ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 55

 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 Title:

Pantex Plant – Occupational Internal Dose

Document Number: ORAUT-TKBS-0013-5 

Revision: 00 PC-1 

Effective Date: 02/08/2007

Type of Document: TBD

Supersedes: None

Subject Expert(s): Eva Eckert Hickey, Donald E. Bihl, and Daniel J. Strom

Site Expert(s): N/A

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

Jack J. Fix, Document Owner

Approval: Signature on File Approval Date: 01/25/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 01/26/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 01/31/2007

Kate Kimpan, Project Director

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

James W. Neton, Associate Director for Science
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 2 of 55

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. 

Jack J. Fix, Document Owner

Signature on File___________________________ ___05/29/2006 

John M. Byrne, Document Owner

Signature on File___________________________ ___04/26/2006 

Edward F. Maher, Task 5 Manager

Signature on File____________________________ __05/26/2006 

Kate Kimpan, Project Director

Signature on File________ ________________ 05/31/2006 

James W. Neton, Associate Director for Science

Signature on File ______________ _ 07/27/2006 

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.
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 3 of 55

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

5.1 Introduction ..............................................................................................................................7

5.1.1 Purpose........................................................................................................................8

5.1.2 Scope...........................................................................................................................8

5.1.3 History of Internal Dosimetry ........................................................................................8

5.1.4 Current Internal Dosimetry Practices ............................................................................9

5.2 Radionuclides with Potential for Internal Dose ....................................................................... 10

5.2.1 Tritium ........................................................................................................................10

5.2.1.1 Internal Assessment for Tritium During Routine Operations.......................... 12

5.2.1.2 Tritium Release Event in 1989...................................................................... 19

5.2.2 Uranium......................................................................................................................19

5.2.2.1 Background .................................................................................................. 19

5.2.2.2 Uranium Reporting Levels or Minimum Detectable Activities ........................ 21

5.2.2.3 Uranium Intakes from Assembly/Disassembly of Weapons for 

 Unmonitored Workers ................................................................................... 22

5.2.2.4 Depleted Uranium Intakes from Machining ................................................... 24

5.2.2.5 Depleted Uranium Intakes from Burning of Contaminated 

 High Explosives ............................................................................................ 25

5.2.2.6 Intakes from Hydroshots ............................................................................... 28

5.2.3 Thorium ......................................................................................................................34

5.2.4 Plutonium and Plutonium Incidents.............................................................................36

5.2.4.1 Plutonium General ........................................................................................ 36

5.2.4.2 1961 Cell Incident ......................................................................................... 38

5.2.4.3 1978 Storage Cylinder Incident..................................................................... 40

5.3 Assessing Occupational Internal Dose from Elevated Radon................................................. 40

5.3.1 Dose from Radon-222 Progeny ..................................................................................41

5.3.2 Radon Monitoring at DOE Facilities............................................................................41

5.3.3 Underground Buildings ...............................................................................................41

5.3.4 Radon Concentrations................................................................................................41

5.3.5 Working Level-Months................................................................................................47

5.4 Intake Summary.....................................................................................................................47

References ........................................................................................................................................49

Glossary ............................................................................................................................................54
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 4 of 55

LIST OF TABLES

TABLE TITLE

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

PAGE

10

5-2 Job titles and descriptions of work with possibility for occupational intake.............................. 11

5-3 Tritium dose data ................................................................................................................... 14

5-4 Tritium urinalysis screening levels, 1991................................................................................ 17

5-5 Potentially missed intake and dose from monthly sampling for chronic intake 

of tritium, 1989 to present....................................................................................................... 17

5-6 Tritium uptakes for unmonitored workers ............................................................................... 18

5-7 Uranium dose to workers ....................................................................................................... 21

5-8 History of uranium urinalysis .................................................................................................. 21

5-9 Retirements in the DOE complex ........................................................................................... 25

5-10 Recorded numbers of hydro test shots and other detonations at Firing Site 5 

or other sites ..........................................................................................................................31

5-11 Estimated number of detonations potentially involving DU dispersion and 

intakes ...................................................................................................................................32

5-12 Thorium dose to workers........................................................................................................ 35

5-13 Number of workers on plutonium bioassay, 1991 to 2002 ...................................................... 36

5-14 Activity composition of weapons-grade plutonium mixtures from Hanford .............................. 37

5-15 Plutonium incident air sample counts, 1961 ........................................................................... 38

5-16 Results of 137 radon measurements in 1990 ......................................................................... 43

5-17 Summary statistics of 1990 radon measurements.................................................................. 46

5-18 Summary of 1969 radon measurements in Cells 1 to 6 .......................................................... 46

5-19 Summary of default intakes.................................................................................................... 48

LIST OF FIGURES

FIGURE TITLE

5-1 Air sample log-probability fit inside hydroshot bunkers ...........................................................

PAGE

29

5-2 Air sample log-probability fit outside hydroshot bunkers......................................................... 30

5-3 Absolute differences between duplicate radon measurements............................................... 46
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 5 of 55

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

Bq becquerel

cd calendar day

CEDE committed effective dose equivalent

CFR Code of Federal Regulations

Ci curie

cm centimeter

cpm counts per minute

d day

DAC derived air concentration

DOE U.S. Department of Energy

DORMS Dosimetry Records Management System

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EU enriched uranium

g gram

gal gallon

GSD geometric standard deviation

HE,50 committed effective dose equivalent

HE high explosive

HEPA high-efficiency particulate air

HERS Historical Exposure Records System

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

IREP Interactive RadioEpidemiological Program

L liter

m meter

mCi millicurie

MDA minimum detectable activity

MeV megaelectron-volt, 1 million electron-volts

min minute

ml milliliter

mrem millirem

NCRP National Council on Radiological Protection and Measurements

NDE nondestructive examination

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH-Office of Compensation Analysis and Support Claims Tracking System
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 6 of 55

PAEC potential alpha energy concentration

PAEE potential alpha energy exposure

pCi picocurie

POC probability of causation

RAMS Radiation Alarm Monitoring System

RST Radiation Safety Technician

s second

SNM special nuclear material

Sv sievert

TBD technical basis document

U.S.C. United States Code

wd working day

WL working level

WLM working level-month

yr year

μCi microcurie

μg microgram

μm micrometer

§ section, sections
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 7 of 55

5.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

As noted above, the statute 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 contains an exclusion with respect to the Naval Nuclear Propulsion Program, 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 dosimetry monitoring results are considered valid for use in 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:

] 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) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 8 of 55

5.1.1

This 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.

Purpose

5.1.2

Activities at the Pantex Plant with the potential for airborne contamination occurred 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. 

Scope

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 and 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. 

5.1.3

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

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 

the facility after the event. A 1967 report describing 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. 

History of Internal Dosimetry

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

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

triggered special bioassays before 1991. 

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 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 9 of 55

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

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

data were not available for those months. 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 will contain the HERS data, but those data do not include 

bioassay data. 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 collected since 1986. No plutonium or thorium bioassay 

results were found in the NOCTS files. Therefore, this TBD provides other approaches for 

determining intakes of radionuclides that dose reconstructors can use when bioassay data are not 

available.

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.

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) or 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. 

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

described the type of job performed rather than the job title. The job titles have changed over the 

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

5.1.4

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, the RadCon Manual, and 10 C.F.R. pt.

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.

Current Internal Dosimetry Practices

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 

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

effective dose equivalent (CEDE or HE,50). 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 C.F.R. § 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 C.F.R. § 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 

uses occurrence-based bioassay sampling to confirm intakes and calculate internal doses. 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 d of an 

assessment of airborne exposures exceeding 4 DAC-hr for an individual actinide.
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 10 of 55

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

Year

Workers 

monitored 

for tritium

Workers 

monitored 

for uranium

Workers 

monitored 

for thorium

Workers 

monitored 

for 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

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) discusses the first four 

radionuclides; the latest version of Section 2 of this Pantex Site Profile (ORAUT 2004a) discusses 

processes and locations where radioactive material could have been present. 

5.2.1

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. A Cockcroft Walton neutron 

generator in use before 1956 produced some tritium in the off-gas, but the amount would have 

Tritium
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 11 of 55

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

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 NDE evaluations with linear accelerators, X-ray 

machines, etc.; conducts telemetry testing; performs 

confirmatory measurements on components, assemblies, 

containers, etc.

1 

Quality Assurance Technician 

II

Performs NDE, 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, driver, 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 See Section 5.2.2.4

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, 

nonoperations management, 

Planner

Performs per job title Environmental only

Tool and dye maker Performs per job title Environmental only

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 Energy Employee’s tasks did not involve 

entry into cells, Gravel Gerties, igloos, or locations with resuspendable contamination. 

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 an environmental dose only unless there is information in the file to indicate 

otherwise. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 12 of 55

produced negligible intakes. 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). 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. 

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. 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 identical to ordinary water and is 

distributed throughout the entire mass of body water. 

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) computer 

program (James 2003), 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 four periods: 1956 to 1971, 1972 to 1982, 1983 to 1988, 

and 1989 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. Because there is no 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. Dose records in the 1990s specifically state 

monitored internal emitters (and give a dose) or state not monitored as “N/M”. No NOCTS files had 

tritium bioassay results for years before 1972, and the files included only a few post-1972 tritium 

results. Tritium doses in the files should be treated the same as those for 1972 to 1982 as described 

below. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 13 of 55

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 detection level and 0.5 μCi/L as the minimum detectable activity (MDA). 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 MDA at that time. Technical Basis for the Internal 

Dosimetry Program and the DOE Pantex Facility (BWXT Pantex 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 MDA as 15 dpm/ml

(0.0068 μCi/L).

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, the following paragraphs provide methods to convert 

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

Tritium intake prior to 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 and 

should apply this same conversion to recorded tritium doses for years before 1972 if they encounter 

any such doses. 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 1983-88

dose rate in rem/d = 8.12 × 10-6 q (5-1)

. 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. 

where q is the uptake in microcuries. Total dose was determined by integrating over the dose rate 

curve. For an acute exposure,

dose in rem = 1.4 × 10-4 q (5-2)

and for chronic exposure,

dose in rem = 8.12 × 10-6 qt (5-3)

where t is the period of chronic exposure in days. 

Equation 5-2 produces a conversion factor of 7.1 μCi/mrem and, assuming a 365-d exposure, 

Equation 5-3 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 5-2 is favorable to the claimant along with the 

assumption of a lognormal distribution with a GSD of 3. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 14 of 55

Table 5-3. Tritium dose data. 

Year

Workers monitored 

for tritium

Maximum recorded 

individual tritium dose 

(mrem)

Maximum 

uptake in 

μCia

Average worker

tritium dose 

(mrem)b

Average uptake 

in μCia

1972 4 12 42 8 28

1973 1 0 0 0 0

1974 0 NAc NA NA NA

1975 0 NA NA NA NA

1976 463 0 0 0 0

1977 466 0 0 0 0

1978 519 0 0 0 0

1979 712 0 0 0 0

1980 14 114 400 43.8 160

1981 41 122 430 14.2 50

1982 5 37 130 20.2 71

1983 NA NA 113d NA 0.070d

1984 0 NA NA NA NA

1985 17 3 21 0.6 4.3

1986 626 6 43 0.1 0.71

1987 481 2 14 0.02 0.14

1988 499 3 21 0.01 0.071

1989 212 1,180 430e 8.5 30

1990 2,341 3 14 0.002 0.010

1991 1,115 5 24 0.02 0.10

1992 879 5 24 0.05 0.24

1993 1,078 14 68 0.2 1.0

1994 1,104 11 53 0.1 0.48

1995 971 12 58 0.1 0.48

1996 940 7 34 0.017 0.082

1997 933 1 5 0.003 0.015

1998 610 1 5 0.0066 0.0066

1999 554 0 0 0 0

2000 535 0 0 0 0

2001 512 0 0 0 0

2002 511 0 0 0 0

2003 441 0 0 0 0

2004 421 0 0 0 0

a. Based on 3.5 μCi/mrem for 1972 to 1982, 7.1 μCi/mrem for 1983 to 88, and based on Equation 5-8b for 1990 to present. 

b. Based on Pantex recorded values.

c. NA = not available. 

d. Obtained directly from urinalysis results.

e. Based on 1981 maximum because of the large single incident in 1989.

