritium to demonstrate compliance with 10 CFR 835.402(c). To 

identify intakes of actinides in a timely manner, the internal dosimetry program is tied closely to the 

quantification of airborne radionuclide concentrations to which workers are exposed. Pantex does not 

have a routine bioassay program for actinides but rather uses occurrence-based bioassay sampling to 

confirm intakes and calculate internal doses (BWXT Pantex 2001). Therefore, with the exception of 

tritium exposure, there is reliance on personal air sampling to determine the need to conduct bioassay 

sampling. At present, bioassay sampling occurs within 2 to 3 days of an assessment of airborne 

exposures exceeding 4 DAC-hr for an individual actinide (BWXT Pantex 2001). For these reasons, 

personnel without specific individual monitoring data listed after the applicable SEC dates were not 

likely to have received occupational intakes greater than those from inhalation of the reported 

maximum environmental airborne concentration(s) at the site.
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 13 of 49

Knowing the job title and a brief description of duties for that title can be helpful in determining the 

correct information to use for assessing dose. Production Technicians (also called Assembly 

Operators) and Radiation Safety Technicians (RSTs) typically had the highest potential for intakes of 

occupational radionuclides. Other workers could have incurred intakes, but the probability of incurring 

an intake was smaller and the magnitude of an intake, if it occurred, would have been smaller [7]. 

Claimant interview files might not state the same job title because the interviewee could have 

described the type of job rather than the job title and because job titles have changed over the years. 

Table 5-2 summarizes job titles, descriptions, and possibilities for intakes. 

Table 5-2. Job titles and descriptions of work with possibility for occupational intake [8]. 

Job title Description of work

Possibility for 

intake

(1 highest)a

Production Technician, Assembler, 

Assembly Operator, Assembly 

Fabrication

Assembles, disassembles, reassembles, 

inspects components. 

1

Quality Assurance Technician I Conducts nondestructive evaluations (NDEs) 

with linear accelerators, X-ray machines, etc.; 

conducts telemetry testing; performs 

confirmatory measurements on components, 

assemblies, containers, etc.

1

Quality Assurance Technician II Performs NDEs, electronic, destructive, 

telemetry, and radiation measurement testing.

1

RST (entry) Performs monitoring and sampling; collects 

samples; assists RST in monitoring personnel.

1a

RST Performs monitoring and sampling; collects 

samples; performs radiation and contamination 

surveys; conducts surveillance work.

1

RST (Senior) Responds to contamination or radiation alarms; 

performs surveillance, monitors radiation 

conditions in workplace.

1

Firing Site Technician Includes hydroshot operators, drivers, anyone 

involved with cleanup of hydroshot 

contamination.

1

Not known, possibly drivers or 

teamsters

Includes burning of HE and cleanup of ash at 

burning ground.

1

Material Handler (pits and cans) Operates material handling/moving equipment; 

transports material; loads and unloads 

materials and containers.

2

Operations Manager, Production 

Supervisor 

Supervises personnel engaged in manufacturing, 

assembly, packaging, material control, etc.

2

Quality Control Inspectors/ 

Auditors

Conducts special audits; different from quality 

assurance technicians.

2

Security, protective force, guard Performs per job title. 2b

Engineer, engineering Performs variety of tasks associated with design, 

testing, procedure development.

2c

Machinist Machining on DU for one weapon design only. 1a

Metrology laboratory staff Performs nonradiological metrology calibrations. Environmental only

Fireman Performs per job title. Environmental only

Computer Programmer, Electronic 

Data Processing Analyst 

Performs computer programming, maintenance. Environmental only

Secretary, Administrator, 

Technical Writer, non-operations 

management, Planner

Performs per job title. Environmental only

Tool and die maker Performs per job title. Environmental only
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 14 of 49

Job title Description of work

Possibility for 

intake

(1 highest)a

Food service Performs tasks associated with operation of 

cafeteria.

Environmental only

Stores Stockman, Clerk, 

Supervisor

Performs tasks associated with general stores. Environmental only

a. Based on actual contact with components or contamination or RSTs assisting potentially contaminated personnel.

b. In general, security personnel had little chance of intakes; however, some small intakes from contamination in cells or 

igloos are possible. The default assumption is to place security personnel in category 2; however, based on other 

information in the file, the dose reconstructors can assign environmental intakes only if they believe the worker’s tasks 

did not involve entry into cells, Gravel Gerties, igloos, or locations with resuspendable contamination [9].

c. Engineering tasks cover a wide range, and most have no potential for intakes. However, some tasks might have 

involved observations during assembly or disassembly work or observations during hydroshots. If the engineer did not 

have a dosimeter or never had recordable dose, assign environmental dose only unless there is information in the file to 

indicate otherwise.

5.1.6 Pantex Workers at Other AEC/DOE Facilities

Due to the nature of the work at Pantex, workers were sometimes required to temporarily perform 

their duties at other U.S. Atomic Energy Commission (AEC)/DOE facilities (e.g., the Nevada Test Site

(now Nevada National Security Site), modification centers, national laboratories, etc.) and might have 

been monitored for occupational radiation exposures by Pantex, the temporary work location, or 

concurrently by both facilities. In such cases, all available monitoring records should be used to 

assign claimant doses. 

5.2 RADIONUCLIDES WITH POTENTIAL FOR INTERNAL DOSE

Only five groups of radioactive materials are of concern for occupational intake at Pantex: tritium, 

uranium, thorium, plutonium, and radon progeny (BWXT Pantex 2001). BWXT Pantex (2001) 

discusses the first four radionuclides; the latest version of ORAUT-TKBS-0013-2, Pantex Plant – Site 

Description (ORAUT 2007) discusses processes and locations where radioactive material could have 

been present.

5.2.1 Tritium

The principal sources of tritium at Pantex were and are the weapons components known as 

reservoirs, which first arrived at Pantex in late 1956 or early 1957 (Martin 2004). A Cockcroft Walton 

neutron generator in use before 1956 produced some tritium in the off-gas, and titanium tritide 

particulate contamination probably existed in the target and the area where the drift tube connected to 

the target, but the amount would have produced negligible intakes [10]. Tritium sealed under high 

pressure in the reservoir units has the potential to leak during disassembly. BWXT Pantex (2001) 

states that tritium could leak through reservoir materials, which presumably refers to concern for 

migration of molecular tritium through welds. The tritium in the reservoirs is 99% gaseous molecular 

hydrogen (DT, HT, or T2) and 1% tritiated water vapor (HTO or T2O) (BWXT Pantex 2001). Tritium 

gas interacts over time with moisture in the air, hydrogenated materials (e.g., hydrocarbons, organic 

compounds, and concrete), and some forms of metals to form tritiated compounds and metal tritides 

(Peterson and Davis 2002).

Tritium gas is far less hazardous than tritiated water, organically bound tritium, or metal tritides, but it 

combines with water vapor in the air or body tissues to form compounds (Johnson and Hill 1993). Of 

particular importance is tritiated water, which the human body absorbs. Elemental tritium is not 

absorbed through the skin to a significant degree. Tritiated water vapor is readily absorbed through 

the skin and lungs and retained in the body. Tritiated water that enters the body is chemically 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 15 of 49

identical to ordinary water and is distributed throughout the entire mass of body water (Johnson and 

Hill 1993).

5.2.1.1 Internal Assessment for Tritium During Routine Operations

Notes: For tritium, uptake refers to total tritium distributed in body fluids regardless of mode 

of intake. Uptake can be thought of as total intake and includes skin absorption. 

Uptake is equivalent to whole body in the Integrated Modules for Bioassay Analysis 

(IMBA) program, and is the product of the urine concentration in activity per liter 

multiplied by 42 L of body fluids.

The following discussion makes no distinction between annual dose and committed 

dose. 

Because the tritium uptakes discussed in this section were determined from Pantex 

dose calculations, which in turn were determined from urine samples, dose 

reconstructors should consider uptakes to be considered HTO (inorganic tritium in 

IMBA) unless otherwise noted in a worker’s records.

Pantex analyzed tritium bioassays on the site. For workers assigned to the tritium bioassay program, 

the frequency of monitoring was monthly. In addition, there were bioassays for new hires and 

terminations (although it is not clear if this was for all new hires or just those expected to be in the 

bioassay program). In addition, for each month, one-twelfth of the worker population received an 

annual urinalysis (Alley 1990).

The following tritium discussion deals with two periods: 1956 to 1971 and 1972 to the present. The 

discussion explains the selection of those periods.

The extent of a routine tritium bioassay program before 1972 is unclear although there are indications 

of sampling of about 10 workers per month in the 1960s [11]. Because there is little evidence that 

workers were monitored for tritium before 1972, dose reconstructors should interpret routine 

occupational records before 1972 that show “0” for internal dose to mean that no information is 

available rather than to indicate a dose below detectable levels [12]. Dose records in the 1990s 

specifically state monitored internal emitters (and give a dose) or state not monitored as “N/M.” Only 

a few NOCTS files had tritium bioassay results for years before 1972. However, the detection 

sensitivity was poor for the analytical method used, and the results are therefore not a good indicator 

of true exposure levels. Tritium doses in the files should be treated the same as those for 1972 to 

1982 as described below.

From 1972 to the present, although tritium bioassay occurred, there are few routine monitoring data in 

individual worker dosimetry records. A few urinalysis records for 1972 show consistent use of 

0.25 μCi/L as a less-than value (MHSMC 1983). A batch of urinalysis records for 1983 shows 

background counts per minute, gross counts per minute, and final concentrations in microcuries per 

liter. It appears that Pantex recorded nonzero concentrations when the gross counts per minute 

exceeded the square root of twice the background counts per minute, which would be a decision 

level. The smallest nonzero concentration recorded was 0.023 μCi/L, so 0.05 μCi/L would be a 

reasonable estimate of the minimum detectable activity (MDA) at that time (MHSMC 1983). Technical 

Basis for the Internal Dosimetry Program and the DOE Pantex Facility (Battelle 1992) lists the tritium 

urinalysis “detection limit” as 14 dpm/mL (0.0063 μCi/L). Although it does not state so directly, this 

document implies that this is the MDA. BWXT Pantex (2001) lists the tritium urinalysis minimum 

detectable activity (MDA) as 15 dpm/mL (0.0068 μCi/L).
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 16 of 49

Based on the scarcity of individual claimant data, more conservative MDAs were used in the 

determination of potentially missed and unmonitored tritium doses. For 1956 to 1990, an MDA of 

0.500 μCi/L [13] was assumed. For 1991 through the present, an MDA of 0.135 μCi/L [14] was 

assumed (Table 5-3). However, in cases where individual claimant data are present in the monitoring 

records, the missed or fitted dose should be calculated based on the claimants’ monitoring data.

Table 5-3. Internal dose for missed and unmonitored tritium intakes.

Year

Maximum 

undetected 

individual annual 

tritium dose 

(mrem)a

Maximum annual 

undetected intake 

(μCi)

Mode of the annual 

undetected 

individual annual 

tritium dose 

(mrem)

Mode of the annual 

undetected intake 

(μCi)

1956–1990 42 619 21 309

1991–present 12 167 6 83.5

a. Calculated based on 1 month of chronic exposure assuming detection limits of 0.500 μCi/L through 1990 and 

0.135 μCi/L for 1991 through the present, and normalized for annual intake using the ICRP Publication 68 dose 

coefficient of 0.067 mrem/μCi (ICRP 1995). Note that these values exceed any recorded doses or intakes the site 

reported for any year of operation, including 1989 when a major tritium released occurred.

5.2.1.1.1 Dose to Uptake

The most complete set of tritium information consists of maximum and average doses for 1972 to 

2001 (Table 5-3). Because it is likely that the dose reconstructor will find only tritium doses rather 

than actual bioassay results in the worker files, Attachment C provide methods to convert from 

recorded dose to uptake (for input into IMBA or the tritium tool).

5.2.1.2 Unmonitored Workers, 1956 to 1971

During weapons assembly, there was little chance that tritium could leak because workers did not 

manipulate the valves on the tritium reservoir [15]. A very small amount of tritium migration through 

reservoir welds occasionally occurred, which is why workers surveyed the reservoirs on arrival [16]. 

