Page 1 of 55

DOE Review Release 02/16/2017

Paducah Gaseous Diffusion Plant – 

Occupational External Dose

ORAUT-TKBS-0019-6 Rev. 05

Effective Date: 02/13/2017

Supersedes: Revision 04

Subject Expert(s): Jodie L. Phillips

Document Owner 

Approval: Signature on File

Jodie L. Phillips, Document Owner

Approval Date: 02/07/2017

Concurrence: Wade C. Morris Signature on File for

John M. Byrne, Objective 1 Manager

Concurrence Date: 02/03/2017

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 02/03/2017

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 02/06/2017

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 02/13/2017

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 2 of 55

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

08/24/2004 00 New technical basis document for the Paducah Gaseous Diffusion 

Plant – Occupational External Dose. First approved issue. Initiated 

by Jay J. Maisler.

03/29/2005 01 Approved issue of Revision 01. Initiated by Jay J. Maisler.

08/30/2006 02 Revision initiated to address comments from earlier unresolved 

comments received on TBD. Approved issue of Revision 02. 

Incorporates additional information obtained through additional data 

capture for the Paducah Gaseous Diffusion Plant. Incorporates 

revised standard language into Purpose section and adds a Scope 

section. This revision addresses comments from the Worker 

Outreach meeting with the United Steelworkers Local 5-550 and 

Security, Police, Fire Professionals of America Local 111 held on 

February 10, 2005. Constitutes a total rewrite of the document. This 

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

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

Initiated by Paul A. Szalinski.

04/04/2007 03 Approved Revision 03 initiated to incorporate Attributions and 

Annotations section. Constitutes a total rewrite of the document. 

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

is required. No further changes occurred as a result of formal 

internal review. Training required: As determined by the Task 

Manager. Initiated by Daniel S. Mantooth.

08/24/2012 04 This document was predominantly revised to address SC&A 

concerns about the document as identified in SC&A-TR-TASK1-

0016, Review of the NIOSH Site Profile for the Paducah Gaseous 

Diffusion Plant. Wording was revised throughout the document to 

add clarity. Acronyms and Abbreviations, References, and Glossary

Sections have been updated. Table 6-1 has been revised to change 

four-element film to two-element film from 1953 through July 1, 1960, 

and to add an annual badge exchange option beginning in 1989. 

Building start dates in Table 6-3 have been clarified. Guidance for 

assigning neutron dose has been revised. Guidance for 99Tc 

exposure has been added as Attachment A. Guidance for coworker 

dose assignment based on ORAUT-OTIB-0031 has been 

incorporated into this TBD as Attachment B. The application of 

construction trade worker doses has been added. Guidance for skin 

dose assignment based on ORAUT-OTIB-0017 has been 

incorporated into this TBD as Attachment C. Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training required: As determined by the Objective 

Manager. Initiated by Jodie L. Phillips.
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 3 of 55

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

02/13/2017 05 Revision initiated to update direction in Sections 6.3.4.2 and 6.4.1 to 

apply exposure to organ dose conversion factors to recorded deep 

dose results for workers for the period through 1988. For 1989 and 

after, reported deep dose values are considered equivalent to 

Hp(10). Incorporates formal internal and NIOSH review comments. 

Added dose conversion factor for the skin in Attachment C. 

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

determined by the Objective Manager. Initiated by Jodie L. Phillips. 
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 4 of 55

TABLE OF CONTENTS

SECTION TITLE PAGE

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

6.1 Introduction .............................................................................................................................. 8

6.1.1 Purpose........................................................................................................................ 9

6.1.2 Scope........................................................................................................................... 9

6.1.3 Special Exposure Cohort.............................................................................................. 9

6.2 Basis of Comparison................................................................................................................ 9

6.3 Dose Reconstruction Parameters........................................................................................... 10

6.3.1 Administrative Practices ............................................................................................. 10

6.3.2 Dosimetry Technology................................................................................................ 15