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. For instance, a May 1991 letter to a 

worker’s file (case 1327) 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 d before the measurement.” Two other cases (3754 and 4276), 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 5-4 to convert from calculated 

dose to uptake assuming a single exponential retention curve with a 10-d retention half-time: 

dose = 1.3 × 10-3 C0 (5-4)
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 15 of 55

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) (5-5a)

uptake in pCi = (1.46 × 107

)(dose in mrem) (5-5b)

For the chronic intake scenario, the same document provides Equation 5-6 for calculating dose from a 

urine sample: 

dose in mrem = [(8.7 × 10-5 t) + 1.3 × 10-3] Ce (5-6)

where Ce is the urine concentration in disintegrations per minute per milliliter. Monthly sampling was 

the normal frequency for workers potentially exposed to tritium so, with a t of 30 d: 

dose in mrem = 3.9 × 10-3 Ce (5-7)

Distributing the tritium in 42,000 ml of body water gives: 

uptake in dpm = (1.07 × 107

)(dose in mrem) (5-8a)

uptake in pCi = (4.82 × 106

)(dose in mrem) (5-8b)

The same equation for an intake period other than 30 d is: 

uptake in pCi = (1.89 × 104

)(dose in mrem)/[(8.7 × 10-5 t) + 1.3 × 10-3] (5-9)

Equation 5-5 differs from Equation 5-8 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. 

Equation 5-8 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:

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 5-8 for monthly sampling frequency or Equation 5-9 for another 

exposure period; or
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 16 of 55

• Dose calculated from a termination sample or single sample when the worker was not on a 

monthly routine, use Equation 5-5.

Regardless of the step used to determine the dose, dose reconstructors should assume a lognormal 

distribution with a GSD of 3.

5.2.1.1.2 Tritium Missed Dose Before 1972

See Sections 5.2.1.1.5 and 5.2.1.1.6 for unmonitored worker discussions. 

5.2.1.1.3 Tritium Missed Dose, 1972 to 1988

The apparent urinalysis MDA in 1972 of 0.5 μCi/L would have resulted from a chronic intake for 30 d 

at 1.6 × 106 pCi/d, and the resultant dose to all organs would have been about 3 mrem. If the annual 

tritium dose is recorded as zero but it appears that some bioassay monitoring occurred, the dose 

reconstructor should assign the dose as a triangular distribution with a minimum of 0 mrem, a mode of 

18 mrem, and a maximum of 36 mrem. These values are small enough that they apply to the entire 

period; however, if more refinement is desired, the values below for 1983 to 1988 could be more 

appropriate. 

The urinalysis MDA of 0.05 μCi/L applicable in 1983 would have resulted from a chronic intake for 

30 d at 1.6 × 105 pCi/d, and the resultant dose would have been far less than 1 mrem. It is unlikely 

that Pantex tracked doses below 1 mrem, so an assignment of a 1-mrem dose for each month of 

potential missed dose is reasonable for the period from 1983 to 1989. If the annual tritium dose is 

recorded as 0 mrem but it appears that bioassay occurred, the calculation should use a triangular 

distribution with a minimum of 0 mrem, a mode of 6 mrem, and a maximum of 12 mrem. 

5.2.1.1.4 Tritium Missed Dose, 1989 to Present

The 1991 internal dose assessment procedure (MHSMC 1991b) lists urinalysis results above which a 

dose assessment is necessary (Table 5-4). 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 change 

dependent on the number of monitoring periods, as listed in Table 5-5. However, the potentially 

missed dose is reasonably close to 1 mrem/30 d, so it is favorable to the claimant 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 5-8). 

This dose would apply equally to all organs. [Using the formula from ANSI Standard N13.14-1983

(HPS 1983), the acute intake in Table 5-4 results in a 1-mrem dose if the time after intake is 7 d, 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]. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 17 of 55

Table 5-4. 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.

Table 5-5. 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.

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. 

5.2.1.1.5 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. A very small amount of tritium migration through 

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

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. Table 5-3 shows this pattern and lists the 

maximum and average recorded tritium doses from 1972 to 2001. Because tritium uptakes before

1972 were unknown but unlikely to be greater than the post-1972 uptakes, this analysis assumed that 

twice the highest uptake from the 1970s was to apply to all the years from 1957 to 1971 (Table 5-6.) 

Dose reconstructors should consider the resultant dose a constant upper bound. 

5.2.1.1.6 Unmonitored Workers, 1972 to 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. 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. Table 5-6 lists those 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 18 of 55

uptakes. 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 5-6 and minimums of zero.

Table 5-6. Tritium uptakes for unmonitored workers. 

Year

Maximum recorded 

individual tritium 

dose (mrem)

Maximum uptake 

(μCi)

a Average worker 

tritium dose (mrem)b

Average uptake (mode 

of dis tribution)

(μCi)

a

1956-71 24c 84c 16c 56c

1972 12 42 8 28

1973 0 5.8d 0 2.9e

1974 NMf 5.8d NM 2.9e

1975 NM 5.8d NM 2.9e

1976 0 5.8d 0 2.9e

1977 0 5.8d 0 2.9e

1978 0 5.8d 0 2.9e

1979 0 5.8d 0 2.9e

1980 114 400 43.8 160

1981 122 430 14.2 50

1982 37 130 20.2 71

1983 NA 113 NA 2.9e

1984 NM 5.8d NM 2.9e

1985 3 21 0.6 4.3

1986 6 43 0.1 2.9e

1987 2 14 0.02 2.9e

1988 3 21 0.01 2.9e

1989 1,180 430g 8.5 30

1990 3 14 0.002 2.9e

1991 5 24 0.02 2.9e

1992 5 24 0.05 2.9e

1993 14 68 0.2 2.9e

1994 11 53 0.1 2.9e

1995 12 58 0.1 2.9e

1996 7 34 0.017 2.9e

1997 1 5.8d 0.003 2.9e

1998 1 5.8d 0.0066 2.9e

1999 0 5.8d 0 2.9e

2000 0 5.8d 0 2.9e

2001 0 5.8d 0 2.9e

2002 0 5.8d 0 2.9e

2003 0 5.8d 0 2.9e

2004 0 5.8d 0 2.9e

a. Based on 3.5 μCi/mrem for 1972 to 1982, 7.1 μCi/mrem for 1983 to 1988, and Equation 5-8b for 1989 to present.

b. Based on Pantex recorded doses.

c. Assumed values based on twice highest values in 1970s.

d. Based on (0.135 μCi/L)(42 L). 

e. Based on (0.5)(0.135 μCi/L)(42 L). 

f. NM = not monitored. 

g. Based on 1981 maximum because of the large single incident in 1989.

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

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

sources. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 19 of 55

5.2.1.2 Tritium Release Event in 1989

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

exposed to tritium. This event is reasonably well documented, and data are available for assessment. 

Data on this event are from two primary sources: Radiation Safety Department Incident Records, 

Date: 5/17/89, Description: Cell 1 Incident (Pantex 1989) 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 hr after the incident. According to 

unclassified information from the classified report, the first bioassay occurred at 1630 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 will be in the worker’s records and are 

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

The latest version of Section 4 of this Site Profile addresses the environmental release from this 

incident (ORAUT 2004b).

5.2.2

5.2.2.1 Background

Uranium

Uranium at Pantex was enriched, natural, or depleted. Natural uranium was in a form referred to as 

tuballoy. Enriched uranium (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 (BWXT Pantex 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 U and DU throughout the Portsmouth, 

Paducah, K-25, Y-12 complex. 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 any intakes of DU: 307 pCi 239Pu/g DU; 3.53 pCi 237Np/g DU; 

509 pCi 99Tc/g DU. 

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 (BWXT Pantex 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 maximize the dose to the specific organ of concern. Exposure to significant quantities 

of type F uranium at Pantex is not credible. 

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. Uranium oxide became most noticeable beginning in the early 1980s and 

was present on eight of the 31 weapons types disassembled at Pantex to date, with types 28 and 55 

apparently having the highest contamination from black dust according to worker interviews. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 20 of 55

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, 

machining of DU-contaminated metal occurred associated with one weapon design. 

Because there is no evidence that workers were routinely monitored for uranium before 1991, unless 

claimant records clearly indicate that uranium bioassay occurred, dose reconstructors should interpret 

routine occupational records before 1991 that show “0” for internal dose as no information available 

rather than as a dose below detectable levels. 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. The technical basis document at the time (BWXT Pantex

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. BWXT Pantex (1992) stated that the 0.15-

dpm screening level would not apply if isotopic ratios implied that the uranium was not derived from 

DU. 

Since 1993, monitoring of uranium exposures has been event-driven and is identified by airmonitoring 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, the implication is that there must have been workplace indications of potential 

uranium intakes. Analysis of Biological Samples for Uranium, Thorium, and 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. 

Table 5-7 lists recorded doses (CEDE) from uranium exposures from 1991 to 2004 as recorded in 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 

(BWXT Pantex 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 code 

(James 2003), and dose reconstructors should use the IMBA values for consistency.
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 21 of 55

Table 5-7. Uranium dose to workers. 

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

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. Table 5-8 summarizes some information 

Table 5-8. History of uranium urinalysis. 

Year Laboratory Sensitivity value Description

1959 Los Alamos Scientific Laboratory 0.5 μga

1960 Tracer Laboratory 10 μg/La Fluorometry sensitivity

1963 Controls for Radiation 0.10 μg/La Less-than value

1965 Controls for Radiation 0.10 μg/La Less-than value

1967 Controls for Radiation 0.15 μg/La Less-than value

1968 Isotopes, Inc. 0.10 μg/La

1978 Control for Environ. Pollution 5 μg/samplea Less-than value

1983 Camp Dresser & McKee 1.4 pCi/La 2σ counting error only; use 3.3 for MDA

1983 Los Alamos Scientific Laboratory 5 μg/L Less-than value

1990-92 Y-12 Bioassay Lab 0.03 pCi/sample MDAb

1994 Y-12 Bioassay Lab Approx. 0.15 dpm/

sample, U-238, U234, U-235. 0.06 

dpm/sample U-236

MDAsa

2001 Y-12 Bioassay Lab U-238, U-234, U235: 0.03 pCi/LMDAc

a. From reports from the laboratories.

b. From BWXT Pantex (1992) and Ealy (1990). 

c. From BWXT Pantex (2001). 

found in the records. In most cases, the sensitivity parameter in the table was based on observed 

less-than values in the records. The equations used to determine the less-than values are not known; 

however, it is likely that the less-than value was more of a decision level than an MDA. Assume the 

MDA is twice the less-than value.

Table 5-8 has temporal gaps. Because uranium bioassays were generally not obtained routinely but 

usually from special bioassay samples obtained after events with potential for intake, it is not known if 

the gaps in the table occur because no bioassay was obtained in those years or if the bioassay 

sensitivities have not been found. If necessary, dose reconstructors should use the last previous 

MDA for years not covered in Table 5-8.
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 22 of 55

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 the claimant to assume the intake was natural 

uranium (BWXT Pantex 2001). However, dose reconstructors should assume DU for intakes 

associated with hydroshots and burning pads. 

5.2.2.3 Uranium Intakes from Assembly/Disassembly of Weapons for Unmonitored Workers

Dose reconstructors should use bioassay data for uranium in worker files to calculate intakes. 