However, weapons brought in for inspection, repair, or disassembly provided a possibility for a small 

release of tritium and subsequent intake for Production Technicians (Assembly Operators), RSTs, and 

Quality Assurance Technicians. Around 1980, disassembly of weapons became more frequent than 

assembly, and releases were more likely to occur [17].

5.2.1.3 Unmonitored Workers, 1972 to the Present

It is unlikely that unmonitored workers had higher intakes than monitored workers. However, for the 

period from 1972 to 1976 and for 1984, there was little or no tritium bioassay monitoring 

(BWXT Pantex 2005). During the years in which monitoring occurred, there is no guarantee that 

everyone exposed to tritium was monitored. Nevertheless, for Production Technicians (Assembly 

Operators), RSTs, and Quality Assurance Technicians, tritium uptakes were possible. This analysis 

estimated uptakes for unmonitored workers in these job categories using Table 5-3 values; the 

average uptake was not allowed to become less than that which would result in a dose of 6 mrem/yr 

[18]. Dose reconstructors should consider tritium uptakes for unmonitored Production Technicians 

(Assembly Operators), RSTs, or Quality Assurance Technicians to be a triangular distribution with the 

modes and maximums in Table 5-3 and minimums of zero.

There is no reason to expect workers other than Production Technicians (Assembly Operators), 

RSTs, and Quality Assurance Technicians incurred uptakes of tritium other than from environmental 

sources [19].
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 17 of 49

5.2.1.4 Tritium Release Event in 1989

During a release event in 1989, significant amounts of tritium were released and four workers were 

exposed. This event is reasonably well documented, and data are available for assessment. Data on 

this event are from two primary sources: an interoffice memorandum with worker exposure records 

information (Griffis 1990) and a classified document. An individual had an acute tritium exposure at 

2:30 p.m. (1430) on May 17, 1989. No alpha contamination was found on the individual. Urine 

specimens were taken 2 and 4 hours after the incident. According to unclassified information from the 

classified report, the first bioassay occurred at 4:30 p.m. using a 0.5-mL aliquot specimen pipetted 

into 10-mL Biofluor. This sample was analyzed using a Tri-Carb Model 2250 CA liquid scintillation 

counter. The tritium activity was 291,000 dpm, which was equivalent to 262 μCi/L of tritium. 

The individual received medical care for the intake that included many special urine and blood 

samples and mandatory forced fluids. Urinalysis results are in the worker’s records and are 

summarized in a letter from the Medical Director (Lang 1990).

The latest version of ORAUT-TKBS-0013-4, Pantex Plant – Occupational Environmental Dose, 

addresses the environmental release from this incident (ORAUT 2014b).

5.2.1.5 Assignment of Tritium Doses for Production Technicians (Assembly Operators), 

RSTs, and Quality Assurance Technicians, 1956 to Present

For workers in which the monitoring records indicate that the individual was monitored for tritium 

intakes and for production technicians, RSTs, and Quality Assurance Technicians, tritium dose should 

be assigned using Table 5-3 as a triangular distribution using zero and the mean and maximum from 

Table 5-3). On occasion, the monitoring records have the individual specific monitoring results. 

Please refer to Attachment A, Examples of Pantex Plant Bioassay Data Files, for examples of the 

layout of the data files when individual specific monitoring data are available.

5.2.2 Uranium

5.2.2.1 Background

Uranium at Pantex was enriched (EU), natural, or depleted (DU). Natural uranium was in a form 

referred to as Tuballoy. EU was in a sealed component with little likelihood of release. No data are 

available to indicate that EU was ever a contaminant in the workplace. The internal dosimetry 

technical basis document (Battelle 1992) stated, “All of the unsealed uranium used at the Pantex 

facility is either depleted uranium or natural uranium.” DU manufactured after 1952 could have 

contained contaminants from movement of recycled uranium and DU throughout the Portsmouth, 

Paducah, K-25, and Y-12 sites. Exact levels of contaminants in Pantex DU have not been discovered 

and probably varied from batch to batch. As an upper bound, dose reconstructors should add the 

following intakes of contaminants to DU intakes: 307 pCi 239Pu/g DU, 3.53 pCi 237Np/g DU, and 

509 pCi 99Tc/g DU [20].

According to BWXT Pantex (2001) and interviews with workers, uranium contamination at Pantex is 

either uranium metal or air-oxidized uranium. Exceptions would be the burning of DU-contaminated 

HE components at the burn pads and explosion of DU during hydro tests, which would have produced 

some thermally oxidized DU. BWXT Pantex (2001) states that uranium compounds at Pantex are 

assumed to exhibit class Y inhalation behavior. However, an earlier internal dosimetry technical basis 

document (Battelle 1992) used an assumption of 80% class Y and 20% class W. Neither assumption 

was based on solubility studies of actual Pantex contamination. Because oxides of uranium can exist 

over a range of solubility, dose reconstructors should assume either absorption type M or S to 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 18 of 49

maximize the dose to the specific organ of concern. Exposure to significant quantities of type F 

uranium at Pantex is not credible [21].

Because components are new during assembly operations, there is little likelihood that significant 

removable DU oxide would have been on them. During disassembly, aged uranium components from 

certain weapons programs had a coating of oxide in the form of black dust that was potentially present 

as airborne contamination [22]. Uranium oxide became most noticeable beginning in the early 1980s 

and was present on eight of the 31 weapons types that were disassembled at Pantex to date, with 

types 28 and 55 apparently having the highest contamination from black dust according to worker 

interviews. Following a contamination event in 1989, consideration for preventing contamination by 

uranium oxide resulted in modifications to disassembly operations, such as the use of downdraft 

tables.

Some DU was released at the burning grounds from burning of contaminated HE and at the hydro test 

firing sites when hydro tests involved DU components (Firing Sites 4, 5, and 10 only). In addition, one 

weapon design required machining of DU-contaminated metal [23].

There is no evidence that workers were routinely monitored for uranium before 1991 and there are 

numerous references that indicate that bioassay was not performed routinely but rather only when 

there was a known release [24]. Dose records in the 1990s specifically state whether internal emitters 

were monitored (and give a dose) or not monitored (N/M). Pantex provided routine urinalysis of 

uranium in 1991 and 1992 [25]. The technical basis document at the time (Battelle 1992) stated that 

the uranium urinalysis method was isotopic analysis using alpha spectrometry that “can detect 

0.03 pCi/isotope/sample.” The document reported an environmental background urinary excretion 

rate of 0.15 dpm/d of 238U based on studies of potentially exposed and unexposed Pantex workers. 

This environmental screening level was carried over to the internal dose assessment procedure 

(MHSMC 1991b), which indicated that dose assessment was to occur for any uranium result with a 

net activity greater than or equal to 0.15 dpm/d. Battelle (1992) stated that the 0.15-dpm/d screening 

level would not apply if isotopic ratios implied that the uranium did not derive from DU.

Since 1993, monitoring of uranium exposures has been event driven and is initiated by air-monitoring 

data. Since the middle 1990s, Pantex has used lapel air samplers to monitor for intakes and trigger 

bioassay measurements. Because Pantex has performed bioassays on more than 300 workers since 

1993 (see Table 5-4), the implication is that there must have been workplace indicators of potential 

uranium intakes [26]. Analysis of Biological Samples for Uranium, Thorium, and/or Plutonium

(MHSMC 1991a) provided the following workplace indicators that would trigger bioassay:

• All personnel ... not wearing ... respiratory protection whose tracked internal 

annual exposure is equal to 40 DAC-hours

• All personnel whose breathing zone monitor indicates that they have been exposed 

to 40 DAC-hours [also lists the DAC for 238U as 6 × 10-11 μCi/mL]

• All personnel found to have skin contamination equal to or greater than ... 

1000 dpm/100 cm2 238U.

BWXT Pantex (2001) decreased the trigger value for special bioassay: “Special bioassay samples 

are collected (usually within 2 to 3 days) when airborne exposures exceed 4 DAC-hours for any single 

actinide (i.e., >4 DAC-hours for 239Pu, 232Th, or 238U creates an occurrence).” This analysis has not 

determined exactly when the change occurred between 1991 and 2001.
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Table 5-4. Uranium dose to workers.a

Year

Workers monitored 

for uranium

Total worker uranium 

dose (person-mrem)

Maximum individual 

uranium CEDE (mrem)

Average worker 

uranium CEDE (mrem)

1990 46 0 0 0

1991 431 109 109 0.25

1992 239 778 502 3.3

1993 90 76 15 0.84

1994 138 0 0 0

1995 37 0 0 0

1996 69 0 0 0

1997 89 0 0 0

1998 12 0 0 0

1999 13 0 0 0

2000 33 0 0 0

2001 65 0 0 0

2002 57 0 0 0

2003 87 10 7 0.11

2004 109 0 0 0

a. Developed from data in DORMS.

Table 5-4 lists recorded doses (CEDE) from uranium exposures from 1991 to 2004 from the facility’s 

dosimetry records management system (DORMS). Although these doses are not directly relevant to 

dose reconstruction, the overall trend is indicative of reduced uranium intakes after 1993.

The weight percent and activity fractions of radionuclides of DU and even natural uranium can be 

variable. Values were listed in the 1992 version of the internal dosimetry technical basis document 

(Battelle 1992), but their origin was not stated and they were not mentioned in BWXT Pantex (2001). 

The 1992 values are not significantly different from the default values in the IMBA program, so dose 

reconstructors should use the IMBA values for consistency.

5.2.2.2 Uranium Reporting Levels or Minimum Detectable Activities

For most of its history, Pantex followed an event-driven approach to uranium bioassay and used many 

laboratories, so the records for bioassay results are spotty [27]. Table 5-5 summarizes information 

found in claimant records. For the years prior to 1990, the MDA value in the table was based on the 

observed less than values listed in the records. 

Table 5-5 has temporal gaps likely associated with the absence of bioassay sample submissions for 

the years in which no data were listed. This is due to the fact that uranium bioassays were generally 

not obtained routinely but usually driven based on special bioassay samples after events with 

potential for intake. If necessary, dose reconstructors should use the last previous MDA for years not 

covered in Table 5-5.

Most documentation of uranium exposure at Pantex focuses on DU, but BWXT Pantex (2001) 

mentions the possibility of exposure to natural uranium. When interpreting bioassay data, if the type 

of uranium exposure is not known, it is favorable to claimants to assume the intake was natural 

uranium (0.685 pCi/μg) (BWXT Pantex 2001).

5.2.2.3 Assessment of Uranium Intake and Internal Dose 

Because of the SEC Class, assessment of uranium intakes at Pantex is limited to actual data in the 

monitoring records. Dose reconstructors should assess occupational intakes in accordance with the 

guidance in ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 2014a). On occasion, the 

monitoring records have the individual specific monitoring results. Please refer to Attachment A, 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 20 of 49

Table 5-5. History of uranium urinalysis.

a

Year Laboratory MDA Value

1959 Los Alamos Scientific Laboratory 0.5 μgb

1960 Tracer Laboratory 10 μg/Lb

1963 Controls for Radiation 0.10 μg/Lb

1965 Controls for Radiation 0.10 μg/Lb

1967 Controls for Radiation 0.15 μg/Lb

1968 Isotopes, Inc. 0.10 μg/Lb

1983 Camp Dresser & McKee 3.30 pCi/Lb [28]

1983 Los Alamos Scientific Laboratory 5.0 μg/L

1990–2001 Y-12 Bioassay Laboratory 0.15 dpm/dc[29]

a. ORAUT 2012.

b. From reports from the laboratories (Author unknown 1952; MHSMC 1978; 

MMES 1990a).

c. The value of 0.15 dpm/d is a conservative estimate of the total uranium MDA for samples 

processed at the Y-12 National Security Complex after 1989 [30].