6.3.2.1 Beta/Photon Dosimeters .............................................................................. 15

6.3.2.2 Neutron Dosimeters ..................................................................................... 16

6.3.3 Calibration .................................................................................................................. 16

6.3.3.1 Beta/Photon Dosimeters .............................................................................. 17

6.3.3.2 Neutron Dosimeters ..................................................................................... 18

6.3.4 Workplace Radiation Fields ........................................................................................ 18

6.3.4.1 Beta/Photon Fields....................................................................................... 18

6.3.4.2 Workplace Beta/Photon Dosimeter Response.............................................. 20

6.3.4.3 Neutron Fields.............................................................................................. 20

6.3.4.4 Workplace Neutron Dosimeter Response .................................................... 21

6.4 Adjustments to Recorded Dose.............................................................................................. 22

6.4.1 Photon Dose............................................................................................................... 22

6.4.2 Nonpenetrating Dose.................................................................................................. 22

6.4.3 Neutron Dose ............................................................................................................. 22

6.5 Missed Dose .......................................................................................................................... 22

6.5.1 Estimating Missed and Unmonitored Photon Deep Dose............................................ 23

6.5.1.1 Non-Construction Trade Workers................................................................. 23

6.5.1.2 Construction Trade Workers ........................................................................ 24

6.5.2 Estimating Missed and Unmonitored Shallow Dose.................................................... 24

6.5.3 Estimating Missed and Unmonitored Neutron Dose.................................................... 25

6.6 Uncertainty............................................................................................................................. 25

6.7 Dose Reconstruction.............................................................................................................. 26

6.8 Organ Dose............................................................................................................................ 26

6.9 Attributions and Annotations .................................................................................................. 26

References ........................................................................................................................................ 28

Glossary ............................................................................................................................................ 33

ATTACHMENT A EXTERNAL EXPOSURE TO TECHNETIUM-99 ............................................. 40
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 5 of 55

ATTACHMENT B COWORKER DOSE ASSIGNMENT............................................................... 45

ATTACHMENT C SKIN DOSE ASSIGNMENT............................................................................ 53

LIST OF TABLES

TABLE TITLE PAGE

6-1 Dosimeter types, periods of use, exchange frequencies, MDLs, and potential annual 

missed doses.........................................................................................................................13

6-2 Major radiation sources, half-lives, energies, and abundances .............................................. 19

6-3 Buildings and periods of operation ......................................................................................... 19

6-4 Dose fractions for PORTS HEU storage vault in Building 345 near an unshielded 252Cf 

source....................................................................................................................................21

6-5 Missed and unmonitored neutron dose assignment ............................................................... 25

A-1 Facilities with 99Tc exposure potential .................................................................................... 41

A-2 Job titles for workers with possible 99Tc exposure .................................................................. 42

B-1 Missed external doses ...........................................................................................................48

B-2 Annual external coworker doses modified to account for missed dose................................... 50

B-3 Annual external penetrating coworker doses modified in accordance with ORAUT-OTIB0052.......................................................................................................................................52

C-1 Examples of skin dose assignments for badge readings in 1970............................................ 54

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Estimated dosimeter photon response characteristics............................................................ 15

6-2 Results of neutron spectrum measurements about 24 in. in front of 93%–96% HEU 

cylinders.................................................................................................................................21

6-3 Historical distribution of deep dose equivalent........................................................................ 23

6-4 Historical distribution of shallow dose equivalent.................................................................... 25
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ACRONYMS AND ABBREVIATIONS

AWE atomic weapons employer

cm centimeter

cpm counts per minute

CTW construction trade worker

d day

DCF dose conversion factor

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

HEU highly enriched uranium

Hp(d) personal dose equivalent at depth d in millimeters in tissue

Hp(0.07) personal dose equivalent at 0.07 millimeters depth in tissue

Hp(10) personal dose equivalent at 10 millimeters depth in tissue

hr hour

IARC International Agency for Research on Cancer

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1 thousand electron-volts

LOD limit of detection

MDL minimum detection level

MED Manhattan Engineer District

MeV megaelectron-volt, 1 million electron-volts

mg milligram

min minute

mm millimeter

mrem millirem

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH-Division of Compensation and Analysis Claims Tracking System