However, uranium bioassays in the files are scarce. If they were not involved in an incident with 

potential for intakes, some production line workers might not have had uranium bioassay; therefore, a 

missed dose calculation for possible frequent intermittent intakes from disassemblies might not be 

possible. The following discussion provides an approach for assigning intakes for workers associated

with assembly and disassembly when no bioassay data are available or when bioassay data are 

insufficient to determine possible missed dose. 

A good set of DU intake data found in the Pantex records is related to a contamination incident in 

February 1989. Approximately 6 months after identification of the incident, workers were given in vivo

counts by Helgeson Scientific Services. The data from these counts were subsequently determined 

to be “flawed with an apparent positive bias” (Brake 1989). Because of this and a recommendation 

from a DOE investigation team, Pantex management committed to perform urinalysis on all workers. 

Note: Dose reconstructors should not

The bioassay samples were taken in late 1989 and early 1990, approximately 1 yr after the 

contamination incident was identified. Internal Dosimetry Urinalysis Studies (BWXT Pantex 2000)

contains bioassay data from urinalyses. This is the oldest set of data that provides isotopic 

determination of uranium alpha activity in urine samples and has sufficient data to perform statistical 

analysis. 

use data from the Helgeson in vivo counts. 

Martin Marietta Energy Systems, Inc., processed the samples at the Y-12 Plant Laboratory. The 

MDA, 0.03 pCi per isotope per sample, was calculated from the formula in Performance Criteria for 

Radiobioassay (HPS 1996). The calculation included a 1,000-min count, a detector efficiency of 

0.0985, and an average recovery of 75%. The average chemical recovery for the sample sets for this 

incident was 70%. Recoveries less than 25% were considered not accurate due to counting statistics 

associated with low recoveries. No dose assessment was provided with the data set. Data with a 

negative value indicated the background was higher than the activity in the sample.

5.2.2.3.1 Analysis of 1990s Uranium Urinalysis Data

From February 1 to April 1, 1990, 305 workers provided urine samples for analysis. The results are 

documented in the HERS microfilm archive, Roll 9030, identified by badge numbers beginning with 

“P.” In addition, these data contain 23 spiked samples (badge number C999999) and 11 “R999999” 

samples. This data set was analyzed to establish a baseline for unmonitored workers (or workers 

whose bioassay results are not found). 

The 305 “P” sample measurements, associated with real individuals, averaged 1,635 ±384 g urine 

and 0.229 ±0.168 dpm total uranium per sample. These measurements are reportedly 24-hr urine 

samples. 

When four individuals for whom radiochemical recovery was listed as zero were removed from the 

data set, the measurements averaged 1,623 ±385 g urine and 0.227 ±0.163 dpm/d total uranium. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 23 of 55

Fitting a lognormal distribution to the non-negative values (299 of 301 observations) with two values 

treated as missing (Strom 1986) gave a median (geometric mean) of 0.188 dpm/d with a GSD of 2. 

Assuming an acute inhalation intake of absorption class S DU 14 months before the excretion results 

in a median intake of 75,000 dpm. Although the urine samples were taken in response to a specific 

event, the measured uranium excretion could have been the result of years of accumulation of DU in 

the body from smaller, frequent, intermittent intakes. An assumption of 10 yr of chronic intake 

produces 50-yr committed doses about twice the acute intake doses; however, the chronic intake 

scenario produces smaller doses until about 7 to 8 yr after the start of intakes. The measured 

excretion probably resulted from a combination of contributions from chronic intakes since 1980 and 

the acute intake in 1989. The differences are not large and, during this period when disassembly 

activity was intense, the chronic intake scenario is the more plausible for most workers. 

Using the IMBA internal dosimetry software, the median inhalation intake for absorption type M 

uranium was determined to be 2.8 dpm/d based on the midpoint (5 yr) of the chronic intake period 

(that is, 0.188 dpm/d excretion after 1,826 d of chronic intake of type M uranium). The range of intake 

from 1 yr of chronic intake to 10 yr of chronic intake (2.99 dpm/d to 2.77 dpm/d) was contained within 

the standard GSD of 3. The median inhalation intake for absorption type S uranium was determined 

to be 41.5 dpm/d based on the midpoint (5 yr) of the chronic intake period, and the range of intake 

from 1 yr of chronic intake to 10 yr of chronic intake (81 dpm/d to 34.1 dpm/d) was contained within 

the standard GSD of 3. 

Similar statistical analyses were performed on the 238U results for 1991 through 1994, using coworker 

analysis methods presented in Analysis of Coworker Bioassay Data for Internal Dose Assignment

(ORAUT 2005). The median results and GSDs were 0.041 (2.7), 0.061 (2.7), 0.061 (3.0), and 0.033 

(3.0) dpm/d for 1991, 1992, 1993, and 1994, respectively. Assuming 238U represents 83% of the total 

alpha activity in DU, the total uranium median activities would be 0.049, 0.074, 0.074, and 0.040 

dpm/d for 1991 through 1994, respectively. These values are less than the 0.188 dpm/d determined 

from the 1990 data, and they provide some confidence that the 1990 data is favorable to the claimant. 

Because ingestion of uranium cannot be ruled out, this TBD analysis estimated chronic ingestion 

intakes. The median ingestion intake of soluble uranium (f1 = 0.02) was calculated to be 9.75 dpm/d, 

and the median ingestion intake of insoluble uranium (f1 = 0.002) was calculated to be 97.5 dpm/d. 

Because equilibrium is quickly established for ingestion, the ranges from 1 yr to 10 yr of chronic intake 

were not significantly different from the median intakes. 

5.2.2.3.2 Depleted Uranium Intakes for Assembly/Disassembly Unmonitored Workers, 

1980 to 1993

When bioassay data are not available for Production Technicians (Assembly Operators), RSTs, and 

Quality Assurance Technicians for 1980 to 1993, dose reconstructors should assume a median 

chronic DU inhalation intake of one of the following: Type M inhalation - 1.3 pCi/d, type S inhalation – 

19 pCi/d, soluble ingestion – 4.4 pCi/d, or insoluble ingestion – 44 pCi/d. The distribution should be 

lognormal with a GSD of 3. 

When bioassay data are not available, to assess the intake of a worker whose job had a lower 

potential for intake but might have had incidental exposure to contamination from disassembly 

activities (such as category 2 in Table 5-2 or the “some potential” category in ORAUT [2004c]), dose 

reconstructors should assign 10% of the intake. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 24 of 55

5.2.2.3.3 Depleted Uranium Intakes for Assembly/Disassembly Unmonitored Workers, 

1994 to the Present

As indicated in Table 5-7, doses to workers were reduced considerably from 1994 to the present due 

to better procedural controls and barriers. All of the 1995 238U bioassay results, for instance, were 

below detection. To account for potential unmonitored dose, dose reconstructors should assign a 

claimant who worked in a job with a high potential for intake with an intake that is 0.2 times the intake 

from 1980 to 1993 to account for improved radiation protection barriers and procedures (lognormal, 

median, GSD = 3) (based on the ratio of urine excretion in the 1994 dataset to the 1989 dataset). 

5.2.2.3.4 Depleted Uranium Intakes for Assembly/Disassembly Unmonitored Workers, 

1961 to 1979

The first disassembly of a weapon with DU oxide at Pantex occurred in 1961 (BWXT Pantex 2004, 

p. 2-10). The number of disassemblies at Pantex is classified. Prior to 1980, approximately 10

disassemblies were performed for maintenance or quality control per year per weapons program at 

Pantex. Table 5-9 lists the number of weapons retired from use throughout the DOE complex. 

Weapons can be retired without being disassembled but the trends over time should be similar, 

although what is not shown in or inferable from the table are the numbers of partial disassemblies that 

resulted in refurbishing of weapons. From 1961 to 1965, most disassemblies were performed at the 

Medina and Clarksville plants (BWXT Pantex 2004, p 2-19), and from 1966 to 1975, disassemblies 

were performed at Pantex and the Iowa Army Ammunition Plant. One can infer from the table that the 

annual number of disassemblies at Pantex for the years before 1980 was probably somewhat less 

than the number for the period from 1980 to 1989. In addition, before about 1980 nuclear weapon 

disassembly operations were generally free of contamination. However, because the documentation 

of the number of disassemblies and partial disassemblies and the contamination levels are not 

specifically available, it is favorable to the claimant to assign unmonitored workers an intake that is the 

same as that from 1980 to 1993 in accordance with the worker’s risk potential from Table 5-2 

(lognormal, median, GSD = 3). 

A few small sets of uranium urinalysis results were found with analysis dates ranging from February 

1963 to March 1967. Although the dates and circumstances for the samples (e.g., routine or postaccident) were not found, the 34 results were analyzed as a group assuming a lognormal distribution 

and coworker analysis methods presented in Analysis of Coworker Bioassay Data for Internal Dose 

Assignment (ORAUT 2005). The median excretion rate was 0.375 dpm/d with a GSD of 4.1. 

Considering that the circumstances of the sampling are not known, this excretion rate compares 

reasonably well with the median excretion of 0.188 dpm/d assuming chronic intake for the large group 

of samples collected in 1990.

5.2.2.4 Depleted Uranium Intakes from Machining

Unlike other sites that performed hydro tests, Pantex did not machine DU as part of the hydro test 

program. However, machining was performed on DU for a period in the early 1960s associated with 

one particular weapon design. A 1960 report describing these operations stated that the workers 

were required to wear respirators during all machining operations and that breathing-zone air

sampling was performed during the machining with all samples showing no detectable activity. 

Judging from the air sample monitoring log at the firing sites at the same time, the air-sampling 

detection limit at Pantex at that time was about 2 pCi/m3 (2 × 10-12 μCi/ml). This is similar to the 

detection limit stated for air sampling at the Hanford uranium fabrication facilities in the same time 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 25 of 55

Table 5-9. Retirements in the DOE complex.

a

Year Retirements

Fraction relative to 1980-89 yearly 

average assuming 100% at Pantex

Fraction relative to 1980-89 yearly 

average assuming 50% at other plants

1961 1,571 1.25 0.326

1962 766 0.611 0.305

1963 830 0.662 0.331

1964 2,534 2.02 1.01

1965 1,936 1.54 0.772

1966 2,357 1.88 0.940

1967 1,649 1.32 0.658

1968 2,194 1.75 0.875

1969 3,045 2.43 1.21

1970 1,936 1.54 0.772

1971 1,347 1.07 0.537

1972 1,541 1.23 0.614

1973 544 0.434 0.217

1974 807 0.644 0.322

1975 2,240 1.79 0.893

1976 2,181 1.74 NAb

1977 998 0.796 NA

1978 1,148 0.915 NA

1979 730 0.582 NA

1980 904 NA NA

1981 1,887 NA NA

1982 1,537 NA NA

1983 749 NA NA

1984 1,143 NA NA

1985 1,322 NA NA

1986 1,224 NA NA

1987 958 NA NA

1988 1,023 NA NA

1989 1,794 NA NA

a. From DOE (2001)

b. NA = not applicable.

frame (Wilson 1958). Assuming exposure for 8 hr/d just at the detection limit, the daily intake would 

have been:

working day (wd) intake = (8 hr/d)(1.2-m3

/hr breathing rate)(2 pCi/m3) = 19 pCi

calendar day (cd) intake = (19 pCi/wd)(250 wd/yr/365 cd/yr) = 13 pCi.

Because the air sample results were below this level and the machinists wore respirators, the 13-

pCi/d inhalation intake rate should be considered a constant upper bound. Ingestion of DU from this 

work cannot be ruled out. Following guidance in Estimation of Ingestion Intakes (NIOSH 2004), the 

estimated ingestion intake would have been: 

ingestion daily intake = (0.2)(2 pCi/m3

) = 0.4 pCi/cd (constant).