Examples of Pantex Plant Bioassay Data Files, for examples of the layout of the data files when 

individual specific monitoring data are available. However, if the monitoring data indicates that the 

analyses were performed by CEP, those data should not be used for the assessment of intake and 

the assignment of internal dose. 

5.2.2.4 Thorium

Thorium at the Pantex Plant exists as thorium metal, thorium alloys, or materials impregnated with a 

thorium compound. Workers handle these forms during assembly and disassembly of certain 

weapons. Because of the relative hazard of thorium, Pantex uses strict workplace monitoring 

practices, such as smear checks of components, to verify the integrity of the thorium encapsulation. It 

is assumed that workers could have encountered oxidized thorium components during disassembly of 

weapons. Pantex has never conducted machining of components containing thorium [31].

Information on source terms of weapons containing thorium is classified, as is the number or 

percentage of weapons that contain thorium. However, there is strong indication that controls for 

contamination have always been in place, as has workplace monitoring for thorium [32].

Natural sources of thorium can exist in rocks and soils (see the latest version of the Pantex TBD on 

occupational environmental dose (ORAUT 2014b). Thorium can be present in measurable amounts 

in biological materials in the environment; ingestion of these materials can result in measurable 

quantities of thorium in bioassay samples collected from workers. Baseline bioassay measurements 

have shown this to be true. Pantex determined the amount of thorium that is naturally present in 

baseline bioassay samples for its workers (BWXT Pantex 2001).

Levels of 232Th and 228Th were analyzed in baseline fecal samples. Environmental levels of 232Th and 

228Th were determined using a lognormal probability analysis on the bioassay data in accordance with 

A Guide for Environmental Radiological Surveillance at U.S. Department of Energy Installations

(Corley et al. 1981). The 95th-percentile results for background levels of thorium in fecal samples 

were 0.4 dpm/sample (0.18 pCi/sample) of 232Th and 0.39 for the 232Th:228Th ratio. Results that 

exceed both screening levels are assumed to represent occupational exposure and are adjusted to a 

net occupational excretion by subtracting the arithmetic mean background excretion rate of 

6.7 × 10-2 dpm/sample (3.0 × 10-2 pCi/sample) from 232Th fecal results (BWXT Pantex 2001, p. 80).

BWXT Pantex (2001) states that thorium at the Plant is inhalation class Y, which would be essentially 

equivalent to absorption type S. This is consistent with the ICRP Publication 68 recommendation that 

thorium oxides are type S (ICRP 1995). Although processing of thorium at Y-12 and Hanford created 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 21 of 49

disequilibrium between 232Th and 228Th, material handled at Pantex would have aged long enough that 

a significant amount of 228Th would have grown back, especially for weapons being disassembled 

[33]. The dose reconstructor should assume equilibrium without evidence to the contrary [34].

There is no evidence that workers were routinely monitored for thorium before 1991, and there are 

numerous references that indicate that bioassay was not performed routinely but rather only when 

there was a known release [35, 36]. Monitoring of thorium exposures has been event-driven since at 

least 1991 [37]. The procedure Analysis of Biological Samples for Uranium, Thorium and/or 

Plutonium provided criteria for when thorium bioassay monitoring was required (MHSMC 1991a). To 

summarize, the criteria were exposure to 40 DAC-hr of thorium in the workplace air after accounting 

for use of respiratory protection, if applicable, or skin contamination equal to or exceeding 

200 dpm/100 cm2

. The only reported doses have occurred since 1999 [38]. This analysis found no 

bioassay data before 1983.

5.2.2.5 Assessment of Thorium Intake and Internal Dose 

Because of the SEC Class, assessment of thorium intakes at the Pantex Plant is limited to actual data 

in the monitoring records. If the monitoring results indicate that the individual was monitored for 

occupational intakes of thorium, dose reconstructors should assess occupational intakes in 

accordance with the guidance in ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 2014a). 

If unmonitored, dose reconstructors should assume occupational intakes of thorium were unlikely and 

assign no thorium internal dose. On occasion, the monitoring records have the individual specific 

monitoring results. Please refer to Attachment A, Examples of Pantex Plant Bioassay Data Files, for 

examples of the layout of the data files when individual specific monitoring data are available. 

However, if the monitoring data indicates that the analyses were performed by CEP, those data 

should not be used for the assessment of intake and the assignment of internal dose.

5.2.3 Plutonium

5.2.3.1 General Information

Plutonium at Pantex is in the encapsulated pits of nuclear weapons. Workers handle the pits during 

weapons assembly and disassembly. Strict workplace monitoring practices ensure the integrity of the 

encapsulation including contamination smear checks during assembly and disassembly (BWXT 

Pantex 2001). If contamination occurred, exposure to plutonium would be acute rather than chronic. 

Table 5-6 lists the numbers of workers given plutonium bioassay by year. There were no recorded 

internal doses associated with these 1991 to 2002 bioassays.

Because the plutonium was encapsulated, it was assumed that the potential for intake was rare (i.e., 

intakes would have been acute rather than chronic) [39]. BWXT Pantex (2001) states that plutonium 

at the Plant should be considered an aged weapons-grade mixture. For the following discussion, the 

intake activities are for the total alpha activity of the mixture. Dose reconstructors should assume the 

20-year aged mixture [40]. Table 5-7 lists the composition of weapons-grade plutonium mixtures. 

Because the source of intake would have been plutonium oxides, dose reconstructors should assume 

inhalation type S (ICRP 1994).

5.2.3.2 Assessment of Plutonium Intake and Internal Dose After 1983

Because of the SEC Class, assessment of plutonium intakes at the Pantex Plant is limited to actual 

data in the monitoring records. If the monitoring results indicate that the individual was monitored for 

occupational intakes of plutonium, dose reconstructors should assess occupational intakes in 

accordance with the guidance in ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 2014a). 

If unmonitored, dose reconstructors should assume occupational intakes of plutonium were unlikely 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 22 of 49

Table 5-6. Number of workers on plutonium 

bioassay, 1991 to 2002 [41]. 

Year

Number of workers 

monitored for plutonium

1991 0

1992 12

1993 0

1994 0

1995 28

1996 17

1997 18

1998 2

1999 1

2000 8

2001 1

2002 10

2003 9

2004 0

Table 5-7. Activity composition of weapons-grade plutonium mixtures from Hanford.

a

Mixture designation: Fresh 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aging:b 0 5 10 15 20 25 30

Specific activity in mixture (Ci/g)

Pu-238 8.56E-03 8.23E-03 7.91E-03 7.60E-03 7.31E-03 7.03E-03 6.75E-03

Pu-239 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02

Pu-240 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02

Pu-241 8.24E-01 6.48E-01 5.09E-01 4.00E-01 3.15E-01 2.48E-01 1.95E-01

Pu-242 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06

Am-241 0 5.83E-03 1.04E-02 1.39E-02 1.66E-02 1.87E-02 2.03E-02

Pu-239+240 7.13E-02 7.13E-02 7.13E-02 7.13E-02 7.12E-02 7.12E-02 7.12E-02

Pu-alpha 7.99E-02 7.95E-02 7.92E-02 7.89E-02 7.85E-02 7.83E-02 7.80E-02

Total alpha 7.99E-02 8.53E-02 8.96E-02 9.28E-02 9.52E-02 9.70E-02 9.83E-02

Activity ratios

Pu-239+240: total alpha 1.00 0.836 0.796 0.768 0.749 0.735 0.725

Pu-238: total alpha 0.107 0.0965 0.0883 0.0819 0.0768 0.0725 0.0687

Pu-241: total alpha 10.3 7.60 5.68 4.31 3.31 2.56 1.98

Am-241: total 0 0.0684 0.116 0.150 0.174 0.193 0.207

a. BWXT Pantex (2001) did not provide a table of isotopic mixtures. The Hanford mixtures should be close enough for the 

default assumptions. The total alpha specific activity changes only about 10% from 10 to 30 years of aging.

b. Time since separation of 241Am from the Pu mix.

and assign no internal dose from plutonium intake. On occasion, the monitoring records have the 

individual specific monitoring results. Please refer to Attachment A, Examples of Pantex Plant 

Bioassay Data Files, for examples of the layout of the data files when individual specific monitoring 

data are available. However, if the monitoring data indicates that the analyses were performed by 

CEP, those data should not be used for the assessment of intake and the assignment of internal 

dose.

5.3 ASSESSING OCCUPATIONAL INTERNAL DOSE FROM ELEVATED RADON

Uranium occurs naturally in virtually all soils, with average levels of about 1 part per million. 

Radium-226 is typically in secular equilibrium with 234U and decays to 222Rn with a half-life of 

1,600 years. Therefore, the noble gas 222Rn is continuously produced in soil where it can be trapped 

in the crystalline structure of minerals or released to the interstices between solid materials. In the 

absence of buildings, 222Rn produced within a meter or so of the soil surface can diffuse into the 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 23 of 49

atmosphere where diffusion and advection dilute it with outdoor air (NCRP 1984a). During the 1980s, 

it was discovered that buildings with heating and air conditioning tend to operate at slightly negative 

pressure [a few tens of pascals, less than 1 in. (water gauge)] in comparison to outdoor air. As a 

result of this negative pressure, soil gas tends to flow actively into indoor air, where it can build to 

higher levels than outdoors due to limited air changes and relatively small dilution volumes. This 

phenomenon is an example of technological enhancement of natural radioactivity (NCRP 1984a,

1984b, 1987).

While the general characteristics of areas with potential for elevated levels of indoor radon as well as 

construction designs that tend to enhance radon levels are known, it is rarely possible to predict 

indoor radon levels for a given structure. In general, structures that exhaust air to the environment 

without adequately engineered replacement air have higher indoor radon levels than structures that 

do not do this, and structures that have exposed soil (dirt floors, sumps) or exposed minerals 

(e.g., gravel) tend to have higher radon levels. Underground structures have a higher ratio of soil 

surface to building volume. All other factors being equal, an underground building would be likely to 

have a higher radon concentration than an aboveground building.

Thorium has 220Rn progeny that is a radioactive noble gas, commonly called thoron, which has a 

much shorter half-life (55.6 seconds) than its parent. In general, 220Rn decays before it can build up to 

significant levels unless there are large quantities of 232Th and its decay products present. There is no 

reason to expect that Pantex had 220Rn of significance. Work on thorium weapon components was 

less frequent.

5.3.1 Dose from Radon-222 Progeny

Radon itself produces far less dose to the bronchial epithelium than its progeny. Because radon 

progeny measurements are more difficult to obtain, measurements of radon are often used as a 

surrogate for progeny measurements. Radon progeny concentrations are expressed as the quantity 

potential alpha energy concentration (PAEC), traditionally measured in working levels (WL). 

Originally, 1 WL was defined as 100 pCi/L (1 × 10−10 Ci/L = 1 × 10−7 Ci/m3 = 1 × 10−7 μCi/cm3

) of radon 

in equilibrium with its short-lived decay products. At present, 1 WL is usually defined as any 

combination of short-lived radon decay products in 1 L of air, without regard to the degree of 

equilibrium, that will result in the ultimate emission of 130,000 MeV of potential alpha energy per liter 

of air. This is almost identical to the original definition. Time-integrated exposures to radon progeny 

are expressed in the quantity potential alpha energy exposure (PAEE), which are traditionally 

measured in working level-months (WLM) and defined as exposure to 1 WL for 170 hours or any 

equivalent concentration and time product (ICRP 1981).

5.3.2 Radon Monitoring at DOE Facilities

In 10 CFR 835.2(a), DOE states, “Background means radiation from ... radon and its progeny in 

concentrations or levels existing in buildings or the environment which have not been elevated as a 

result of current or prior activities....” Because background is specifically excepted from monitoring 

requirements, DOE and its contractors generally do not monitor for radon and its short-lived decay 

products. However, if radon and its progeny are elevated due to DOE activities, then DOE requires 

monitoring. This has been the policy of DOE and its predecessor agencies [42] (the U.S. Energy 

Research and Development Administration, the AEC, and the Manhattan Engineer District). 