NTA nuclear track emulsion, type A

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

OW open window

PER program evaluation report

PGDP Paducah Gaseous Diffusion Plant
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POC probability of causation

PORTS Portsmouth Gaseous Diffusion Plant

PPE personal protective equipment

QF quality factor

RU recycled uranium

S shielded

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

TEPC tissue-equivalent proportional counter

TLD thermoluminescent dosimeter

TLND thermoluminescent neutron dosimeter

U.S.C. United States Code

yr year

§ section
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6.1 INTRODUCTION

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

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

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

dose reconstructions at particular Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

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

information is obtained about the affected DOE or AWE facility(ies). These documents may be used 

to assist NIOSH staff in the evaluation of Special Exposure Cohort (SEC) petitions and the completion 

of the individual work required for each dose reconstruction.

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

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

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7384l(5) 

and (12) of the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA), respectively. An AWE facility means “a facility, owned by an atomic weapons employer, 

that is or was used to process or produce, for use by the United States, material that emitted radiation 

and was used in the production of an atomic weapon, excluding uranium mining or milling.” 42 U.S.C. 

§ 7384l(5). On the other hand, a DOE facility is defined as “any building, structure, or premise, 

including the grounds upon which such building, structure, or premise is located ... in which 

operations are, or have been, conducted by, or on behalf of, the [DOE] (except for buildings, 

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” 

and with regard to which DOE has or had a proprietary interest, or “entered into a contract with an 

entity to provide management and operation, management and integration, environmental 

remediation services, construction, or maintenance services.” 42 U.S.C. § 7384l(12). The Department 

of Energy (DOE) determines whether a site meets the statutory definition of an AWE facility and the 

Department of Labor (DOL) determines if a site is a DOE facility and, if it is, designates it as such.

Accordingly, a Part B claim for benefits must be based on an energy employee’s eligible employment 

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

period and location (i.e., covered employee). After DOL determines that a claim meets the eligibility 

requirements under EEOICPA, DOL transmits the claim to NIOSH for a dose reconstruction. 

EEOICPA provides, among other things, guidance on eligible employment and the types of radiation 

exposure to be included in an individual dose reconstruction. Under EEOICPA, eligible employment 

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures be incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction.

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

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

dose reconstructions for covered employees (NIOSH 2010):

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

conventional structures

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

reasons
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6.1.1 Purpose

The purpose of this technical basis document (TBD) is to provide technical data and other key 

information that may be used in the evaluation of external occupational dose for EEOICPA claimants 

who were employed at the Paducah Gaseous Diffusion Plant (PGDP).

6.1.2 Scope

PGDP workers, especially those during the peak production decades (1950s, 1960s, and 1970s), 

have been exposed to radiation types and energies associated with enrichment of natural and 

recycled uranium (RU). PGDP used facility and individual worker monitoring methods to measure and 

control radiation exposure to workers (UCC 1976). Before about July 1960, personnel dosimeters 

were not assigned to all workers (PACE and University of Utah 2000). Records of radiation dose to 

individuals who wore dosimeters are available beginning in 1953. Doses from these dosimeters were 

recorded at the time of measurement, routinely reviewed by PGDP operations and radiation safety 

personnel for compliance with radiation control limits, and routinely made available to individual 

workers. OCAS-IG-001, External Dose Reconstruction Implementation Guideline (NIOSH 2007), 

indicates that these represent the highest quality records for assessment and reconstruction of doses.

Initial radiation dosimetry practices were based on experience from several decades of radium and 

X-ray medical diagnostic and therapy applications. In general, these practices were well advanced at 

the start of the Manhattan Engineer District (MED) program to develop nuclear weapons, which began 

on August 13, 1942.