5.2.2.5 Depleted Uranium Intakes from Burning of Contaminated High Explosives

A 1971 letter from the Plant Manager to the manager of the DOE Albuquerque Operations Office on 

the burning of HE at the burning grounds stated that, “during the past years, uranium (238U) parts have 

been burned in conjunction with the waste HE ... ” (Drummond 1971). Dose from intakes by 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 26 of 55

operators of the burning pad was considered occupational internal dose; exposure of other workers 

from plumes was considered environmental exposure. During a telephone interview, Herman Phillips, 

who has worked as a safety engineer at Pantex since 1952, provided details on the burning 

operations (Bihl and Martin 2004). Burning HE was a part-time task for Transportation Department 

workers involving a few hours approximately once a week. Only a few workers would have been 

involved during any one burn. The ignition operators were about 100 yards from the burn pad during 

a burning. Access for all other workers was restricted to about 300 yards or more. Ash from the burn 

was collected and placed in 10- to 20-gal cans for burial. Operators wore half-face respirators with 

high-efficiency particulate air filters during the ash collection task. 

The burning grounds have operated since 1952 (DOE 1997). Air sample results from the burning 

grounds cover 1960 to 1967, with no results for 1963 (MHSMC c. 1967). Two categories of results 

are listed as during burning and during cleanup. Some results are recorded as disintegrations per 

minute per cubic meter and others as counts per minute. 

In relation to the air samples taken during burning, 24 were listed as zero or background, 9 had 

results that ranged from 4 to 112 dpm/m3

, and 17 had nonzero results in counts per minute. (In the 

logs, the results are written as d/m/m3

.) No information was available for converting the counts-perminute results to concentrations, but if one assumes a 25% counting efficiency, the nonzero results 

would represent a range from 2 to 77 dpm/m3

. Therefore, the air sample result of 112 dpm/m3 is 

probably the highest concentration measured. Considering that there were 50 total samples, it can 

reasonably be assumed that the highest result represents the 95th percentile or higher. The exact 

locations of the samplers were not shown, but reference was frequently made to the pad, implying 

that the samplers were close to the burning material and not the worker locations. 

To estimate the concentration at the location of the workers, this analysis used the following equations 

from National Council on Radiation Protection and Measurements (NCRP) Report 123 I (NCRP 

1996), assuming the air sampler was about 2 m and the workers about 100 m from the burning 

material: 

C = fQ/πuσyσz (5-10)

where C is the concentration from a ground-level release in activity per cubic meter, f is the fraction of 

time the wind blows in the direction of interest, Q is the release rate in activity per second, u is the 

wind speed in meters per second, and the sigma values are the horizontal and vertical atmospheric 

diffusion parameters. 

In this application, the change in concentration from 2 to 100 m for the same burn is being calculated, 

so fQ/πu is constant. The equations then become: 

C(air sample) = K/σyσz at 2 m (5-11)

C(workers) = K/σyσz at 100 m (5-12)

The diffusion parameters are determined by:

σy = (0.08x)(1+0.0001x)

-1⁄2 (5-13)

σz = (0.06x)(1+0.0015x)-1⁄2 (5-14)

where x is the distance from the burning point to the air sampler or workers. When expressed as a 

ratio of concentrations, the K term cancels, so: 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 27 of 55

C(workers) = C(air sample) σyaσza/σywσzw (5-15)

using the subscript w to refer to the worker’s location and a to refer to the air sample location. 

Using Equations 5-13, -14, and -15 at distances of 2 and 100 m and the 95th-percentile air sample 

concentration of 112 dpm/m3

, the estimated concentration of DU for workers during burning was 

4.8 × 10-2 dpm/m3

. The intake was determined as the product of air concentration, breathing rate, and 

exposure time. Sampling times of some air sample results were logged; the times were usually from 

1 to 2 hr. [ERDA (1976) states that, “Explosives burn rapidly and amounts of 500 lbs or more are 

usually consumed in less than 15 minutes. However, occasionally (about once a month) smoke can 

be observed for up to 30 minutes.” Nevertheless, worker exposure was assumed to be 2 hr to be 

consistent with the air sample run times.] Therefore: 

95th-percentile intake in pCi 

 = (4.8 × 10-2 dpm/m3

)(1.2 m3/hr)(2 hr)/2.22 dpm/pCi = 0.052 pCi DU per burn (5-16)

According to Mr. Phillips, burns might have been performed about once a week (Bihl and Martin 

2004). The air sample log shows far fewer entries than weekly in the years sampled; this analysis 

could not determine if air samples were taken for every burn. It is reasonable that different workers 

were involved with different burns. Therefore, the once-a-week estimate seems to be a reasonable 

upper bound for each worker. Under the assumption that the same workers incurred intakes when 

the activity referred to in the logs as “pickup” or “cleanup” occurred, the daily intake was: 

(weekly intake)(50 weeks/yr)/365 d/yr = 7.1 × 10-3 pCi/d (5-17)

Twenty-one air samples were taken during cleanup activities. Only 1 result was listed as background; 

6 were nonzero results recorded in disintegrations per minute per cubic meter; and 14 were nonzero 

results listed in counts per minute. When plotted as a lognormal distribution, the results in 

disintegrations per minute per meter indicated a median value of about 22 dpm/m3 (1 × 10-11 μCi/ml) 

and a 95th percentile of 900 dpm/m3

. The results in counts per minute fell within this distribution with 

an exception of one at 1,040 dpm/m3 under a counting efficiency assumption of 25%. No information, 

such as sampler location, was available on the representativeness of the air sample concentration in 

relation to the air breathed by the workers, but the large GSD is adequate to cover uncertainties in the 

representativeness of the samples. Mr. Phillips stated that the workers wore half-face respirators 

during cleanup activities (Bihl and Martin 2004), but credit was not taken for this use of respirators. 

On the few occasions that the time of sampling was logged, the times ranged from 40 to 90 min. A 2-

hr exposure time for cleanup of the ash was assumed. There was no definitive information about the 

frequency of cleanup activities; the air sample logs show cleanup at a less frequent rate than burning. 

This analysis made the favorable to claimant assumption that cleanup occurred once a week. 

Therefore, the 95th-percentile default intake from cleanup of DU-contaminated ash is:

95th percentile intake in pCi/d 

= (900 dpm/m3

)(1.2 m3/hr)(2 hr)(50/365)/2.22 dpm/pCi = 130 pCi/d (5-18)
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 28 of 55

Assuming the same workers performed the cleanup as those who performed the burn, the intake from 

the burn itself is negligible in relation to the cleanup activities. Therefore, the default intake rate of DU 

for burning ground workers is 130 pCi/d for 1952 to the present. This intake would represent the 

upper bound so should be considered a constant distribution. Dose reconstructors should assume 

either absorption type M or S. 

5.2.2.6 Intakes from Hydroshots 

5.2.2.6.1 DU

Pantex has used firing sites for HE quality control and research since 1952. Some of the test firings 

at Firing Sites 4, 5, and 10 involved DU through 1985 (DOE 1997; MHSMC 1990, Chapter 7). 

According to a former employee who worked at the firing sites from 1959 to 2000 and was supervisor 

of the firing sites since 1960, the first hydroshot that involved DU occurred in late 1959 or 1960 (Bihl 

2006). Dose from intakes by operators of the hydroshots was considered occupational internal dose; 

exposure of other workers from plumes was considered environmental exposure. Mr. Phillips 

described the hydroshot operations (Bihl and Martin 2004). Operators were in a bunker a few tens of 

feet from the detonation site. The bunker was fully enclosed with a ventilation system that was shut 

down during the detonation. After the detonation, the operators walked to ground zero to retrieve their 

instruments. A driver, who was outside the fenced area (approximately 2,000 ft from ground zero), 

drove to the detonation site to retrieve the operators. The total exposure time for the operators was 

less than 30 min. The cloud from the detonations was clearly visible and the operators and driver 

avoided direct exposure to the cloud. 

The prevailing winds carried the cloud basically northward away from most Pantex buildings and 

worker locations. The water treatment plant and sewage treatment plants are about 1 and 2 km, 

respectively, northeast of Firing Site 5; they are a bit closer but nearly due east from Firing Site 4. 

This analysis found data providing air concentrations inside and outside the bunker at Firing Site 4 for 

October 1959 to January 1962 (MHSMC c. 1967). The data list 94 results for inside the bunker and 

79 results for outside. Eighty-five percent of the results are recorded as 0 dpm/m3 with the lowest 

nonzero value recorded as 1 dpm/m3 (only one significant figure was recorded). Figure 5-1 shows a

log-probability plot of the inside-bunker data, and Figure 5-2 shows that for the outside-bunker data 

with the air concentrations expressed in picocuries per cubic meter. The inside concentrations fit a 

log-probability curve with a median concentration of 0.18 pCi/m3 and a 95th-percentile value of 

6.2 pCi/m3

. The outside concentrations had a median concentration of 0.82 pCi/m3 and a 

95th-percentile value of 24 pCi/m3

. The following conditions could have caused the large spread in 

the air concentrations:

• Different hydroshots could have contained different masses of DU or HE.

• The cloud could have been sharply defined close to the detonation spot and shortly after the 

detonation time, such that under rare weather conditions the samplers were within the 

boundaries of the initial cloud, whereas most of the time they were not.

• A sample filter could have occasionally picked up a large particle of DU.
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 29 of 55

Pantex obtained air concentrations from hydroshots in the 1970s using a radio-controlled drone 

mounted with air samplers on each wing. The drone flew through the cloud within a minute or two of 

the test at a location expected to sample the highest concentration. These results were used to 

predict offsite concentrations rather than exposure of the test operators. Mr. Phillips expressed an 

intuitive estimate that ground-level concentrations for the operators would have been at least 100 

times lower (Bihl and Martin 2004). 

This analysis found results from the drone samplers for March 1971, various dates in 1972, and two 

dates in the summer of 1974 (Gidley 1971; Alexander 1972a,b,c, 1974). It is not known if these data 

Lognormal Percentages

Inside Bunker (pCi/m3)

1 5 10 50

90 95 98 99 99.5

Figure 5-1. Air sample log-probability fit inside hydroshot bunkers. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 30 of 55

Lognormal Percentages

Outside Bunker (pCi/m3)

0.5 1.0 5.0 10.0 50.0

80 90 95 98 99

Figure 5-2. Air sample log-probability fit outside hydroshot bunkers. 

comprise the complete set of data from drone flights or merely what has been found to date. The 

design of the firing pad at Firing Site 5 changed from a flat pad to a sand-filled pit in the summer of 

1971, but the drone results do not show an appreciable change in cloud concentration before and 

after the change. The log-probability fit to the drone sampler results had a median concentration of 

70 pCi/m3 and a 95th-percentile concentration of 944 pCi/m3

, so the median ground concentration 

outside the bunker in the 1960s was about 85 times less than the median cloud concentration 

measured in the 1970s. Considering the large GSDs and the dynamics of cloud concentrations 

measured very near to the cloud origin, the two sets of measurements are reasonably consistent; that 

is, they do not indicate that the 1970s shots produced significantly higher or lower ground-level 

concentrations than the 1960s shots. 

The March 1971 drone results were from a study that compared the cloud concentration of DU from a 

hydroshot that contained DU and two detonations from Firing Site 5 that contained only HE (Gidley 

1971). The cloud concentration of DU was similar for all three shots, and the DU from the HE-only 

shots was resuspended contamination from the firing site area. The effect of the sandy pit on 

resuspension of DU is not known, but by 1977 there was a policy to not use Firing Site 5 for any nonDU detonations because, “this could cause the resuspension of some of the residue of shots fired at 

that bunker in the past” (Boettner 1977; Wilcox 1984; Alexander 1984). Therefore, intakes of DU by 

firing site personnel were possible after every shot at Firing Site 5, not just the hydroshots. 