5.3.3 Underground Buildings

At Pantex, the Gravel Gertie cells are in Buildings 12-44, 12-85, 12-96, and 12-98, which are 

considered to be underground even though they are not below grade. Bays, which are also 

considered underground, are in Buildings 12-17, 12-19, 12-21, 12-56, 12-64, 12-84 East, 12-84 West, 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 24 of 49

12-99, 12-104, and 12-117. Workers in these buildings were likely to have greater exposures to 

radon and its decay products than workers in other buildings [43].

5.3.4 Radon Concentrations

A DOE-wide survey of radon levels (UNC Geotech 1990) sampled 137 locations at Pantex and made 

duplicate measurements at 13 locations. Attachment B lists complete survey data.

Eight buildings at Pantex measured above 4 pCi/L (4 ×10-9 μCi/cm3

), which is the U.S. Environmental 

Protection Agency reference point for considering remedial action for indoor radon (UNC Geotech 

1990). As listed in Table 5-8, the average for all buildings was 1.62 ±1.24 pCi/L with a geometric 

mean (median) of 1.37 pCi/L and a geometric standard deviation (GSD) of 1.68. Values ranged from 

0.8 to 8.1 pCi/L. Underground buildings had a higher average, and aboveground buildings had a 

lower average.

Table 5-8. Summary statistics of 1990 radon measurements [44]. 

Parameter All buildings Underground buildings Aboveground buildings

Mean (pCi/L) 1.62 1.81 1.56

Standard deviation (pCi/L) 1.24 1.35 1.21

Coefficient of variation 0.77 0.75 0.77

Geometric mean (pCi/L) 1.37 1.51 1.33

GSD 1.68 1.75 1.66

Minimum (pCi/L) 0.8 0.8 0.8

Maximum (pCi/L) 8.1 7.1 8.1

Maximum/Minimum 10.1 8.9 10.1

Count 137 31 106

Considering the uncertainty in these measurements, the average absolute difference between 

duplicate measurements was 0.27 pCi/L with no obvious dependence on the average value of the 

measurement (Figure 5-1). The average ratio was 1.03, which indicates no significant bias.

Figure 5-1. Absolute differences between duplicate radon 

 measurements [45]. 

Far more important than measurement uncertainty is the issue of representativeness (i.e., an 

uncertainty that cannot be quantified from available measurements). Most of these Pantex 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 25 of 49

measurements were made over a 2-month period during the winter, which is normally expected to be 

the time with the highest radon concentrations because buildings are closed and heated most of the 

time [46].

There is an earlier set of radon measurements. For 6 months at the beginning of 1969, Pantex 

monitored radon levels in Cells 1 to 6 on a twice-monthly basis using Eberline-supplied radon film 

badges. The raw results were reported as number of tracks in exposed and covered areas, and the 

integrated radon concentration (in picocurie-hours per cubic meter) was inferred from the net number 

of tracks (McFall 1969). The integrated radon concentration was converted to an average radon 

concentration in picocuries per liter by dividing by the number of hours of exposure and multiplying by 

1,000 cm3

/L. Of the 66 radon film badges issued, this TBD analysis found no record of analysis for 6, 

and 60 had reported analyses. Of the 60 reported analyses, 6 were damaged. Of the 54 undamaged 

results, 33 were reported as zero. When the zeros are included, the overall mean concentration in the 

cells was 4.2 ± 8.6 pCi/L with a range from 0 to 47.2 pCi/L. Fitting a lognormal distribution to all 54 

points yielded a median of 1.2 pCi/L with a GSD of 6.7. This median is slightly lower than the median 

from the 1990 data, and the GSD is considerably larger. These results probably reflect (1) the time of 

year of the sampling and the longer period over which the sampling took place and (2) the difference 

in the sampling methods [47]. Table 5-9 summarizes the statistics from the 1969 radon datasets.

Table 5-9. Summary of 1969 radon measurements in 

Cells 1 to 6 [48]. 

Parameter pCi/L WL

Mean 4.24 0.0170

Standard deviation 8.58 0.0343

Coefficient of Var. 202%

Minimum 0 0

Maximum 47.22 0.1889

Count 54

Using 54 measurements (including 33 zeroes):

Lognormal median 1.20 0.0048

GSD 6.70 6.70

Lognormal mean 7.33 0.0293

Lognormal std. dev. 44.2 0.1767

Using 21 nonzero measurements:

Lognormal median 7.34 0.0293

GSD 2.47 2.47

Lognormal mean 10.91 0.0437

Lognormal standard deviation 10.89 0.0436

5.3.5 Working Level-Months

The Pantex-measured radon concentrations were converted to equilibrium equivalent concentrations 

by multiplying the radon concentration C by the equilibrium factor F using an assumed value of 0.4 as 

recommended by the ICRP (1981) and the United Nations Scientific Committee on the Effects of 

Atomic Radiation (UNSCEAR 1993). The equilibrium equivalent concentration was divided by 

100 pCi/L/WL to arrive at the PAEC. These operations were combined to create:

PAEC C F = × / 100 pCi/L/WL (5-1)

where C is the radon concentration in picocuries per liter and PAEC is in WL. Dose reconstructors 

should multiply the PAEC by the months per year of exposure to determine the WLM for input to the 

Interactive RadioEpidemiological Program (IREP).
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 26 of 49

For workers who spent most of their time in a facility with an earthen cover (Level 1 from Table 5-2), 

dose reconstructors should use the 1990 median value for underground buildings of 1.5 pCi/L for C

and 12 months for the period (unless the person only worked for part of a year). This results in an 

annual average exposure of:

(1.5 pCi/L)(0.4)(12 months)/100 pCi/L/WL 0.027 WLM/yr = (5-2)

For workers with possible occasional entries into underground buildings (e.g., those with a risk 

ranking of 2 in Table 5-2), dose reconstructors should assume that the WLM are one-tenth of those 

from Equation 5-2 [49].

Radon exposure applies from 1958 when the Gravel Gerties were completed to the present [50]. The 

exposure distribution is lognormal [51]. Parameter 1 is the median value in WLM from Equation 5-2. 

Parameter 2 is the GSD. Dose reconstructors should use a GSD of 3 to allow for uncertainties in the 

application of the 1990 radon measurements to a full year (rather than only winter months) and to 

account for possible yearly differences in radon due to frozen ground or snow cover [52].

5.4 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 (SRDB).

[1] Hickey, Eva Eckert. Oak Ridge Associated Universities (ORAU) Team. Principal Health 

Physicist. April 2004.

Information on the use of the cell was verified during an interview trip.

[2] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

Review of records from data capture and interview trips revealed no routine bioassay before 

1972 (see Table 5-1). Review of the HERS database confirmed this.

[3] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

Review of materials showed letters and memoranda in the 1960s that discussed bioassays. 

Because the statement is of a general nature and does not provide specific data, a reference 

was not given.

[4] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

Comment made based on data reviewed from the records in DORMS as well as discussion in 

the Pantex Internal Dosimetry Technical Basis Document (Battelle 1992).

[5] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

The author and others on the Pantex team repeatedly requested air-monitoring and -sampling 

data. They were told that the data were not retrievable and the data would not be useful for 

analyzing intakes from radionuclides.

[6] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

Based on discussions with Plant employees familiar with the work during data capture and 

interview trips to the Plant and confirmed by a former manager.
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[7] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

The Cell 1 tritium incident and the W28 disassembly/uranium issues are the basis for this 

comment.

[8] Hickey, Eva Eckert. BPNNL. Principal Health Physicist. November 2003. 

The list in this table was developed from discussions with Pantex employees, including a 

former manager. The list was then confirmed using information on Job Titles/Classifications 

from 1954 to 2003 (MHSMC 1983). 

[9] Hickey, Eva Eckert, and Bihl, Donald E. BPNNL. Principal Health Physicists. April 2006. 

This statement is based on discussions with Plant employees familiar with the work and with a 

former manager. The risk of intakes was highest in the cells during disassembly operations, 

but some contamination, and therefore risk of intake, might have been present in cells after 

disassembly operations ceased when security personnel were performing security checks. 

Security personnel also had potential for intake while performing duties in weapons storage 

igloos. Security personnel were in areas with potential for contamination only a small 

percentage of their work hours, as opposed to workers who were involved in disassembly 

operations, who spent more daily hours in potential contamination areas.

[10] Bihl, Donald E. ORAU Team. Principal Health Physicist. September 2006.

Assuming a resuspension factor of 10-4/m, an acute 2-hr exposure during target changeouts, 

no respiratory protection, and changeouts once a year, the titanium tritide contamination level 

needed to produce an intake that would result in a 1-mrem dose to the lung would have been 

1.6 × 108 dpm/100 cm2

. Doses to all other organs would have been less. Although actual 

contamination levels have not been found, the author judged that they would have not reached 

this level.

[11] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. November 2005. 

These data are from DORMS and were accessed November 17, 2005, by M. Prather (BWXT 

Pantex 2005). In addition, there was an occasional reference to bioassay in memoranda from 

the 1960s or 1970s. No data were available from these memoranda.

[12] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. February 2004. 

Review of dosimetry records from the 1960s and 1970s often showed a zero in the column for 

“internal emitters”; however, there are numerous references that indicate that bioassay was 

not performed routinely but rather only when there was a known release. Therefore, the dose 

reconstructors need to be careful and not assume that a zero means that someone was 

monitored. The dose records in more recent years (1990s forward) clearly indicate when an 

individual was monitored for internal dose.

[13] Thomas, Dale D., ORAU Team, Senior Health Physicist, October 2014. 

Review of the records shows that the detection limit for tritium analysis was somewhat less 

than 0.5 uCi/L. However, some individual records listed the detection limit as high as 0.5 

uCi/L; therefore, the unmonitored tritium dose was calculated based on the higher detection 

limit to ensure favorability to claimants through 1990. 

[14] Thomas, Dale D., ORAU Team, Senior Health Physicist, October 2014. 

Similar to [13] except that the value was selected based on the highest detection limits for 

years after 1990. 

[15] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. May 2004. 

Based on discussions with Plant employees familiar with the work during data capture and 

interview trips to the Plant, and was confirmed by a former manager.
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[16] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. May 2004. 

Based on discussions with Plant employees familiar with the work during data capture and 

interview trips to the Plant. Confirmed by a former manager.

[17] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. May 2004. 

This information was verified during an interview trip and was originally interpreted from a slide 

presentation from a data capture trip (BWXT Pantex ca. 2000).

[18] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2004. 

The mode of the triangular distribution for tritium missed-dose calculations is 6 mrem. It was 

used for unmonitored workers based on the premise that unmonitored workers were at less 

risk than monitored workers, especially for tritium because the source of tritium exposure 

came from handling tritium reservoirs during disassembly. The possibility of casual exposure 

to tritium during the disassembly operations by unmonitored workers was considered remote 

because of the security and other safety requirements of the work.

[19] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2004. 

Calendar year 1981 had the highest maximum uptake of any year monitored except 1989, so 

this value was considered favorable to claimants as a surrogate for 1989.

[20] Hickey, Eva Eckert, and Bihl, Donald E. ORAU Team. Principal Health Physicists. May 2004. 

Tritium exposure occurred during disassembly. Tritium that leaked into the cell was removed 

by ventilation, so residual contamination was not an issue (at normal release levels). 

Therefore, those workers in the cell during disassembly incurred the tritium intakes. There is 

always the possibility that other workers were occasionally in the cell or close by during a 

tritium leak, but these would have been infrequent and the doses would have been less than 

1 mrem/yr.

[21] Chew, Melton H. ORAU Team. Principal Health Physicist. April 2006. 

The values for the impurities were provided by Mel Chew based on research at Y-12 (ORAUT 

2007).

[22] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

The Internal Dosimetry Technical Basis and Quality Assurance Document (BWXT Pantex 

2001) states, “The compounds for uranium at Pantex are pure metal or air-oxides; it is 

assumed that all forms encountered will exhibit class Y aerosol behavior.” This is the basis for 

the statement that there would not be significant quantities of type F uranium at Pantex.