6.1.3 Special Exposure Cohort

PGDP is one of the original sites that was designated by Congress as part of the SEC under 

EEOICPA [42 U.S.C. § 7384l(14)]. This designation is as follows:

(A) The employee was so employed for a number of work days aggregating at least 

250 work days before February 1, 1992, at a gaseous diffusion plant located in 

Paducah, Kentucky, Portsmouth, Ohio, or Oak Ridge, Tennessee, and, during such 

employment— 

(i) was monitored through the use of dosimetry badges for exposure at the plant of 

the external parts of employee’s body to radiation; or

(ii) worked in a job that had exposures comparable to a job that is or was monitored 

through the use of dosimetry badges. 

Dose reconstruction guidance in this TBD is presented to provide a technical basis for dose 

reconstructions for nonpresumptive cancers that are not covered in the SEC class through 

January 31, 1992. Dose reconstructions for individuals who were employed at PGDP before 

February 1, 1992, but who do not qualify for inclusion in the SEC, may be performed using this 

guidance as appropriate.

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

justification, or clarification of the associated information, are presented in Section 6.9.

6.2 BASIS OF COMPARISON

Since the start of the MED on August 13, 1942, various radiation dose concepts and quantities have 

been used to measure and record occupational dose. The basis of comparison for reconstruction of 
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dose is the personal dose equivalent, Hp(d), where d identifies the depth (in millimeters) and 

represents the point of reference for dose in tissue. For weakly penetrating radiation of significance to 

skin dose, d = 0.07 mm and is noted as Hp(0.07). For penetrating radiation of significance to wholebody dose, d = 10 mm and is noted as Hp(10). Both Hp(0.07) and Hp(10) are the radiation quantities 

the International Commission on Radiation Units and Measurements (ICRU) has recommended for 

use as operational quantities for radiological protection (ICRU 1998). In addition, Hp(0.07) and 

Hp(10) are the radiation quantities the DOE Laboratory Accreditation Program (DOELAP) has used to 

accredit the Department’s personnel dosimetry systems since the 1980s (DOE 1986). The 

International Agency for Research on Cancer (IARC) Three-Country Combined Study (Fix et al. 1997) 

and the IARC Collaborative Study (Thierry-Chef et al. 2002) selected Hp(10) as the quantity to assess 

error in historical recorded whole-body dose for workers in IARC nuclear worker epidemiologic 

studies. This TBD uses Hp(10) and Hp(0.07) as deep dose and shallow dose, respectively.

6.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of beta, photon (X- and gamma rays), and neutron energies and geometries of 

exposure, and the characteristics of PGDP dosimeter responses, are crucial for assessment of the 

original recorded doses. Bias and uncertainty for current dosimetry systems are typically well 

documented (MMES 1994). The performance of current dosimeters can often be compared to the 

performance of dosimetry systems in the same, or highly similar, facilities or workplaces. In addition, 

current performance testing techniques can be applied to earlier dosimetry systems to achieve a 

consistent evaluation of all dosimetry systems. Dosimeter response characteristics for radiation types 

and energies in the workplace are crucial to the overall analysis of error in recorded dose.

Overall, accuracy and precision of the original recorded individual worker doses and their 

comparability to be considered in using NIOSH (2007) guidelines depend on the following factors 

(Fix et al. 1997):

• Administrative practices adopted by facilities to calculate and record personnel dose based 

on technical, administrative, and statutory compliance considerations; 

• Dosimetry technology, including physical capabilities of the dosimetry system such as the 

response to different types and energies of radiation, in particular in mixed radiation fields; 

• Calibration of the respective monitoring systems and similarity of methods of calibration to 

sources of exposure in the workplace; and

• Workplace radiation fields that could include mixed types of radiation, variations in exposure 

geometries, and environmental conditions. 

The accuracy of PGDP worker doses has been the subject of DOE investigations (PACE and 

University of Utah 2000). An evaluation of the original recorded doses as available, combined with 

detailed examinations of workplace radiation fields, is the recommended option to provide the best 

estimate of Hp(0.07) for the shallow dose and Hp(10) for the deep dose for individual workers.