Table 5-10 lists what was discovered about the number of DU-related detonations. The number of 

shots in 1960 and 1961 was assumed to equal the number of air samples in the air sample log (often 

more than one sample per day), namely 30 and 58, respectively. These shots occurred at Firing Site 

4. The shots from 1971 to 1986 occurred at Firing Site 5. Test firing using DU ended in 1986 

(MHSMC 1990, Chapter 7). As shown, there are gaps in the data. This analysis used annual 

environmental reports to fill some of the gaps (Alexander 1975, 1976; Alexander, Cornelius, and 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 31 of 55

Horton 1978; Laseter 1986). Because the first DU shot may have occurred in late 1959, the number 

of shots for 1959 was chosen to be half of the number of shots in 1960. 

Table 5-10. Recorded numbers of hydro test shots and 

other detonations at Firing Site 5 or other sites. 

Year Hydroshots Other shots at FS-5 Total

1967 67a NDb ND

1968 3a ND ND

1969 37a ND ND

1970 54a ND ND

1971 49c ND ND

1972 45c ND ND

1973 28c 52c 80

1974 ND ND ND

1975 ND ND ND

1976 7c 17c 24

1977 ND ND ND

1978 ND ND ND

1979 3c 8c 11

1980 5c 0 5

1981–86 ND ND ND

a. From ERDA (1976). Specific site was not identified. 

b. Not discovered.

c. From Johnson (1973, 1974, 1980, 1981, 1982) and Alexander (1977).

Some of the annual environmental reports listed release activities of DU based on estimates of the 

amount of DU that went airborne from the firing site. Table 5-11 summarizes that data. The ratio of 

1973 release activities over those from 1976 was 5, whereas the ratio of the number of hydroshots 

was 4 and the ratio of total shots at Firing Site 5 was 3.3. Therefore, it is favorable to the claimant to 

use total shots in 1973 as the base and the ratio of release activities to estimate total shots for years 

where the numbers of shots are missing. For instance, the estimated number of shots at Firing Site 5 

for 1977 becomes (0.001)(24)/0.00004 = 600. However, a limit was set at 500 shots per year based 

on a reasonable limit of two shots per working day to which any one worker could have been 

exposed. Table 5-11 lists the recorded or estimated total number of shots potentially involving 

dispersion of DU. 

The default intake for firing site operators and drivers, assuming a 30-min exposure to each shot, 

would be: 

daily chronic intake in pCi 

 = (air concentration pCi DU/m3

)(1.2-m3/hr breathing rate)(0.5 hr)(no. of shots/yr)/(365 d/yr) (5-19)

Because of the uncertainty as to whether an operator was or was not in the cloud for any particular 

shot and because the operators and driver spent part of their time outside of the bunker retrieving 

instruments, the 95th percentile outside air concentration was used to determine the intake per shot

(24 pCi/m3

). Substituting this air concentration and simplifying the math, the formula for the default 

intake becomes: 

daily chronic intake in pCi = (3.9 × 10-2)(no. of shots/yr) (5-20)
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 32 of 55

Table 5-11. Estimated number of detonations 

potentially involving DU dispersion and 

intakes. 

Period Total shots Intake (pCi/d)

1959 15a 0.6

1960 30 1.2

1961 58 2.3

1962-72 500/yrb 20

1973 80 3.2

1974 60 2.4

1975 120 4.7

1976 24 0.95

1977 500b 20

1978 500b 20

1979 11 0.43

1980 5 0.20

1981 6 0.24

1982 6c 0.24

1983 6c 0.24

1984 0 0

1985 6c 0.24

1986 6 0.24

a. Assumed to be half of the number for 1960.

b. Assumed 2 shots per working day.

c. Assumed equal to highest year in 1980s.

Because the 95th percentile was used, which certainly would not apply to all shots, dose 

reconstructors should consider the result to be a constant upper bound and assume either absorption 

type M or S.

5.2.2.6.2 Other Possible Radioactive Sources

The Pantex Plant Radiological Investigation Report (BWXT Pantex 2004) states that contaminated 

soil from the firing sites potentially contained low levels of beryllium, DU, strontium, and thorium. The 

more detailed discussion of the firing sites does not mention use or residual contamination of 

radioactive strontium, and when questioned directly if there had been a hydroshot involving 

radioactive strontium, Mr. Phillips said there had not been. Mr. Phillips said there had been one 

hydroshot involving thorium (Traub 2007). Until then, dose reconstructors should assign the intake 

below, once, to anyone involved with hydroshots]. Assuming that the air concentration from the 

thorium hydroshot was no worse than the 95th-percentile concentration outside the bunker from the 

DU shots (24 pCi/m3

), and assuming a 30-min exposure, the one-time acute intake is: 

acute intake in pCi 

= (air concentration pCi Th/m3

)(1.2-m3/hr breathing rate)(0.5 hr) (5-21)

acute intake in pCi = 14 pCi.

As discussed in Section 5.2.3, the state of equilibrium of 232Th with its progeny is not known. It is 

favorable to the claimant to assume that the 14 pCi is for 232Th and the progeny are in equilibrium. 

Either absorption type M or S may be assumed. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 33 of 55

5.2.2.6.3 Clean Up of Firing Site 5

Firing Site 5 was decontaminated in the late 1990s. According to the Radiation Safety Section 

Manager, the work was done by a subcontractor under Radiation Work Permits with monitoring by site 

Radiological Control Technicians and bioassays were collected. This work was conducted during the 

time Pantex was using lapel air samplers and using 40 DAC-hr as the trigger for initiating bioassay. 

Therefore, if it is determined that an energy employee was an unmonitored worker associated with the 

Firing Site 5 cleanup, an assumption of chronic intake of 50 DAC-hr per year of DU can be made. 

[This assumes that intakes below 1 DAC-hr on any given air sampler are disregarded and there might 

be 50 such results per year.] Fifty DAC-hr of type S uranium equates to an intake of 600 pCi, or 

1.6 pCi/d.

5.2.2.6.4 Clean Up of Firing Site 23

In the 1980s a device was constructed at Firing Site 23 for destruction of small, prototype-sized 

weapons with HE and DU that had been tested off the site and returned to Pantex. The device, 

referred to as the Silver Bullet, was a contained system with air effluents vented through a highefficiency particulate air (HEPA)-filtered system. Within the Silver Bullet, a device was surrounded 

with lead plates and bags of gravel. The Silver Bullet was used for one weapons design only and only 

about 10 shots were conducted. After the explosions, the contaminated gravel was removed by 

shoveling onto a conveyor belt that dumped the gravel into a wooden box for disposal. Workers wore 

full anti-contamination clothing and respirators during both setup of the shot and cleanup afterwards. 

Cleanup of the gravel was conducted after each shot. 

Urinalysis data associated with one Firing Site 23 cleanup was found (Copeland 1983a,b). The exact 

period of the cleanup activities is not known; a batch of urine samples dated October 4, 1983 are 

labeled as background samples; the other sample dates range from November 9 to December 5, 

1983. The November and December DU urinalysis results were analyzed as a group using the 

coworker analysis methods in Analysis of Coworker Bioassay Data for Internal Dose Assignment

(ORAUT 2005). The results ranged from <5 to 50 μg/L, with all but one exceeding the analysis 

detection level. The median urinalysis result was 16 μg/L with a GSD of 2.1. 

The intake scenario was difficult to determine based on the pattern of the bioassay. Tracy Ikenberry, 

who was a radiological engineer at Pantex at the time, believed that actual removal of the 

contaminated gravel would have been a 1- or 2-d task at the most for a small crew of workers, 

although he did not witness any of the cleanups. Several of the workers had samples taken on 

November 9, 14, and 21, and the latest samples usually had the higher concentrations. Therefore, 

two intake scenarios were investigated: 1) The median result was modeled as a sample obtained four 

days after the start of two days of exposure, and 2) the median result was modeled as a sample 

obtained at the end of 12 d of chronic exposure (November 10-21). The first assumption produced 

the largest total intake (1.6 × 105 pCi/d DU for 2 d), which was about 3 times greater than the intake 

from the second assumption. Therefore, the uncertainty from the intake assumption was greater than 

the uncertainty in the distribution of the urinalysis results; and a GSD of 3 is reasonable. 

If urinalysis results are not available after subsequent shots and clean-up activities, the following 

default intakes can be assigned. Personal recollections of persons associated with use of the Silver 

Bullet indicate that there were 10 shots with the last shot in 1988. Because these were rare, shortterm events spread over many years, an assumption of acute intake is best. Assuming the first shot 

occurred in October or November 1983, the number of shots with cleanups can be distributed as one 

in 1983, two in each year from 1984 to 1987, and one in 1988. They should be modeled as acute 

intakes of 3.2 × 105 pCi of type S DU (GSD of 3) with intakes dates of November 9, 1983, January 2, 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 34 of 55

1984, July 1, 1984, January 2, 1985, July 1, 1985, January 2, 1986, July 1, 1986, January 2, 1987, 

July 1, 1987, and January 2, 1988. These intakes should be applied to workers identified as being 

involved with hydroshots or firing sites. The workers involved in the 1983 cleanup are identified in the 

records, and it is favorable to the claimant to assume they were involved in subsequent cleanups if 

consistent with their employment history. 

5.2.3

Thorium at Pantex 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. 

Thorium

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. 

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

occupational environmental dose, ORAUT 2004b). 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. 

232Th and 228Th levels 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) 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 

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. 

The dose reconstructor should assume equilibrium without evidence to the contrary. 

Because there is no evidence that workers were routinely monitored for thorium before 1991, unless 

the worker records clearly indicate bioassay for thorium, dose reconstructors should interpret routine 

occupational records before 1991 that show “0” for internal dose as no information available rather 

than necessarily as a dose below detectable levels. Dose records in the 1990s specifically state 

whether internal emitters were monitored (and a dose is given) or not monitored (N/M). Monitoring of 

thorium exposures has been event-driven since at least 1991. Table 5-12 lists the number of 

individuals monitored and the dose results. 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 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 35 of 55

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. This 

analysis found no bioassay data before 1983.

Table 5-12. Thorium dose (CEDE) to workers. 

Year

Number monitored 

for thorium

Total worker thorium 

dose (person-rem)

Maximum individual 

thorium dose (mrem)

Average worker 

thorium dose (mrem)

1991 0 0 0 0

1992 17 0 0 0

1993 0 0 0 0

1994 0 0 0 0

1995 67 0 0 0

1996 56 0 0 0

1997 13 0 0 0

1998 1 0 0 0

1999 16 25 25 1.6

2000 9 14 14 1.6

2001 16 221 150 13.8

2002 11 48 48 4.4

In early 1992, several workers were given baseline thorium urine bioassays. The analyses were

performed by Controls for Environmental Pollution, Inc. The results were less than 0.05 dpm/L. 

Therefore, dose reconstructors should assume that the MDA was about 0.1 dpm/L (0.045 pCi/L). All 

the follow-up results were less than the detection limit. 

It is not certain that NOCTS records will include bioassay results even for the monitored workers listed 

in Table 5-12. Considering the extent of workplace controls used during disassembly of thorium 

weapons, the very limited amount of available data in relation to thorium intakes, and the inability to 

determine the frequency of handling weapons containing thorium, dose reconstructors should: 

• For workers who had the highest possibility of intake (from Table 5-2), for each year from 1980 

to 2000, assume a single acute intake of 40 DAC-hr (48 pCi) of 232Th (in equilibrium with 

progeny) [(1 × 10-12 μCi/ml)(106 ml/m3

)(1.2 m3/hr)(40 hr)(106 pCi/μCi)]. Assign 0.1 times this 

intake for category 2 workers in Table 5-2. These intakes are modes of triangular distributions 

with a minimum of zero and a maximum of 10 times the mode to account for the possibility of 

more than one intake per year and possible unrepresentativeness of the air-sampling system. 

There is no record of disassembly of thorium weapons before 1980.