[23] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

The presence of black dust DU contamination was mentioned by many Pantex employees 

during site visits. It is also discussed in a presentation on DU contamination during 

investigation of the 1989 incident (BWXT Pantex ca. 2000).

[24] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

As an example of improvements in contamination control, a presentation from 1989 discussed 

the lessons learned and improvements for contamination control in a presentation on DU 

contamination during investigation of the 1989 incident (BWXT Pantex ca. 2000).

[25] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. July 2004. 

Review of dosimetry records from the 1960s and 1970s often showed a zero in the column for 

“internal emitters”; however, there are numerous references that indicate that bioassay was 

not performed routinely but rather only when there was a known release. Therefore, the dose 

reconstructors need to be careful and not assume that a zero means that someone was 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 29 of 49

monitored. The dose records in more recent years (1990s forward) clearly indicate when an 

individual was monitored for internal dose.

[26] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

Table 5-4 provides the data for uranium dose to workers from 1990 to 2004. This table was 

created using data from DORMS.

[27] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

This statement was made based on inference from the statement that uranium exposures 

were monitored only based on air-monitoring data (i.e., that the presence of uranium was 

detected). If bioassays were made on 300 workers, there must therefore have been the 

potential for uranium intakes.

[28] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2005. 

If 1.4 is the 2-sigma value, then the 1-sigma value is 0.7. One frequently used formula for 

MDA is 4.65σ. 0.7 x 4.65 = 3.3. The sigma value in the MDA equation is arguably not just the 

counting error, but there are not enough data to do a more rigorous calculation.

[29] Thomas, Dale D. ORAU Team. Senior Health Physicist, October 2013.

The total uranium MDA of 0.15 dpm per day was derived based on reasonable assumptions 

regarding possible uranium enrichments and an isotopic specific MDA of nominally 

0.03 pCi/sample for each predominant uranium isotope. The value of 0.15 dpm per day 

accounts for all potential uranium enrichments and normalized twenty-four hour excretion 

rates. 

[30] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2005. 

The MDAs for the Y-12 analyses were printed on the report but they varied for each sample. 

The values shown in Table 5-5 were generally the higher of the various MDAs listed on the 

report (Reichert 1994, pp. 14–17).

[31] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. July 2004. 

The determination that there have been strict workplace restrictions in relation to thorium 

comes from discussions with the staff at Pantex during interview trips and from the Internal 

Dosimetry Technical Basis Manual (BWXT Pantex 2001).

[32] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. July 2004. 

The determination that there have been strict workplace restrictions in relation to thorium 

comes from discussions with the staff at Pantex during interview trips and the Internal 

Dosimetry Technical Basis Manual (BWXT Pantex 2001).

[33] Bihl, Donald E. ORAU Team. Principal Health Physicist. September 2005. 

The age of thorium during disassembly is not known and probably varied from weapon to 

weapon. Thorium-228 activity decreases relative to 232Th activity upon purification of the 

thorium (i.e., removal of the 228Ra) and reaches a minimum of about 44% in 5 to 6 years, then 

increases slowly to reach 90% in about 23 years (see, for instance, West 1965).

[34] Bihl, Donald E. ORAU Team. Principal Health Physicist. September 2005.

Because the degree of disequilibrium of the 232Th progeny is not known, it was judged 

favorable to claimants to assume equilibrium. That assumption increases the activity of 

progeny during intake and consequently increases annual organ doses. 
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[35] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. July 2004.

Review of dosimetry records from the 1960s and 1970s often showed a zero in the column for 

“internal emitters”; however, there are numerous references that indicate that bioassay was 

not performed routinely but rather only when there was a known release. Therefore, the dose 

reconstructors need to be careful and not assume that a zero means that someone was 

monitored. The dose records in more recent years (1990s forward) clearly indicate when an 

individual was monitored for internal dose.

[36] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. September 2005. 

This information was obtained by review of dose records from this period.

[37] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

This statement was made based of review of dosimetry data in DORMS.

[38] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. April 2004. 

This statement was made based of review of dosimetry data in DORMS.

[39] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. July 2004. 

The assumption that the potential for intake was rare is inferred from discussions in BWXT 

Pantex (2001).

[40] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. July 2004. 

BWXT Pantex (2001) states that the plutonium at Pantex is likely to be an aged weaponsgrade mixture. Therefore, the assumption in this section was made to be a 20-year-aged 

mixture.

[41] Hickey, Eva Eckert. ORAU Team. Principal Health Physicist. July 2004.

The data are from the DORMS database.

[42] Strom, Daniel J. ORAU Team. Principal Health Physicist. February 2004. 

This is certainly the policy that has been stated verbally by various DOE policymakers over 

time and by inference from the lack of requirements to monitor for radon in most DOE 

buildings. Finding unequivocal documentation of this policy is difficult, but DOE Order 5480.11 

states in article 9.b, “Note: Natural background and therapeutic and diagnostic medical 

exposures are not to be included in dose records or in assessment of dose against limiting 

values” (DOE 1988). The exact history of the policy is not particularly germane to dose 

reconstruction except to explain the lack of radon and progeny measurements at DOE and 

predecessor organization facilities over the years.

[43] Strom, Daniel E. ORAU Team. Principal Health Physicist. February 2004. 

This statement follows from the first paragraph in Section 5.3 and the fact that a building that is 

surrounded by soil on more sides than just the floor generally has more radon emanating into 

the building than a building with just soil under the floor, unless barriers to radon emanation 

are employed.

[44] Strom, Daniel J. ORAU Team. Principal Health Physicist. February 2004. 

These values are the result of calculations by the author from the data in Table B-1.

[45] Strom, Daniel J. ORAU Team. Principal Health Physicist. February 2004. 

These values are the result of calculations by the author from the Eberline data (McFall 1969).
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[46] Strom, Daniel E. ORAU Team. Principal Health Physicist. February 2004. 

Soil gas is the principal source of indoor radon. In winter, buildings are closed and heated, 

which leads to decreased air pressure in the building because hot air rises and heat leaks 

draw air out of the building. This effect depressurizes the building in relation to soil gas, which 

draws radon in at an increased rate. This effect is especially strong when there is a 

combustion source (with a flue) in a building (e.g., a gas hot water heater or a gas or oil 

furnace). In addition, exhaust ventilation systems and clothes dryers generally depressurize 

buildings and, with closed windows limiting air makeup, there is more soil gas infiltration than 

in seasons when windows are open.

[47] Strom, Daniel J. ORAU Team. Principal Health Physicist. February 2004. 

These values are the result of calculations by the author.

[48] Strom, Daniel J. ORAU Team. Principal Health Physicist. February 2004. 

These values are the result of calculations by the author from the data in Table B-1.

[49] Bihl, Donald E. ORAU Team. Principal Health Physicist. June 2004. 

The use of one-tenth intakes for workers in the second risk category is consistent throughout 

the TBD. It is based on two exposure modes: either occasional exposure to the same air 

concentrations as the high risk workers (modeled as 4 instead of 40 hr/wk), or continuous 

exposure to one-tenth of the contamination levels incurred by the high risk workers, such as 

reduced airborne concentrations in hallways or rooms in the same buildings but distant from 

the cells and bays.

[50] Bihl, Donald E. ORAU Team. Principal Health Physicist. June 2004. 

Radon intakes apply to unusual structures that enhance radon beyond normal concentrations; 

Gravel Gerties were constructed with gravel roofs and soil high on the sides, so they fit the 

criterion.

[51] Bihl, Donald E. ORAU Team. Principal Health Physicist. June 2004. 

Lognormal distribution for air concentrations of contaminants is the usual default unless there 

are data that clearly show otherwise. Radon air concentrations have been shown to have 

lognormal distributions (see, for instance, NCRP 1984a, p. 49).

[52] Bihl, Donald E. ORAU Team. Principal Health Physicist. June 2004.

Table B-1 shows the radon concentrations for the underground buildings to have a GSD of 

1.75 for that set of measurements, which is less than the Project default of 3. Therefore, the 

Project default of 3 takes precedence. For matters that influence internal dose, the smallest 

allowable GSD is 3.

[53] Bihl, Donald E. ORAU Team. Principal Health Physicist. July 2004.

Extrapolation prior to 1972 is based on the issue date of ICRP Publications 10 (1968) and 2 

(1959), so it was assumed that the equations were used throughout these years.

[54] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2004.

This statement is based on personal review of records from Pantex in several claims files.

[55] Hickey, Eva Eckert, and Bihl, Donald E. ORAU Team. Principal Health Physicists. May 2004.

The factor of 2 was a professional judgment made to be favorable to claimants. As explained 

in the text, the risk of tritium intake was less during assembly than disassembly and fewer 

disassemblies took place from 1956 to 1972 than afterward.
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[56] Bihl, Donald E. ORAU Team. Principal Health Physicist. July 2004. 

This general statement was made by a former manager and by other Pantex employees 

during site visits. The documentation that would verify the statement is classified. However, 

Table 5-9 provides information that supports the statement in a general way.

[57] Bihl, Donald E. ORAU Team. Principal Health Physicist. April 2004. 

This statement is made because missed dose calculations assume chronic rather than acute 

intake as would be applicable to incident follow-up samples.

[58] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2004.

The mode of the triangular distribution for tritium missed-dose calculations is 6 mrem. It was 

used for unmonitored workers based on the premise that unmonitored workers were at less 

risk than monitored workers, especially for tritium because the source of tritium exposure 

came from handling tritium reservoirs during disassembly. The possibility of casual exposure 

to tritium during the disassembly operations by unmonitored workers was considered remote 

because of the security and other safety requirements of the work.
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 33 of 49

REFERENCES

Alley, C. D., 1990, “Bioassay Information,” letter to A. G. Ladino (U.S. Department of Energy, Amarillo 

Area Office), Mason & Hanger-Silas Mason Company, Pantex Plant, Amarillo, Texas, 

December 18. [SRDB Ref ID: 11092]

Author unknown, 1952, X-ray records of two early employees - 1952, Pantex Plant, Amarillo, Texas. 

[SRDB Ref ID: 14320]

Battelle, 1992, Technical Basis for the Internal Dosimetry Program and the DOE Pantex Facility, 

Rev. 1, Pantex Plant, Amarillo, Texas, February 14. [SRDB Ref ID: 13312]

BWXT Pantex, 2000, Internal Dosimetry Studies, Urinalysis in Historical Exposure Records System 

(HERS), Pantex Plant, Amarillo, Texas, March. [SRDB Ref ID: 14531]

BWXT Pantex, ca. 2000, presentation on W28 Program after Accident, Pantex Plant, Amarillo, Texas. 

[SRDB Ref ID: 26391]

BWXT Pantex, 2001, Internal Dosimetry Technical Basis & Quality Assurance Document, RSD-TBD0001, Rev. 03, Pantex Plant, Amarillo, Texas, August 23. [SRDB Ref ID: 11070]

BWXT Pantex, 2005, Information Obtained from DORMS (Dosimetry Records Management System) 

November 17, 2005 by Mark Prather (Pantex), Pantex Plant, Amarillo, Texas, November 17. 

[SRDB Ref ID: 20774]

CEP (Controls for Environmental Pollution), 1992, urinalysis report, Santa Fe, New Mexico, May 4. 

[SRDB Ref ID: 17084, p. 11]

Corley J. P., D. H. Denham, R. E. Jaquish, D. E. Michels, A. R. Olsen, and D. A. Waite, 1981, A 

Guide for: Environmental Radiological Surveillance at U.S. Department of Energy 

Installations, DOE/EP-0023, U.S. Department of Energy, Office of Operational Safety, 

Washington, D.C., July. [SRDB Ref ID: 70296] 

Davis, C. E., 1967, Health Protection Survey Report, Pantex Plant, December 13-15, 1967. [SRDB 

Ref ID: 13310]

DOE (U.S. Department of Energy), 1988, Radiation Protection for Occupational Workers, 

DOE 5480.11, Office of Environment, Safety and Health, Washington D.C., December 21. 