6.3.1 Administrative Practices

The PGDP radiation monitoring program has used portable instruments, contamination surveys, zone 

controls, and personnel dosimeters to measure exposure in the workplace (Ely et al. 1957; UCNC

1957a, 1957b; Becher and Baker 1964; UCC 1976; Affel et al. 1980). The program improved as 

better technology and more information became available. Results from personnel dosimeters were 

used to measure and record doses from external radiation exposure to PGDP workers. These 

dosimeters included one or more of the following:
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• Personnel whole-body beta/photon dosimeters, 

• Pocket ionization chamber dosimeters, and

• Personnel neutron dosimeters. 

For low-energy beta radiation, the dosimeters were probably incapable of furnishing accurate doses in 

terms of Hp(0.07). This TBD analysis does not include extremity doses, which were generally not 

assessed (PACE and University of Utah 2000).

In 1953, PGDP began using Oak Ridge National Laboratory (ORNL) to provide and process 

dosimeters (Baker ca. 1995). There is evidence that PGDP might have processed its own dosimeters 

for a period; a review of the limited documentation available indicated that practices were similar to 

those at ORNL and other major sites at that time (UCNC 1957a). PGDP used the ORNL services 

through the implementation of a newer film badge in 1961 for all Union Carbide Corporation – Nuclear 

Division (UCC-ND) facilities (ORAUT 2009). Beginning in 1980, the TLD system for UCC-ND facilities 

was issued by the Y-12 Plant laboratory (UCND 1980). Beginning in 1989, the Centralized External 

Dosimetry System (CEDS) operated by the Y-12 Plant laboratory was implemented. (ORAUT-TKBS0014-6, ORAUT 2007a). The CEDS dosimetry system is DOELAP-accredited, so reported deep and 

shallow dose values can be treated as representative of Hp(10) and Hp(0.07) for dose reconstruction. 

Table 6-1 summarizes PGDP personnel beta/photon and neutron dosimeter characteristics [dosimeter 

type, exchange, minimum detection level (MDL), and potential missed annual dose]. ORNL, which 

was then the Clinton Laboratory, had based its dosimetry methods on the personnel beta/photon 

dosimeter design from the Metallurgical Laboratory at the University of Chicago (Pardue, Goldstein, 

and Wollan 1944). 

The precise detection levels in Table 6-1 are difficult to estimate, particularly for older systems. 

Current PGDP commercial thermoluminescent dosimeter (TLD) system MDLs are identified in ORNL 

documentation (MMES 1994) based on a DOELAP-accredited laboratory testing protocol (DOE 

1986). During earlier years, MDLs were subject to additional uncertainty because factors such as 

radiation field, film type, and processing, developing, and reading systems cannot now be tested 

(Thornton, Davis, and Gupton 1961). The estimates of film dosimeter MDLs in Table 6-1 were based 

on information from NIOSH (1993), Lalos (1989), Wilson et al. (1990), and site personnel. 

Examination of older records, when available, indicated that the Hp(0.07) MDL values were about 

3 times those for Hp(10) for film. The current TLD MDLs were obtained from ORNL (MMES 1994). 

The film badge was replaced by the TLD in 1980 (UCC 1980). Parameters of the PGDP 

administrative practices significant to dose reconstruction involve policies to:

• Assign dosimeters to workers, 

• Exchange dosimeters, 

• Record notional dose (i.e., some identified value for lower dosed workers, often based on a 

small fraction of the regulatory limit), 

• Estimate dose for missing or damaged dosimeters, 

• Replace destroyed or missing records, 

• Evaluate and record dose for incidents, and

• Obtain and record occupational dose to workers for other employer exposure. 