• For each year from 2001 to the present, for workers with the highest possibility of intake, 

assume a single acute intake of 4 DAC-hr (4.8 pCi) unless the recorded dose for the worker 

exceeds 5 mrem CEDE. If the recorded dose exceeds 5 mrem, convert the recorded dose to 

intake using 0.86 pCi/mrem [(3.2 × 103 Bq/Sv from ICRP 1989)(10-5 Sv/mrem)(27 pCi/Bq)]. 

Assigning 0.1 times this intake for category 2 workers in Table 5-2. Because Pantex has used 

lapel sampling during this period, these intakes are the maximum of a triangular distribution 

with a minimum of zero and a mode of one-half the maximum. 

A check on the reasonableness of the above estimates was made by analyzing 232

Th bioassay 

results. Two hundred fifty-eight worker urine samples were analyzed between 1992 and 1996. Only 

one result arguably exceeded the detection level; the median of the distribution was 0.000 pCi/L and 

the 95th percentile was 0.004 pCi/L (less than detectable). One hundred fifty-one worker fecal 

samples were analyzed between 1996 and 2000. About half were above the analytical detection 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 36 of 55

level, but only four exceeded the expected natural excretion of about 0.32 pCi/d (2.9 μg/d × 

1.1 × 10-

7 μCi/μg 232Th × 106 pCi/μCi [ICRP 1975]. An acute intake of 48 pCi would result in less than 

0.32 pCi/d excretion over about 6 d after the intake, so the intake estimate above and the fecal data 

agree reasonably well. 

5.2.4

5.2.4.1 Plutonium General 

Plutonium and Plutonium Incidents

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. If 

contamination occurred, exposure to plutonium would be acute rather than chronic. Table 5-13 lists 

the number of workers given plutonium bioassay by year. There were no recorded internal doses 

associated with these 1991 to 2002 bioassays.

Table 5-13. Number of workers on 

plutonium bioassay, 1991 to 2002. 

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

Although exposure to plutonium has been strictly controlled at Pantex, there is indication of past 

concern about potential plutonium intakes. Based on two sets of bioassay data, it appears that in 

1963 and 1966 monthly urine sample collection and analysis occurred for workers who could have 

been exposed to plutonium. This analysis found additional bioassay data for 1961, 1968, the early 

1980s, and 1994 to present. A 1961 incident that resulted in the release of plutonium is discussed 

below. 

Dose estimates from these bioassays were not found and, because they are not listed in the records, 

doses were probably determined to be less than whatever recording level was used at that time. For 

1961 data, the results were not detectable and the lower limit of detection was 0.2 cpm/sample or 

about 0.8 dpm/sample. Although not specified in these data, other bioassay data indicated that 24-hr 

samples were taken. The results for the 1968 samples were all less than 0.3 dpm/L. MDAs were not 

provided for 1963 and 1966 or for data from the 1980s. Bioassay data from March 30, 1994, had 

isotopic MDAs ranging from 0.01 to 0.04 dpm/sample for 238Pu and 239/240Pu. A number of different 

vendors analyzed plutonium bioassay samples over the years. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 37 of 55

Several individuals who had urine bioassays for plutonium in the 1960s were identified as claimants, 

and their claimant files were reviewed. The analysis found no indication of the bioassay history or 

doses from these bioassays in the records. Therefore, claimant files probably will not contain reliable 

bioassay data. If bioassay data are not available, dose reconstructors should use the following 

approach for assigning potentially missed doses to workers.

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). 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-yr aged mixture. Table 5-14 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.

In addition, dose reconstructors should use the following methods based on exposure period:

• For workers who had the highest possibility of intake (from Table 5-2), for each year of 

possible exposure from 1980 to 2000, when the number of disassemblies was highest, 

assume a single acute intake of 40 DAC-hr (290 pCi) [(6 × 10-12 μCi/ml)(106 ml/m3

)(1.2 

m3

/hr)(40 hr)(106 pCi/μCi)]. Assign 0.1 times this intake for category 2 workers in Table 5-2. 

These intakes are modes of triangular distributions with a minimum of zero and a maximum of 

10 times the mode to account for the possibility of more than one intake per year and possible 

unrepresentativeness of the air-sampling system. 

Table 5-14. 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 agingb: 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 yr of aging. 

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

• For each year from 2001 to the present, for workers with the highest possibility of intake, 

assume a single acute intake of 4 DAC-hr (29 pCi). Assign 0.1 times this intake for category 2 

workers in Table 5-2. Because Pantex used lapel sampling results as criteria for when to 

conduct bioassay during this period, these intakes are the maximum of a triangular distribution 

with a minimum of zero and a mode of one-half the maximum. 
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• For the period from 1958 to 1979 (except 1961, as discussed below), it is unknown what air 

sample levels would have triggered bioassay; however, it is expected that fewer disassemblies 

occurred. In addition, there would have been less oxidation of plutonium metal during the 

earlier years as well. However, because the documentation of the number of disassemblies 

and the contamination levels is not available, it is favorable to the claimant to assign 

unmonitored workers an intake that is the same as the intake from the 1980 to 2000 period in 

accordance with their risk potential from Table 5-2. 

Twenty-three non-incident (reason codes annual, routine, or termination) plutonium urinalysis results 

have been recorded in the Pantex DORMS since 1989; these samples were obtained in 1995 and 

1996. All results were below the detection limit (median = 0.00026 pCi/d). Incident-related urine 

sample results since 1989 were also analyzed as a group and all results except possibly one were 

below the detection limit (median = 0.0000 pCi/d). Although daily urine excretion of type S plutonium 

is a small fraction of the intake, these excretion results are consistent with the 40-DAC-hr intake 

assumed for the 1980 to 2000 period above. 

5.2.4.2 1961 Cell Incident

An incident of plutonium exposure occurred in 1961. The details of the event are classified, but some 

data were available that dose reconstructors can use to estimate intakes. The appropriate intake 

should be applied to all workers known either to have been involved in this event or to have worked in 

the cells or bays in 1961 even if it is not clear that they were involved.

Three people were in the cell at the time of the incident, and all were contaminated. The problem was 

recognized as soon as the incident happened, and the three individuals immediately left the cell. The 

two operators were wearing respirators, but the foreman standing approximately 6 ft from the release 

point was not. Initial contamination readings were as high as 450,000 dpm, but there was no 

contamination outside the cell. There was a statement that an initial urinalysis performed immediately 

after the incident showed no internal deposition of plutonium, but there is no information on exactly 

how long after the incident the urinalysis occurred. A urinalysis within a few hours would be unlikely 

to find anything. In addition, because the material was most likely type S, urinalysis would not have 

been a particularly sensitive indicator of intake. The air sample data in Table 5-15 were obtained from 

reports.

Table 5-15. Plutonium incident air sample counts (dpm/m3

), 1961. 

Location 0-2 hr + 3 d + 4 d

Assembly cell 1,900 1.03 0.43

A cubicle 880 0.55 0.156

B cubicle --- 28 0.185

Equipment room 1,030 Filter paper damage 0.156

Analysis of this incident made the following assumptions:

• The breathing rate of the three workers was 3 m3

/hr (ICRP 1994) or 0.05 m3/min (heavy 

exercise, adult male).

• The workers were exposed for 5 min (assumes time in the cell and immediately following while 

they removed contaminated clothing).

• The workers closest to the pit were exposed to the highest air concentration (1,900 dpm/m3

).
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 39 of 55

• The foreman was not as close to the release, so the second highest air concentration 

(1,030 dpm/m3

) was used for him.

• The air samples were removed for counting 2 hr after the incident, but essentially all activity on 

the filters was obtained in the first 5 min. Therefore, the air concentration breathed by the 

workers in those 5 min was 120/5 min, or 24 times greater than as reported. 

• The material was aged weapons-grade plutonium mixture, absorption type S. 

• Assuming the plutonium was produced 10 yr before the incident, isotopic ratios can be 

obtained from Table 5-14. 

• A protection factor of 5 (0.2) was allowed for the two workers in respirators. This factor 

underestimates the protection of most respirators when worn properly with a good fit. 

However, the quality of the respiratory protection program at that time was not known, and

some intake could have occurred during undressing.

Using these assumptions, an estimation of the intake can be made as follows:

• For the workers wearing respirators:

Total alpha = 0.05 m3

/min × 5 min × 1,900 dpm/m3 × 24 × 0.2 = 2,280 dpm = 1,027 pCi

• For the foreman without a respirator:

Total alpha = 0.05 m3

/min × 5 min × 1,030 dpm/m3 × 24 = 6,180 dpm = 2,784 pCi

Both of these intakes produce expected daily urine excretion far below the likely detection level for the 

urinalysis. The method of analysis is noted as “N.T.A.” The MDA for the analysis was not given, but 

given the state of the art at that time, sample results less than 0.04 dpm/sample are unlikely to be 

different from background. Urinalysis of these three workers occurred 8 d after the incident on 

November 14, 1961. The results were 0, 0.01, and 0.03 dpm. It is favorable to the claimant to use 

the MDA of 0.04 dpm/d as the bioassay result in IMBA (absorption type S), which results in an intake 

of 66,000 pCi. Because this is substantially higher than expected from the air sample data, it should 

be considered a constant upper bound. This is the total alpha intake. Assuming 10-yr-old weaponsgrade plutonium, intakes of the specific radionuclides are: 

Plutonium-239: 52,000 pCi

Plutonium-238: 5,800 pCi

Plutonium-241: 370,000 pCi

Americium-241: 7,700 pCi

In addition to the incident itself, there was decontamination of the cell following the accident. 

Decontamination of Building 12-44-6 Following Radiation Accident on November 6, 1961 (MHSMC

1962) contains information on decontamination activities. Exhibit 3 of that report provides urinalysis 

data for the workers who participated in cleanup operations. Only three samples were 0.04 dpm or 

higher. These workers wore full-face respirators and full protective clothing during decontamination 

activities. The first three individuals listed in Exhibit 3 of the accident report are the workers involved 

in the incident. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 40 of 55

5.2.4.3 1978 Storage Cylinder Incident

Sometime just before November 14, 1978, a nuclear materials inventory of a Nuclear Weapons 

Accident Residue (NWAR) storage cylinder (a concrete cylinder about 6 ft in diameter buried vertically 

and mostly belowground) was attempted. The mounded earth overcap was removed and the cylinder 

was opened. Heavy rainfall occurred during the time the cylinder was open, and the cylinder was 

flooded, which soaked various cans storing radioactive waste. The cans were removed, surveyed, 

and moved to Magazine 4-75 (also referred to as Igloo 75). Although no contamination was found 

during the initial survey, the cans were wet so the alpha survey was ineffective. A subsequent survey 

on November 14 found alpha contamination associated with a small hole in one 11-M can that 

contained mostly plutonium waste (uranium and tritium contamination was also possible). The igloo 

was secured. Potentially exposed workers were given bioassay on or about November 17. 

A special decontamination procedure was developed which included, for example, full anticontamination clothing and full-face respirators, continuous radiation monitoring, high-volume air 

sampling, and HEPA-filtered vacuum cleaners. Cleanup began on January 23, 1979, and concluded 

by February 8. 

Details are provided in MHSMC (1979), which includes the decontamination procedure, results of 

surveys, pictures, notes associated with the original movement of the cans from the cylinder and with 

cleanup of the igloo, and results of bioassay samples. It is possible that claims files for energy 

employees associated with this work do not have these bioassay results; if an energy employee was 

employed at Pantex in 1978 or 1979, the bioassay results in this file should be checked. 

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 yr. 

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 atmosphere 

where diffusion and advection dilute it with outdoor air. 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 inch (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,b, 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 of 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 weapons was infrequent.
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 41 of 55

5.3.1

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 hr or any 

equivalent concentration and time product.

Dose from Radon-222 Progeny

5.3.2

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 [the U.S. Energy 

Research and Development Administration, the U.S. Atomic Energy Commission (AEC), and the 

Manhattan Engineer District].

Radon Monitoring at DOE Facilities

5.3.3

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, 

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. 