[SRDB Ref ID: 8073]

DOE (U.S. Department of Energy), 1992, U.S. Department of Energy Radiological Control Manual, 

DOE/EH-0256T, DOE N 5480.6, Office of Environment, Safety and Health, Washington D.C., 

June. [SRDB Ref ID: 11515]

DHHS (U.S. Department of Health and Human Services), 2011, DHHS Designation of Additional 

Members of the Special Exposure Cohort under the Energy Employees Occupational Illness 

Compensation Program Act of 2000, Designating a Class of Employees from Pantex Plant, 

Amarillo, Texas, U.S. Department of Health and Human Services, Office of the Secretary, 

Washington, D.C., December 21. [SRDB Ref ID: 116796]
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DHHS (U.S. Department of Health and Human Services), 2013, DHHS Designation of Additional 

Members of the Special Exposure Cohort under the Energy Employees Occupational Illness 

Compensation Program Act of 2000, Designating a Class of Employees from the Pantex Plant, 

Amarillo, Texas, U.S. Department of Health and Human Services, Office of the Secretary, 

Washington, D.C., September 30. [SRDB Ref ID: 128051]

Griffis, D. W., 1990, “File Holdings of Radiation Safety Department in Relation to [Privacy Act 

Information],” memorandum to J. A. Cross, Mason & Silas-Hanger Mason Company, Pantex 

Plant, Amarillo, Texas, July 25. [SRDB Ref ID: 11086]

ICRP (International Commission for Radiological Protection), 1959, Report of Committee II on 

Permissible Dose for Internal Radiation, Publication 2, Pergamon Press, Oxford, England. 

[SRDB Ref ID: 11106]

ICRP (International Commission on Radiological Protection), 1968, Report of Committee IV on 

Evaluation of Radiation Doses to Body Tissues from Internal Contamination due to 

Occupational Exposure, Publication 10, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1981, Limits for Inhalation of Radon 

Daughters by Workers, Publication 32, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1982, Limits for Intakes of Radionuclides 

by Workers, Part 3, Publication 30, Part 3, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22733] 

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

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22732]

ICRP (International Commission on Radiological Protection), 1995, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22731]

Johnson, J. R., and R. L. Hill, 1993, “Metabolism and Dosimetry of Tritium,” Health Physics, volume 

65, number 6, pp. 628–647.

Lang, C. M., 1990, “Medical Management Case of [Privacy Act Information] Following Tritium Incident 

May 17, 1989,” letter to B. L. Gage (U.S. Department of Energy, Amarillo Area Office), Mason 

& Hanger-Silas Mason Company, Pantex Plant, Amarillo, Texas, March 1. [SRDB Ref ID: 

11090]

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Universities Team, Oak Ridge, Tennessee, May 27. [SRDB Ref ID: 14542]

McFall, E. D., 1969, “Radon Film Badge Reports,” letter to B. Crisman (Mason & Hanger-Silas Mason 

Company), Eberline Instrument Corporation, Santa Fe, New Mexico, June 25. [SRDB Ref ID: 

13457]

MHSMC (Mason & Hanger-Silas Mason Company), 1978, soil, fecal, and urine sample results - 1978, 

Pantex Plant, Amarillo, Texas. [SRDB Ref ID: 14194]

MHSMC (Mason & Hanger-Silas Mason Company), 1983, collection of plutonium analysis procedures 

and results, Pantex Plant, Amarillo, Texas. [SRDB Ref ID: 14330] 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 35 of 49

MHSMC (Mason & Hanger-Silas Mason Company), 1991a, Analysis of Biological Samples ofr [sic] 

Uranium, Thorium and/or Plutonium, IOP D5466, Issue No. 1, Pantex Plant, Amarillo, Texas, 

January 4. [SRDB Ref ID: 12319]

MHSMC (Mason & Hanger-Silas Mason Company), 1991b, Internal Dose Assessment, IOP D5491, 

Issue No. 2, Pantex Plant, Amarillo, Texas, May 23. [SRDB Ref ID: 12549]

MMES (Martin Marietta Energy Systems), 1990, uranium bioassay samples, 1990, Pantex Plant, 

Amarillo, Texas. [SRDB Ref ID: 14196]

NCRP (National Council on Radiation Protection and Measurements), 1984a, Evaluation of 

Occupational and Environmental Exposures to Radon and Radon Daughters in the United 

States, Report 78, Bethesda, Maryland, May.

NCRP (National Council on Radiation Protection and Measurements), 1984b, Exposures from the 

Uranium Series with Emphasis on Radon and Its Daughters, Report 77, Bethesda, Maryland.

NCRP (National Council on Radiation Protection and Measurements), 1987, Exposure of the 

Population in the United States and Canada from Natural Background Radiation, Report 94, 

Bethesda, Maryland. [SRDB Ref ID: 22736]

NIOSH (National Institute for Occupational Safety and Health), 2008, SEC Petition Evaluation Report, 

Petition SEC-00068, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

August 5.

NIOSH (National Institute for Occupational Safety and Health), 2010, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October 5.

ORAUT (Oak Ridge Associated Universities Team), 2007,Pantex Plant – Site Description, ORAUTTKBS-0013-2, Rev. 02, Oak Ridge, Tennessee, May 8.

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Occupational Internal Dose, ORAUT-TKBS-0014-5, Rev 03, Oak Ridge, Tennessee, 

March 12.

ORAUT (Oak Ridge Associated Universities Team), 2014a, Internal Dose Reconstruction, ORAUTOTIB-0060, Rev. 01, Oak Ridge, Tennessee, September 8. 

ORAUT (Oak Ridge Associated Universities Team), 2014b, Pantex Plant – Occupational 

Environmental Dose, ORAUT-TKBS-0013-4, Rev. 02, Oak Ridge, Tennessee, January 3. 

Peterson, S-R., and P. A. Davis, 2002, “Tritium Doses from Chronic Atmospheric Releases: A New 

Approach Proposed for Regulatory Compliance,” Health Physics, volume 82, number 2, 

pp. 213–225. [SRDB Ref ID: 46071]

Reichert, M. M., 1994, “Pantex Plant Bioassay Uranium Samples Received 02-14-1994,” letter to 

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[SRDB Ref ID: 14537, p. 14]
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 36 of 49

UNC Geotech, 1990, Results of the U.S. Department of Energy Indoor Radon Study, Volume 1, 

DOE/ID/12584-75 Vol. 1, CNG/GJ-TP-1, U.S. Department of Energy, Grand Junction, 

Colorado, August. [SRDB Ref ID: 12287]

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and Effects of Ionizing Radiation, UNSCEAR 1993 Report to the General Assembly, with 

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11596]
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 37 of 49

GLOSSARY

age or aging

In relation to reactor fuel and mixtures of plutonium isotopes, time since the step in the 

refinement process that separates americium from the mixture.

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium.

dose of record

(1) Dose records that the U.S. Department of Energy provided to the National Institute for 

Occupational Safety and Health as part of each worker’s file. (2) Individual recorded dose 

such as that on a dosimetry card or in a dosimetry database.

equilibrium factor (F)

In relation to the potential alpha energy of radon and its progeny in air, the ratio of the 

equilibrium equivalent concentration to the actual activity concentration of radon. See 

potential alpha energy concentration. 

Gravel Gertie

Facility with the distinguishing characteristic of having blow-out roof panels overlain with gravel 

to dissipate the pressure surge and energy of a conventional high-explosive detonation. This 

design was developed to allow the energy of the blast to be dissipated while minimizing the 

spread of contamination of any radioactive material present. Also called Gertie.

hydroshot

Detonation of a mixture of explosives and depleted uranium used as a quality control

technique for measuring the performance of plastic-bonded explosives.

Parameters 1, 2, and 3

Columns in the Interactive RadioEpidemiological Program template where the dose 

reconstructor enters the calculated doses and expected variance based on the distribution 

applied. Multiple entries based on year of employment, type of radiation, appropriate energy 

ranges, and internal and external exposures are possible.

potential alpha energy concentration (PAEC)

Kinetic energy in units of working levels potentially released in a unit volume of air by alpha 

particles emitted by the short-lived radioactive progeny of 222Rn (218Po, 214Pb, 214Bi, and 214Po) 

and 220Rn (216Po, 212Pb, 212Bi, 212Po). See potential alpha energy exposure and working level. 

potential alpha energy exposure (PAEE) 

Average potential alpha energy concentration to which a worker is exposed multiplied by the 

time of exposure in working months of 170 hours (units of working level months). PAEE is the 

potential alpha energy concentration multiplied by time. See potential alpha energy 

concentration and working level month. 

progeny

Nuclides that result from decay of other nuclides. Also called decay products and formerly 

called daughter products.

radon (Rn)

Radioactive gaseous element with atomic number 86. Radon is a decay product (progeny) of 

other radioactive elements such as thorium and radium.
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 38 of 49

thoron

Informal name for 220Rn.

U.S. Atomic Energy Commission (AEC)

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

working level (WL)

Unit of concentration in air of the short-lived decay products of 222Rn (218Po, 214Pb, 214Bi, and 

214Po) and 220Rn (216Po, 212Pb, 212Bi, 212Po) defined as any combination of the short-lived 

radioactive progeny of radon or thoron in 1 liter of air, without regard to the degree of 

equilibrium, that results in the ultimate emission of 130,000 megaelectron-volts of alpha 

energy; 1 WL equals 2.083 × 10-5 joules per cubic meter. See potential alpha energy 

concentration. 

working level month (WLM)

Unit of exposure to radon progeny defined as exposure for 1 working month (170 working 

hours) to a potential alpha energy concentration of 1 WL; 1 WLM equals 1 WL times 

170 hours, which is 0.00354 joule-hours per cubic meter. See potential alpha energy 

exposure and working level. 
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 39 of 49

ATTACHMENT A

EXAMPLES OF PANTEX SITE BIOASSAY DATA FILES
 

ATTACHMENT B 

RESULTS OF 1990 RADON MEASUREMENTS

Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 40 of 49

Table B-1. Results of 137 radon measurements in 1990 (UNC Geotech 1990).

RPIS Bu-InsBldg CodeBuilding 

number

General 

description

Gross 

area 

(ft2)

Number 

of floors

Radon 

(pCi/L)

Dupl. 

radon 

(pCi/L)

Install 

date

Retrieve 

date Room

01001DOE 

BLDG

DOE Building DOE Building (12-

36) is of brick 

construction (with 

bricks from the 

Panhandle area). 

8.1 01/10/90 02/16/90 Main Office

0100111-48 11-48 11-48 3,200 1 7.8 01/10/90 02/16/90 Maintenance Shop

0100112-104 12-104 

East

12-104 is bay 

building

99,680 2 7.1 01/10/90 02/16/90 Bay 5

0100112-66 12-66 12-66 is special 

nuclear material 

(SNM) warehouse

25,900 1 5.4 01/10/90 02/16/90 Center Of building

01001FS-01 FS-1 FS-01 is an earth 

covered storage 

facility for HE

5,364 1 5.2 5.9 01/10/90 02/16/90 Break room

0100112-23 12-23 12-23 3,200 1 4.5 01/10/90 02/16/90 North wall middle

0100112-15 12-15 12-15 16,800 1 4.2 01/10/90 02/16/90 Training Room 103

0100112-60 12-60 12-60 is Mass 

Properties Facility

8,600 1 4.1 01/10/90 02/16/90 Office - Vault

0100112-79 12-79 12-79 is warehouse/

loading dock

28,700 1 3.2 01/10/90 02/16/90 South Warehouse Area

0100112-104 12-104 12-104 is bay 

building

99,680 2 3 01/10/90 02/16/90 106f R Collins Office

0100112-15 12-15 12-15 16,800 1 2.8 01/10/90 02/16/90 Training Graphic Arts

0100112-58 12-58 12-58 is bay building 2,600 1 2.8 01/10/90 02/16/90 East Wall Between Bay 

4&5

0100112-15 12-15 12-15 16,800 1 2.6 01/10/90 02/16/90 Training Office

0100112-44 12-44-4 12-44 is Gravel 

Gertie cell building

27,100 1 2.6 01/10/90 02/16/90 Round Room

0100112-6 12-6 12-6 23,700 1 2.6 01/10/90 02/16/90 Room 131 Quality 

Records

0100112-14 12-14 12-14 900 1 2.4 01/10/90 02/16/90 Office

0100112-26 12-26 Tooling 

Warehouse

12-26 is bay building 87,500 1 2.4 01/10/90 02/16/90 Tooling Warehouse Office

0100112-26 12-26 12-26 also has pit 87,500 1 2.3 01/10/90 02/16/90 Bay 28
 

ATTACHMENT B 

RESULTS OF 1990 RADON MEASUREMENTS (continued)

Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 41 of 49

RPIS Bu-InsBldg CodeBuilding 

number

General 

description

Gross 

area 

(ft2)

Number 

of floors

Radon 

(pCi/L)

Dupl. 

radon 

(pCi/L)

Install 

date

Retrieve 

date Room

vault

0100112-35 12-35 12-35 13,400 1 2.3 2.1 01/10/90 02/16/90 Area Mechanics Office

0100111-17 11-17 12-17is bay building 6,700 1 2.2 01/10/90 02/16/90 Bay 7 Lab

0100112-1 12-1 12-1 is cafeteria/

change room

27,600 2 2.2 01/10/90 02/16/90 Lepor Colony

0100112-5 12-5 12-5 74,400 1 2.2 01/10/90 02/16/90 Electric Shop Office

0100112-86 12-86 12-86 is an Inert 

Assembly and Test 

Facility

2.2 01/10/90 02/16/90 Electrical Testing Area

0100111-7 11-7 12-7 34,100 1 2.1 2.2 01/10/90 02/16/90 Break room

0100112-61 12-61 12-61 24,000 1 2.1 01/10/90 02/16/90 Warehouse Area

0100112-44 12-44 12-44 is Gravel 

Gertie cell building

27,100 1 2 01/10/90 02/16/90 Cell 8

0100112-64 12-64 12-64 is bay building 32,000 1 2 01/10/90 02/16/90 Bay 9

01001STATI

ON 30

Zone 4

Station 30

4-30 is underground 

igloo SNM storage 

building

2 01/10/90 02/16/90 Control Room

0100112-28 12-28 3,500 1 1.9 01/10/90 02/16/90 Quality Hallway

0100112-37 12-37 22,700 1 1.9 01/10/90 02/16/90 Room 120 Control Room

0100112-42 12-42 

Radiation 

Safety

47,400 2 1.9 01/10/90 02/16/90 12-42 Cr

0100112-84 12-84 Bay 1 1 1.9 01/10/90 02/16/90 Bay 13

0100111-29 11-29 4,200 1 1.8 01/10/90 02/16/90 Office

0100112-37 12-37 22,700 1 1.8 01/10/90 02/16/90 Room 121 Tech Doc

0100112-37 12-37 22,700 1 1.8 01/10/90 02/16/90 Room 112 Mail Room

0100112-5 12-5 74,400 1 1.8 01/10/90 02/16/90 Plant Eng Annette 

Covington

0100112-5 12-5 74,400 1 1.8 1.8 01/10/90 02/16/90 Plant Design Eng

0100116-2 16-2 Courier 20,072 1 1.8 01/10/90 02/16/90 114 Break room

0100110-9 10-9 15,500 1 1.7 01/10/90 02/16/90 Office

0100111-27 11-27 5,100 2 1.7 01/10/90 02/16/90 Room 119

0100112-50 12-50 1,400 1 1.7 01/10/90 02/16/90 West Of 12-50 Door

0100112-75 12-75 21,862 1 1.7 01/10/90 02/16/90 Desk Lieutenants Office

0100112-99 12-99 60,716 1 1.7 01/10/90 02/16/90 Break Room
 

ATTACHMENT B 

RESULTS OF 1990 RADON MEASUREMENTS (continued)

Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 42 of 49

RPIS Bu-InsBldg CodeBuilding 

number

General 

description

Gross 

area 

(ft2)

Number 

of floors

Radon 

(pCi/L)

Dupl. 

radon 

(pCi/L)

Install 

date

Retrieve 

date Room

0100111-20 11-20 16,600 1 1.6 01/10/90 02/16/90 Office South Wall

0100112-16 12-16 5,000 1 1.6 1.1 01/10/90 02/16/90 Plastic Shop Office

0100112-2 12-2 Safety 13,456 1 1.6 01/10/90 02/16/90 Dosimetry Lab Room 157

0100112-2B 12-2B 3,220 1 1.6 01/10/90 02/16/90 North Wall By Clock

0100112-42 12-42 47,400 2 1.6 01/10/90 02/16/90 South Vault

0100112-6 12-6 23,700 1 1.6 01/10/90 02/16/90 Room 103 Cafeteria

0100112-84 12-84 1 1.6 01/10/90 02/16/90 125d

0100112-36 12-36 29,400 1 1.5 01/10/90 02/16/90 Emergency Preparedness

0100112-52B 12-52B1 1.5 01/10/90 02/16/90 Meteorology

0100112-9 12-9 18,500 3 1.5 01/10/90 02/16/90 HE Side

0100112-96 12-96 Gravel Gertie 7,865 1 1.5 01/10/90 02/16/90 Round Room

0100112-99 12-99 Bay 60,716 1 1.5 01/10/90 02/16/90 Bay 7

0100112-26 12-26 Bay 87,500 1 1.4 01/10/90 02/16/90 Bay 30

0100112-107 12-107 

South

10,000 1 1.4 01/10/90 02/16/90 By C.L. Saban’s Office

0100112-44E 12-44-E 12-44 is Gravel 

Gertie cell building

1,900 1 1.4 01/10/90 02/16/90 Marion Everett’s Office

0100112-6 12-6 23,700 1 1.4 01/10/90 02/16/90 Room 112 Tom Folks

0100112-68 12-68 35,900 1 1.4 01/10/90 02/16/90 Machine Shop Office

0100112-69 12-69 9,800 1 1.4 01/10/90 02/16/90 Emmett Hallway

0100111-5 11-5 Bay 9,000 2 1.3 01/10/90 02/16/90 Control Bay

0100112-2 12-2 13,456 1 1.3 2 01/10/90 02/16/90 Medical Office

0100112-24 12-24 South Bay 1.3 01/10/90 02/16/90 Bay 27

0100112-49 12-49 3,900 1 1.3 01/10/90 02/16/90 Electronics Room

0100112-6 12-6 23,700 1 1.3 01/10/90 02/16/90 104 Room Standards

0100116-12 16-12 28,500 2 1.3 01/10/90 02/16/90 Employment

0100112-61 12-61 24,000 1 1.2 01/10/90 02/15/90 Break Room

0100112-107 12-107 North 10,000 1 1.2 01/10/90 02/16/90 Preventive Maint Section

0100112-11 12-11 2,900 1 1.2 01/10/90 02/16/90 Data Management

0100112-42 12-42 47,400 2 1.2 1 01/10/90 02/16/90 Upstairs Assembly Ops

0100112-6 12-6 23,700 1 1.2 01/10/90 02/16/90 Room 121 Elaine Miller

0100112-64 12-64 Bay 32,000 1 1.2 1.3 01/10/90 02/16/90 Bay 15

0100112-69 12-69 9,800 1 1.2 01/10/90 02/16/90 DOE Office

0100112-9 12-9 18,500 3 1.2 01/10/90 02/16/90 Office

0100112-97 12-97B 10,000 1 1.2 01/10/90 02/16/90 By Refrigerator in Break 
 

ATTACHMENT B 

RESULTS OF 1990 RADON MEASUREMENTS (continued)

Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 43 of 49

RPIS Bu-InsBldg CodeBuilding 

number

General 

description

Gross 

area 

(ft2)

Number 

of floors

Radon 

(pCi/L)

Dupl. 

radon 

(pCi/L)

Install 

date

Retrieve 

date Room

Room

0100112-98 12-98 Gravel Gertie 34,358 1 1.2 1.2 01/10/90 02/16/90 Cell 2

0100112-21 12-21 

Gas Lab

29,300 2 1.1 01/10/90 02/16/90 Break Area

0100111-18 11-18 Control

Room

1,500 1 1.1 01/10/90 02/16/90 Control Room

0100111-2 11-2 9,600 2 1.1 01/10/90 02/16/90 110

0100111-51 11-51 11,600 1 1.1 01/10/90 02/16/90 Office

0100112-32 12-32 

South Side

7,600 1 1.1 01/10/90 02/16/90 Above Phone on Ramp

0100112-98 12-98-3 Gravel Gertie 34,358 1 1.1 0.8 01/10/90 02/16/90 Round Room

0100111-36 11-36 5,000 2 1 01/10/90 02/16/90 Office

0100111-50 11-50 22,151 1 1 01/10/90 02/16/90 Room 110 Office

0100112-100 12-100 4,360 1 1 01/10/90 02/16/90 Environmental Protection

0100112-11A 12-11A 5,200 1 1 01/10/90 02/16/90 Quality Hallway

0100112-5 12-5 74,400 1 1 01/10/90 02/16/90 General Stores Office

0100112-52B 12-52B 1 01/10/90 02/16/90 Meteorology

0100112-

52C

12-52C 3,600 1 1 0.9 01/10/90 02/16/90 Meteorology

0100112-6 12-6 23,700 1 1 01/10/90 02/16/90 Assy Eng Office

0100112-6 12-6 23,700 1 1 01/10/90 02/16/90 Room 700 Stoddard

0100112-61 12-61 24,000 1 1 01/10/90 02/16/90 Office Area

0100112-82 12-82 Bay 6,800 1 1 01/10/90 02/16/90 E-Bay Office

01001STATI

ON C

Station C 1 01/10/90 02/16/90 West Wall By Exit

0100112-21 12-21 29,300 2 1 01/10/90 02/16/90 X-Ray Office

0100112-2B 12-2B 3,220 1 0.9 01/10/90 02/16/90 South Wall - Nancy’s 

Office

0100112-104 12-104 Bay 99,680 2 0.9 01/10/90 02/16/90 Bay 13

0100112-

104-EAS

12-104 East Bay 0.9 01/10/90 02/16/90 Bay 2

0100112-106 12-106 5,400 1 0.9 01/10/90 02/16/90 Across From Room 105 

Janitor

0100112-24 12-24 North Bay 0.9 01/10/90 02/16/90 Bay 10

0100112-31 12-31 Bay 7,600 1 0.9 0.8 01/10/90 02/16/90 Bay 3 Outside
 

ATTACHMENT B 

RESULTS OF 1990 RADON MEASUREMENTS (continued)

Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 44 of 49

RPIS Bu-InsBldg CodeBuilding 

number

General 

description

Gross 

area 

(ft2)

Number 

of floors

Radon 

(pCi/L)

Dupl. 

radon 

(pCi/L)