Underground Buildings

5.3.4

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

duplicate measurements at 13 locations. Table 5-16 lists complete survey data. 

Radon Concentrations

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 5-17, the average for all buildings was 1.62 ±1.24 pCi/L with a geometric 

mean (median) of 1.37 pCi/L and a 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.

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 5-3). The average ratio was 1.03, which indicates no significant bias.

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. 00 Effective Date: 07/27/2006 Page 42 of 55

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. 

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. Table 5-18 summarizes the statistics from the 1969 radon datasets.
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 43 of 49

Table 5-16. Results of 137 radon measurements in 1990 (UNC Geotech 1990).

RPIS Bu-Ins-Bldg 

Code Building number General description

Gross sq 

ft

No 

flrs

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 3200 1 7.8 01/10/90 02/16/90 Maintenance Shop

0100112-104 12-104 East U 12-104 is bay building 99680 2 7.1 01/10/90 02/16/90 Bay 5

0100112-66 12-66 12-66 is SNM [special nuclear material] warehouse 25900 1 5.4 01/10/90 02/16/90 Center Of building

01001FS-01 FS-1 U FS-01 is an earth covered storage facility for HE 5364 1 5.2 5.9 01/10/90 02/16/90 Break room

0100112-23 12-23 12-23 3200 1 4.5 01/10/90 02/16/90 North wall middle

0100112-15 12-15 12-15 16800 1 4.2 01/10/90 02/16/90 Training Room 103

0100112-60 12-60 12-60 is Mass Properties Facility 8600 1 4.1 01/10/90 02/16/90 Office - Vault

0100112-79 12-79 12-79 is warehouse/loading dock 28700 1 3.2 01/10/90 02/16/90 South Warehouse Area

0100112-104 12-104 U 12-104 is bay building 99680 2 3 01/10/90 02/16/90 106f R Collins Office

0100112-15 12-15 12-15 16800 1 2.8 01/10/90 02/16/90 Training Graphic Arts

0100112-58 12-58 U 12-58 is bay building 2600 1 2.8 01/10/90 02/16/90 East Wall Between Bay 4&5

0100112-15 12-15 12-15 16800 1 2.6 01/10/90 02/16/90 Training Office

0100112-44 12-44-4 U 12-44 is Gravel Gertie cell building 27100 1 2.6 01/10/90 02/16/90 Round Room

0100112-6 12-6 12-6 23700 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

U 12-26 is bay building 87500 1 2.4 01/10/90 02/16/90 Tooling Warehouse Office

0100112-26 12-26 U 12-26 also has pit vault 87500 1 2.3 01/10/90 02/16/90 Bay 28

0100112-35 12-35 12-35 13400 1 2.3 2.1 01/10/90 02/16/90 Area Mechanics Office

0100111-17 11-17 U 12-17is bay building 6700 1 2.2 01/10/90 02/16/90 Bay 7 Lab

0100112-1 12-1 12-1 is cafeteria/change room 27600 2 2.2 01/10/90 02/16/90 Lepor Colony

0100112-5 12-5 12-5 74400 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 34100 1 2.1 2.2 01/10/90 02/16/90 Break room

0100112-61 12-61 12-61 24000 1 2.1 01/10/90 02/16/90 Warehouse Area

0100112-44 12-44 U 12-44 is Gravel Gertie cell building 27100 1 2 01/10/90 02/16/90 Cell 8

0100112-64 12-64 U 12-64 is bay building 32000 1 2 01/10/90 02/16/90 Bay 9

01001STATION 30 Zone 4 

Station 30

U 4-30 is underground igloo SNM storage building 2 01/10/90 02/16/90 Control Room

0100112-28 12-28 3500 1 1.9 01/10/90 02/16/90 Quality Hallway

0100112-37 12-37 22700 1 1.9 01/10/90 02/16/90 Room 120 Control Room

0100112-42 12-42 Radiation SAFET 47400 2 1.9 01/10/90 02/16/90 12-42 Cr

0100112-84 12-84 U Bay 1 1 1.9 01/10/90 02/16/90 Bay 13

0100111-29 11-29 4200 1 1.8 01/10/90 02/16/90 Office

0100112-37 12-37 22700 1 1.8 01/10/90 02/16/90 Room 121 Tech Doc

0100112-37 12-37 22700 1 1.8 01/10/90 02/16/90 Room 112 Mail Room

0100112-5 12-5 74400 1 1.8 01/10/90 02/16/90 Plant Eng Annette Covington

0100112-5 12-5 74400 1 1.8 1.8 01/10/90 02/16/90 Plant Design Eng

0100116-2 16-2 Courier 20072 1 1.8 01/10/90 02/16/90 114 Break room

0100110-9 10-9 15500 1 1.7 01/10/90 02/16/90 Office

0100111-27 11-27 5100 2 1.7 01/10/90 02/16/90 Room 119

0100112-50 12-50 1400 1 1.7 01/10/90 02/16/90 West Of 12-50 Door

0100112-75 12-75 21862 1 1.7 01/10/90 02/16/90 Desk Lieutenants Office

0100112-99 12-99 60716 1 1.7 01/10/90 02/16/90 Break Room

0100111-20 11-20 16600 1 1.6 01/10/90 02/16/90 Office South Wall

0100112-16 12-16 5000 1 1.6 1.1 01/10/90 02/16/90 Plastic Shop Office

0100112-2 12-2 Safety 13456 1 1.6 01/10/90 02/16/90 Dosimetry Lab Room 157

0100112-2B 12-2B 3220 1 1.6 01/10/90 02/16/90 North Wall By Clock

0100112-42 12-42 47400 2 1.6 01/10/90 02/16/90 South Vault

0100112-6 12-6 23700 1 1.6 01/10/90 02/16/90 Room 103 Cafeteria
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 44 of 49

RPIS Bu-Ins-Bldg 

Code Building number General description

Gross sq 

ft

No 

flrs

Radon 

(pCi/L)

Dupl. 

radon 

(pCi/L)

Install 

date

Retrieve 

date Room

0100112-84 12-8411 1.6 01/10/90 02/16/90 125d

0100112-36 12-36 294001 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 18500 3 1.5 01/10/90 02/16/90 HE Side

0100112-96 12-96U Gravel Gertie 78651 1.5 01/10/90 02/16/90 Round Room

0100112-99 12-99U Bay 607161 1.5 01/10/90 02/16/90 Bay 7

0100112-26 12-26U Bay 875001 1.4 01/10/90 02/16/90 Bay 30

0100112-107 12-107 South 100001 1.4 01/10/90 02/16/90 By C.L. Saban’s Office

0100112-44E 12-44-EU 12-44 is Gravel Gertie cell building 19001 1.4 01/10/90 02/16/90 Marion Everett’s Office

0100112-6 12-6 237001 1.4 01/10/90 02/16/90 Room 112 Tom Folks

0100112-68 12-68 359001 1.4 01/10/90 02/16/90 Machine Shop Office

0100112-69 12-69 98001 1.4 01/10/90 02/16/90 Emmett Hallway

0100111-5 11-5 U Bay 9000 2 1.3 01/10/90 02/16/90 Control Bay

0100112-2 12-2 134561 1.32 01/10/90 02/16/90 Medical Office

0100112-24 12-24 SouthU Bay 1.3 01/10/90 02/16/90 Bay 27

0100112-49 12-49 39001 1.3 01/10/90 02/16/90 Electronics Room

0100112-6 12-6 237001 1.3 01/10/90 02/16/90 104 Room Standards

0100116-12 16-12 285002 1.3 01/10/90 02/16/90 Employment

0100112-61 12-61 240001 1.2 01/10/90 02/15/90 Break Room

0100112-107 12-107 North 100001 1.2 01/10/90 02/16/90 Preventive Maint Section

0100112-11 12-11 29001 1.2 01/10/90 02/16/90 Data Management

0100112-42 12-42 474002 1.21 01/10/90 02/16/90 Upstairs Assembly Ops

0100112-6 12-6 237001 1.2 01/10/90 02/16/90 Room 121 Elaine Miller

0100112-64 12-64 U Bay 32000 1 1.2 1.3 01/10/90 02/16/90 Bay 15

0100112-69 12-69 98001 1.2 01/10/90 02/16/90 DOE Office

0100112-9 12-9 185003 1.2 01/10/90 02/16/90 Office

0100112-97 12-97B 100001 1.2 01/10/90 02/16/90 By Refrigerator in Break Room

0100112-98 12-98U Gravel Gertie 343581 1.2 1.2 01/10/90 02/16/90 Cell 2

0100112-21 12-21 Gas Lab 293002 1.1 01/10/90 02/16/90 Break Area

0100111-18 11-18 Control Room 15001 1.1 01/10/90 02/16/90 Control Room

0100111-2 11-2 96002 1.1 01/10/90 02/16/90 110

0100111-51 11-51 116001 1.1 01/10/90 02/16/90 Office

0100112-32 12-32 South Side 76001 1.1 01/10/90 02/16/90 Above Phone on Ramp

0100112-98 12-98-3U Gravel Gertie 343581 1.1 0.8 01/10/90 02/16/90 Round Room

0100111-36 11-36 500021 01/10/90 02/16/90 Office

0100111-50 11-50 2215111 01/10/90 02/16/90 Room 110 Office

0100112-100 12-100 436011 01/10/90 02/16/90 Environmental Protection

0100112-11A 12-11A 520011 01/10/90 02/16/90 Quality Hallway

0100112-5 12-5 7440011 01/10/90 02/16/90 General Stores Office

0100112-52B 12-52B1 01/10/90 02/16/90 Meteorology

0100112-52C 12-52C 360011 0.9 01/10/90 02/16/90 Meteorology

0100112-6 12-6 2370011 01/10/90 02/16/90 Assy Eng Office

0100112-6 12-6 2370011 01/10/90 02/16/90 Room 700 Stoddard

0100112-61 12-61 24000 1 1 01/10/90 02/16/90 Office Area

0100112-82 12-82U Bay 680011 01/10/90 02/16/90E-Bay Office

01001STATION C Station C1 01/10/90 02/16/90 West Wall By Exit

0100112-21 12-21 2930021 01/10/90 02/16/90X-Ray Office

0100112-2B 12-2B 32201 0.9 01/10/90 02/16/90 South Wall - Nancy’s Office

0100112-104 12-104U Bay 996802 0.9 01/10/90 02/16/90 Bay 13

0100112-104-EAS 12-104 EastU Bay 0.9 01/10/90 02/16/90 Bay 2

0100112-106 12-106 54001 0.9 01/10/90 02/16/90 Across From Room 105 Janitor

0100112-24 12-24 NorthU Bay 0.9 01/10/90 02/16/90 Bay 10

0100112-31 12-31U Bay 76001 0.9 0.8 01/10/90 02/16/90 Bay 3 Outside
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 45 of 49

RPIS Bu-Ins-Bldg 

Code Building number General description

Gross sq 

ft

No 

flrs

Radon 

(pCi/L)

Dupl. 

radon 

(pCi/L)

Install 

date

Retrieve 

date Room

0100112-64 12-64 32000 1 0.9 01/10/90 02/16/90 D&I Office

0100112-99 12-99 607161 0.9 01/10/90 02/16/90 105-F Manufacturing Office

0100112-84 12-84-East11 0.8 01/10/90 02/16/90 Break Room

0100112-84 12-84U Bay11 0.8 01/10/90 02/16/90 Bay 4

0100112-84 12-8411 0.8 01/10/90 02/16/90 Break Room

0100112-84 12-84U Bay11 0.8 1.3 01/10/90 02/16/90 Bay 7

0100112-101 12-101 Portable Maint 53341 0.8 01/10/90 02/16/90 By Sign-Out Board

0100112-102 12-102 57781 0.8 01/10/90 02/16/90 Tech Applications

0100112-103 12-103 236081 0.8 01/10/90 02/16/90 Smoking Area

0100112-104 12-104 West 996802 0.8 01/10/90 02/16/90 128f Manufacturing Office

0100112-111 12-111 74161 0.8 01/10/90 02/16/90 Carpenter Shop

0100112-112 12-112 65251 0.8 01/10/90 02/16/90 Camera Room

0100112-17 12-17 325002 0.8 01/10/90 02/16/90 Break Area

0100112-19 12-19 EAST 325002 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 20001 0.8 01/10/90 02/16/90 Transportation