Install 

date

Retrieve 

date Room

0100112-64 12-64 32,000 1 0.9 01/10/90 02/16/90 D&I Office

0100112-99 12-99 60,716 1 0.9 01/10/90 02/16/90 105-F Manufacturing 

Office

0100112-84 12-84-

East

1 0.8 01/10/90 02/16/90 Break Room

0100112-84 12-84 Bay 1 1 0.8 01/10/90 02/16/90 Bay 4

0100112-84 12-84 1 1 0.8 01/10/90 02/16/90 Break Room

0100112-84 12-84 Bay 1 1 0.8 1.3 01/10/90 02/16/90 Bay 7

0100112-101 12-101 

Portable Maint

5,334 1 0.8 01/10/90 02/16/90 By Sign-Out Board

0100112-102 12-102 5,778 1 0.8 01/10/90 02/16/90 Tech Applications

0100112-103 12-103 23,608 1 0.8 01/10/90 02/16/90 Smoking Area

0100112-104 12-104 

West

99,680 2 0.8 01/10/90 02/16/90 128f Manufacturing Office

0100112-111 12-111 7,416 1 0.8 01/10/90 02/16/90 Carpenter Shop

0100112-112 12-112 6,525 1 0.8 01/10/90 02/16/90 Camera Room

0100112-17 12-17 32,500 2 0.8 01/10/90 02/16/90 Break Area

0100112-19 12-19 

EAST

32,500 2 0.8 01/10/90 02/16/90 Break Area East Side

0100112-20 12-70 0.8 01/10/90 02/16/90 Cafeteria

0100112-3 12-3 2,000 1 0.8 01/10/90 02/16/90 Transportation

0100112-35 12-35 13,400 1 0.8 01/10/90 02/16/90 Utilities Console Room

0100112-39 12-39 Fire

Department

8,200 1 0.8 01/10/90 02/16/90 Sleeping Room

0100112-41A 12-41A 3,000 1 0.8 01/10/90 02/16/90 North Wall

0100112-42A 12-42A 19,900 1 0.8 01/10/90 02/16/90 Outer Wall by Sandia 

Sign

0100112-5 12-5 74,400 1 0.8 01/10/90 02/16/90 Master Mechanics

0100112-59 12-59 8,300 1 0.8 01/10/90 02/16/90 Chem Lab Office

0100112-5C 12-5C 21,700 2 0.8 01/10/90 02/16/90 Sheet Metal Shop

0100112-84 12-84 Bay 1 0.8 01/10/90 02/16/90 Bay 12

0100112-86 12-86 0.8 01/10/90 02/16/90 86-2e-5

0100112-86 12-86 0.8 01/10/90 02/16/90 206s Upstairs Assembly 

Ops Office

0100112-86 12-86 Bay 0.8 01/10/90 02/16/90 Bay 10
 

ATTACHMENT B 

RESULTS OF 1990 RADON MEASUREMENTS (continued)

Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 45 of 49

RPIS Bu-InsBldg CodeBuilding 

number

General 

description

Gross 

area 

(ft2)

Number 

of floors

Radon 

(pCi/L)

Dupl. 

radon 

(pCi/L)

Install 

date

Retrieve 

date Room

0100112-97 12-97A 10,000 1 0.8 01/10/90 02/16/90 By Clock in Hallway

0100112-97 12-97C 10,000 1 0.8 01/10/90 02/16/90 Above Fire Ext by Copier

0100112-99 12-99 Bay 60,716 1 0.8 01/10/90 02/16/90 Bay 6

0100116-1 16-1 VMF 54,200 1 0.8 01/10/90 02/16/90 Office

0100116-12 16-12 28,500 2 0.8 01/10/90 02/16/90 Purchasing

01001STATI

ON B

Station B 0.8 01/10/90 02/16/90 East Wall Center

01001Trailer Parking Lot 0.8 01/10/90 02/16/90 West Trailer from 12-2

01001Trailer Parking Lot 0.8 01/10/90 02/16/90 East Trailer from 12-2
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 46 of 49

ATTACHMENT C

ASSESSING TRITIUM INTAKES

Tritium Intakes, before 1983

To convert from tritium dose back to uptake for 1972 to 1982, dose reconstructors should use a dose 

conversion factor of 3.5 μCi/mrem. This conversion results from the approach in ICRP Publication 10 

(ICRP 1968) explained in NUREG-0938 (Brodsky 1983; 1.5 mCi = 425 mrem). It assumes an acute 

intake and a quality factor of 1.7 for tritium beta particles. Dose reconstructors should use a 

lognormal distribution with a geometric standard deviation (GSD) of 3 (ORAUT 2014a) and should 

apply this same conversion to recorded tritium doses for years before 1972 if they encounter any such 

doses [53]. Because this conversion produces a dose lower than the recorded dose by nearly 44% 

when input into IMBA, it is permissible to use the doses as recorded for likely noncompensable, 

maximum internal dose cases. 

Tritium Intakes, 1983 to 1988

Ikenberry (1983) described the uptake to dose calculation method used at that time, which was based 

on ICRP Publication 2 (ICRP 1959) but used a quality factor of 1: 

6 doserate in rem d / 8.12 10 q − = × (C-1)

where q is the uptake in microcuries. Total dose was determined by integrating over the dose rate 

curve. For an acute exposure: 

4 dose in rem 1.4 10 q − = × (C-2)

and for chronic exposure: 

6 dose in rem 8.12 10 qt − = × (C-3)

where t is the period of chronic exposure in days. 

Equation C-2 produces a conversion factor of 7.1 μCi/mrem and, assuming a 365-day exposure, 

Equation C-3 produces a conversion factor of 0.33 μCi/mrem. Ikenberry (1983) does not specify a 

particular intake scenario, so it is not known which of the two equations produced the reported doses. 

Both were probably used to fit whichever intake scenario was appropriate for each worker but, for the 

purpose of establishing a default intake, Equation C-2 is favorable to claimants along with the 

assumption of a lognormal distribution with a GSD of 3 (ORAUT 2014a). 

Tritium Intakes, 1989 to Present

In 1989, DOE Order 5480.11 (DOE 1988) required sites to convert to ICRP 30 internal dose 

methodology (ICRP 1982). Spot-checking of case files showed that Pantex used both acute and 

chronic assumptions for different cases [54]. For instance, a May 1991 letter to a worker’s file (case 

xxxx) states that a 0.5-mL aliquot was analyzed by liquid scintillation and that “Doses were calculated 

by the use of computer algorithms incorporating an assumption of a single intake 30 day before the 

measurement.” Two other cases (xxxx and xxxx), dated 1991 and 1993, respectively, showed 

outputs from the REMedy© internal dosimetry computer code, and both assumed a chronic intake 

mode to calculate the dose. 

For the acute intake scenario, BWXT Pantex (1992) provided Equation C-4 to convert from calculated 

dose to uptake assuming a single exponential retention curve with a 10-day retention half-time: 

3 dose in mrem 1.3 10 C0

− = × (C-4)
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 47 of 49

ATTACHMENT C 

ASSESSING TRITIUM INTAKES (continued)

where C0 is the initial body water concentration in disintegrations per minute per milliliter and the dose 

is in millirem. The concentration is distributed in 42,000 mL of body water, so the uptake in 

disintegrations per minute is 42,000C0. Therefore: 

uptake in dpm = (42,000)(dose)/(1.3 × 10-3) = (3.23 × 107)(dose in mrem) (C-5a)

and

7 uptake in pCi (1.46 10 )( ) = × dose in mrem (C-5b)

For the chronic intake scenario, the same document provides Equation C-6 for calculating dose from 

a urine sample:

( ) 5 3

e dose in mrem 8.7 10 1.3 10 t C − − = × +×     (C-6)

where Ce is the urine concentration in disintegrations per minute per milliliter. Monthly sampling was 

the normal frequency for workers potentially exposed to tritium [55] so, with a t of 30 days: 

3

e dose in mrem 3.9 10 C− = × (C-7)

Distributing the tritium in 42,000 mL of body water gives:

7 uptake in dpm (1.07 10 )( ) = × dose in mrem (C-8a)

6 uptake in pCi (4.82 10 )( ) = × dose in mrem (C-8b)

The same equation for an intake period other than 30 days is:

( )

4

5 3

(1.89 10 )( ) 8.7 10 1.3 10dose in mrem uptake in pCi

t − −

× = × +×

(C-9)

Equation C-5 differs from Equation C-8 by a factor of 3, so if the doses are large it could be important 

to know if the original calculation of the recorded tritium dose assumed the chronic or acute scenario. 

It is implied, although not explicitly stated, in the Pantex procedure Internal Dose Assessment

(MHSMC 1991b) that chronic intakes were applied to workers receiving routine monthly bioassay and 

an acute intake scenario was applied to workers receiving termination or infrequent bioassay. 

Disassemblies were occurring more often than assemblies during this period, so chronic intakes were 

more likely [56]. 

Equation C-8 and IMBA produce a slightly smaller dose than originally recorded. For instance, a 

recorded dose of 10 mrem for a monthly sample results in a recalculated dose of 7.4 mrem. 

Therefore, if the only information available is the recorded tritium dose, and it is evident from the 

records that the worker was on a monthly sampling frequency, dose reconstructors can use the 

recorded dose directly for the likely noncompensable, maximum internal dose approach. Therefore, 

for tritium doses in the records for 1989 to the present, the dose reconstructor should apply the 

following steps:
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 48 of 49

ATTACHMENT C 

ASSESSING TRITIUM INTAKES (continued)

If the worker’s file provides (in order of priority):

• Actual bioassay results and an acute intake date or chronic exposure period, use that 

information to determine dose; or

• Dose and a chronic intake period, use the recorded dose unless better accuracy is required, in 

which case use Equation C-8 for monthly sampling frequency or Equation C-9 for another 

exposure period; or

• Dose calculated from a termination sample or single sample when the worker was not on a 

monthly routine, use Equation C-5.

Regardless of the step used to determine the dose, dose reconstructors should assume a lognormal 

distribution with a GSD of 3 (ORAUT 2014a).

Tritium Missed Dose, 1972 to 1988

See Sections 5.2.1.1.5 and 5.2.1.1.6 for unmonitored worker discussions. 

Tritium Missed Dose, 1989 to Present

The 1991 internal dose assessment procedure (MHSMC 1991b) lists urinalysis results above for 

which a dose assessment is necessary (Table C-1). By inference, results below the values in the 

table did not need dose assessment because, as the procedure states, “The activities cited below 

have been calculated to result in 1 mrem of exposure based on methods described in ANSI N13.14” 

(HPS 1983). This TBD analysis has not established how far back in time these screening values were 

in place, but it is plausible that Pantex started using them in 1989 with the implementation of DOE 

Order 5480.11 (DOE 1988). Assuming a chronic intake for the monthly sample period and a 

urinalysis result of 0.135 μCi/L at the end of the period, the IMBA Version 3.1.99 calculates a daily 

uptake of 4.28 × 105 pCi/d (or 0.87 mrem to all organs). The daily uptake rate and the total potentially 

missed dose are dependent on the number of monitoring periods, as listed in Table C-2. However, 

the potentially missed dose is reasonably close to 1 mrem/30 d, so it is favorable to claimants and 

efficient to use 1 mrem for each monitoring period. Therefore, if a worker’s record shows, for 

example, 2 mrem for three monitoring periods in a year and zero dose for the remaining nine periods, 

the unrecorded dose would be 9 mrem and the recorded dose would be 6 mrem (or could be adjusted 

using Equation C-8). This dose would apply equally to all organs [see, for instance, ICRP 1995, 

Table 5.1.2(d)]. [Using the formula from ANSI Standard N13.14-1983 (HPS 1983), the acute intake in 

Table C-2 results in a 1-mrem dose if the time after intake is 7 days, so Pantex must have used the 

0.357-μCi/L value for incident follow-up samples. It would not be appropriate for potentially missed 

dose estimation] [57]. 

If the annual tritium dose is recorded as zero but it appears that bioassay occurred, the dose should 

be assigned as a triangular distribution with a minimum of 0 mrem, a mode of 6 mrem, and a 

maximum of 12 mrem [58]. 

Table C-1. Tritium urinalysis screening levels, 1991.

a

Analysis period Urine tritium concentration (μCi/L)

Termination 1.35 E-2

Monthly 1.35 E-1

Acute 3.57 E-1

a. From MHSMC (1991b); assumed to apply to 1989 to present.
Document No. ORAUT-TKBS-0013-5 Revision No. 03 Effective Date: 11/13/2014 Page 49 of 49

ATTACHMENT C 

ASSESSING TRITIUM INTAKES (continued)

Table C-2. Potentially missed intake and dose from monthly 

sampling for chronic intake of tritium, 1989 to Present.

a

Monthly periods 

missed

Daily intake

( × 105 pCi)

Total missed dose 

(mrem) (all organs)

1 4.28 0.869

2 3.73 1.51

3 3.62 2.23

4 3.58 2.93

5 3.56 3.64

6 3.55 4.37

7 3.54 5.08

8 3.54 5.82

9 3.54 6.55

10 3.54 7.27

11 3.54 7.98

12 3.54 8.73

a. Based on 0.135 μCi/L excretion at end of total period from Table 5-4.