0100112-3512-35 134001 0.8 01/10/90 02/16/90 Utilities Console Room

0100112-39 12-39 Fire Department 82001 0.8 01/10/90 02/16/90 Sleeping Room

0100112-41A 12-41A 30001 0.8 01/10/90 02/16/90 North Wall

0100112-42A 12-42A 199001 0.8 01/10/90 02/16/90Outer Wall by Sandia Sign

0100112-5 12-5 74400 1 0.8 01/10/90 02/16/90 Master Mechanics

0100112-59 12-59 83001 0.8 01/10/90 02/16/90 Chem Lab Office

0100112-5C 12-5C 217002 0.8 01/10/90 02/16/90 Sheet Metal Shop

0100112-84 12-84U Bay11 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 12-86U Bay 0.8 01/10/90 02/16/90 Bay 10

0100112-97 12-97A 100001 0.8 01/10/90 02/16/90 By Clock in Hallway

0100112-97 12-97C 100001 0.8 01/10/90 02/16/90 Above Fire Ext by Copier

0100112-99 12-99U Bay 607161 0.8 01/10/90 02/16/90 Bay 6

0100116-1 16-1 VMF 542001 0.8 01/10/90 02/16/90 Office

0100116-12 16-12 285002 0.8 01/10/90 02/16/90 Purchasing

01001STATION 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. 00 Effective Date: 07/27/2006 Page 46 of 55

Table 5-17. Summary statistics of 1990 radon measurements. 

Parameter All buildings

Underground 

buildings

Aboveground 

buildings

Mean (pCi/L) 1.62 1.81 1.56

SD (pCi/L) 1.24 1.35 1.21

CV 0.77 0.75 0.77

GeoMean (pCi/L) 1.37 1.51 1.33

GSD 1.68 1.75 1.66

Min (pCi/L) 0.8 0.8 0.8

Max (pCi/L) 8.1 7.1 8.1

Max/Min 10.1 8.9 10.1

Count 137 31 106

Figure 5-3. Absolute differences between duplicate radon 

measurements. 

Table 5-18. Summary of 1969 radon measurements in 

Cells 1 to 6. 

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 std. dev. 10.89 0.0436
Document No. ORAUT-TKBS-0013-5 Revision No. 00 Effective Date: 07/27/2006 Page 47 of 55

5.3.5

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

by multiplying the radon concentration 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: 

Working Level-Months

PAEC = C × F/100 pCi/L/WL (5-21)

where C is the radon concentration in picocuries per liter and PAEC is in WLs. 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). 

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 = 0.072 WLM/yr (5-22)

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 their WLMs are one-tenth of those 

from Equation 5-22. 

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

exposure distribution is lognormal. Parameter 1 is the median value in WLM from Equation 5-22. 

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. 

5.4 INTAKE SUMMARY

Table 5-19 provides a summary of the default intakes developed in Sections 5.2 and 5.3. These 

default intakes should be used when bioassay data are missing or inadequate. 
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 48 of 49

Table 5-19. Summary of default intakes. 

Work or worker category Period Material Mode

Absorption 

type

Intake 

(pCi/d or noted) Distribution GSD

1 Production techs, QA techs, RSTs, 

assemblers/disassemblers

1956–present Tritium Chronic inhalation/

absorption 

NAa From Table 5-6 Triangular

(0, mean from Table 5-

6, max from Table 5-6)

NA

2 Production techs, QA techs, RSTs, 

assemblers/disassemblers

1961–1993 DU or U Chronic inhalation

or 

chronic ingestion 

M 

or

S

1.3 

19

Lognormal 3 

Soluble

or

insoluble

4.4 

44 

3 Production techs, QA techs, RSTs, 

assemblers/disassemblers

1994–present DU or U Use 20% of values in row 2

4 Production techs, QA techs, RSTs, 

assemblers/disassemblers

1980–2000 Th-232 or Th-228 One acute intake per year S 48 pCi Triangular (0, 48, 480) NA

5 Production techs, QA techs, RSTs, 

assemblers/disassemblers

2001–present Th-232 or Th-228 One acute intake per year S 2.4 pCib Triangular (0, 2.4, 4.8) NA

6 Production techs, QA techs, RSTs, 

assemblers/disassemblers

1958–1979 Pu One acute intake per year S 290 pCi Triangular (0, 290, 

2900)

NA

7 Production techs, QA techs, RSTs, 

assemblers/disassemblers

1980–2000 Pu One acute intake per year S 290 pCi Triangular (0, 290, 

2900)

NA

8 Production techs, QA techs, RSTs, 

assemblers/disassemblers

2001–present Pu One acute intake per year S 14.5 pCi Triangular (0, 14.5, 29) NA

9 Production techs, QA techs, RSTs, 

assemblers/disassemblers

1958–present Radon Chronic NA 0.072 WLM/yrc Lognormal 3

10 Category 2 in Table 5-2 or some risk 1961–1993 DU or U 10% of values in row 2

11 Category 2 in Table 5-2 or some risk 1994–present DU or U 2% of values in row 2

12 Category 2 in Table 5-2 or some risk 1980–2000 Th-232 or Th-228 One acute intake per year S 10% of value in row 4i Triangular (0, 4.8, 48) NA

13 Category 2 in Table 5-2 or some risk 2001–present Th-232 or Th-228 One acute intake per year S 10% of value in row 5b Triangular (0, 0.24, 

0.48)

NA

14 Category 2 in Table 5-2 or some risk 1958–1979 Pu One acute intake per year S 10% of value in row 6 Triangular (0, 29, 290) NA

15 Category 2 in Table 5-2 or some risk 1980–2000 Pu One acute intake per year S 10% of value in row 7 Triangular (0, 29, 290) NA

16 Category 2 in Table 5-2 or some risk 2001–present Pu One acute intake per year S 10% of value in row 8 Triangular (0, 1.45, 2.9) NA

17 Category 2 in Table 5-2 or entry into 

bays and Gravel Gerties

1958–present Radon Chronic NA 0.0072 WLM/yrc Lognormal 3

18 Machinists 1960–1965 DU Chronic inhalation M or S 13 Lognormal 3

19 Machinists 1960–1965 DU Chronic ingestion Soluble or 

insoluble

0.4 Lognormal 3

20 Burning ground techs. /operators 1952–present DU Chronic inhalation M or S 130 Constant NA

21 Firing site techs/operators 1959–1986 DU Chronic inhalation M or S Per Table 5-11 Constant NA

22 Firing site techs/operators Unknown, apply once in 

employment history

Th-232 in 

equilibrium 

Acute M or S 14 pCi Constant NA

23 Firing Site 5 cleanup if no bioassay 

was obtained

1994 to 1999 DU Chronic S 1.6 pCi/d Triangular (0, 1.6, 3.2) NA

24 Firing Site 23 cleanup if no bioassay 

was obtained

Nov. 10, 1983; Jan. 2, 

1984, 85, 86, 87, 88; July 

1, 1984, 85, 86, 87

DU Acute S 3.2 E5 pCi Lognormal 3

25 Involved in 1961 accident in bay Nov. 1961 Pu Acute S (all pCi) Pu-238: 5,800 , 

Pu-239: 52,000, Pu-241: 

370,000, Am-241: 7,700 

Constant NA

a. NA = not applicable.

b. Unless there is a recorded internal dose exceeding 5 mrem, in which case use 0.86 pCi/mrem.

c. Can be prorated by month.
Document No. ORAUT-TKBS-0013-5 Revision No. 00 PC-1 Effective Date: 02/08/2007 Page 49 of 55

REFERENCES

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Hanger-Silas Mason Company, Pantex Plant, Amarillo, Texas. [SRDB Ref ID: 13097]

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Brake, R., 1989, untitled letter to G. Runkle (U.S. Department of Energy), University of California, Los 

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BWXT Pantex, 1992, Technical Basis for the Internal Dosimetry Program and the DOE Pantex 

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BWXT Pantex, 2004, Pantex Plant Radiological Investigation Volume I of II Report for the U.S. 

Department of Energy, National Nuclear Security Administration, Pantex Plant, Amarillo, 

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20778]

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, Washington, D.C., July.

DOE (U.S. Department of Energy), 1988, Radiation Protection for Occupational Workers, Order 

5480.11, Office of Environment, Safety and Health, Washington D.C., December 21. [SRDB 

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GLOSSARY

aged

In the context of mixtures of plutonium isotopes, aging refers to the time since 241Am separated 

from the plutonium mixture and then grew back in from decay of 241Pu. 

U.S. Atomic Energy Commission (AEC)

Original agency established for nuclear weapons and power production; a predecessor to the 

U.S. Department of Energy.

dose of record

The dose files provided by DOE to NIOSH as part of the individual worker files. 

DU

Depleted uranium; uranium having less than the natural mass of 235U; used as components in 

nuclear weapons or as a surrogate for enriched uranium or plutonium in testing. 

Gertie

A facility covered with crushed gravel used to suppress potential radioactive contamination 

from the accidental explosion of a nuclear weapon during assembly or disassembly. Also 

referred to as a Gravel Gertie. 

equilibrium factor (F)

In relation to the potential alpha energy of radon and its progeny in air, the ratio of the 

equilibrium equivalent concentration (EEC) to the actual activity concentration of radon.

hydroshot

Detonation of a mixture of explosives and DU used as a quality control technique for 

measuring the performance of plastic-bonded explosives.

likely noncompensable, maximum internal dose approach

An efficiency method of assigning organ (or in some cases whole body) dose whereby the 

organ is assigned the maximum plausible dose that could have been received. If the resulting 

probability of causation is 45% or less at the 99% confidence level, the dose reconstruction is 

considered complete. Also called the overestimate approach. 

nuclear emulsion

Often referred to as “NTA” film and normally used to measure personnel dose from neutron 

radiation; in addition, Pantex apparently used NTA film at one time to measure alpha radiation 

from radon progeny in air.

potential alpha energy concentration (PAEC)

The kinetic energy potentially released in a unit volume of air by alpha particles emitted by the 

short-lived radioactive progeny of 222Rn (i.e., 218Po and 214Po) or 220Rn (i.e., 216Po, 212Bi, and 

212Po). PAEC is expressed in working levels (WL).

potential alpha energy exposure (PAEE) 

The average potential alpha energy concentration (PAEC) to which a worker is exposed, 

multiplied by the time of exposure in working months of 170 hours; that is, PAEE = PAEC × 

time. PAEE is expressed in working level months (WLM).
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Parameter 1

The column in the IREP template where the dose reconstructor will enter the calculated dose. 

Multiple entries based on year of employment, type of radiation, and appropriate energy 

ranges; internal and external exposures are possible.

Parameter 2

The column in the IREP template where the dose reconstructor will enter the lower limit of the 

dose distribution based on the radiation type and the dose distribution type.

progeny

Radionuclides that result from decay of a parent radionuclide.

radon

Unless otherwise specified, the isotope 222Rn.

thoron

The isotope 220Rn.

working level (WL)

The unit of potential alpha energy concentration (PAEC), 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 will result in the ultimate emission of 130,000 MeV of alpha energy (1 WL = 

2.083 × 10-5 J/m3

) (10 C. F. R. pt. 835). 

working level month (WLM)

The unit of potential alpha energy exposure (PAEE), defined as exposure for 1 working month 

(of 170 hours) to an airborne concentration of 1 WL. (1 WLM = 1 WL × 170 hours = 0.00354 

J.h/m3

).