PGDP policies appear to have been in place for all these parameters. From startup until July 1960, 

PGDP issued dosimeters to a limited number of individuals (PACE and University of Utah 2000). This 
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population of monitored individuals represents those with the highest exposure potential. After July 

1960, routine practice required the assignment of dosimeters to all workers who entered a controlled 

radiation area (BJC 2000). Dosimeters were exchanged on a routine schedule (UCNC 1957a; UCC 

1977; DOE 2000a). For workers in some areas the frequency was monthly, but for the general 

population it was quarterly. Employees on the monthly exchange cycle were primarily involved in 

chemical processing, maintenance of chemical processing facilities, and uranium metal production 

(DOE 2000). All dosimeters were processed, and measured results were recorded and used to 

estimate dose.

Current administrative practices are generally available (MMES 1994), as is detailed information for 

each worker in the PGDP exposure history documentation. Summary documents provide information 

on historical practices at PGDP (PACE and University of Utah 2000; BJC 2000; UCNC 1957a; Affel et 

al. 1980; Baker ca. 1995).
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Table 6-1. Dosimeter types, periods of use, exchange frequencies, MDLs, and potential annual missed doses.

Hp(10) beta/photon dosimeters

Dosimeter Period of use Monitored population

Exchange 

frequency

Laboratory 

MDL(rem)a

Maximum annual 

missed dose 

equivalent (rem)b

Two-element film 1953 through 07/1960 Selected workers based on activities 

performed

Weekly (n = 50) 0.04 1.0

Four-element film After 07/1960 through 

1979

Workers in Buildings C-340, C-400, 

and C-410

Monthly (n = 12) 0.04 0.24

Four-element film After 07/1960 through 

1979

Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.04 0.08 

Four-element film After 07/1960 through 

1979

Workers and visitors not likely to 

exceed 0.1 of applicable guidelines

Annual (n = 1) 0.04 0.02

Harshaw two-chip TLD 1980 through 1988 Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.02 0.04

Harshaw two-chip TLD 1980 through 1988 Workers and visitors not likely to 

exceed 0.1 of applicable guidelines

Annual (n = 1) 0.02 0.01

Harshaw four-chip TLD, 

8800 series

1989 through present Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.02 0.04

Harshaw four-chip TLD, 

8800 series

1989 through present Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Annual (n = 1) 0.02 0.01
 

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Hp(0.07) beta/photon dosimeters

Dosimeter Period of use Monitored population

Exchange 

frequency

Laboratory 

MDL(rem)a

Maximum annual 

missed dose 

equivalent (rem)b

Two-element film 1953 through 7/1960 Selected workers based on activities 

performed

Weekly (n = 50) 0.12 3.0

Four-element film After 7/1960 through 

1979

Workers in Buildings C-340, C-400, 

and C-410

Monthly (n =12) 0.12 0.72

Four-element film After 7/1960 through 

1979

Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.12 0.24

Four-element film After 7/1960 through 

1979

Workers and visitors not likely to 

exceed 0.1 of applicable guidelines

Annual (n = 1) 0.12 0.06

Harshaw two-chip TLD 1980 through 1988 Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.03 0.06

Harshaw two-chip TLD 1980 through 1988 Workers and visitors not likely to 

exceed 0.1 of applicable guidelines

Annual ( n = 1) 0.03 0.015

Harshaw four-chip TLD, 

8800 series

1989 through present Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.02 0.04

Harshaw four-chip TLD, 

8800 series

1989 through present Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Annual ( n = 1) 0.02 0.01

Neutron dosimetersc

Dosimeter Period of use Monitored population

Exchange 

frequency

Laboratory 

MDL(rem)a

Maximum annual 

missed dose 

equivalent (rem)b

Harshaw TLND 1998 through present Selected workers based on activities 

performed

Quarterly (n = 4) 0.015 0.03

a. Estimated film dosimeter detection levels based on NIOSH (1993), Lalos (1989), and Wilson et al. (1990). TLD detection levels from MMES (1994) and personal 

communication with site personnel.

b. Maximum annual missed dose (NIOSH 2007).

c. The potential annual missed dose based on laboratory irradiations is not applicable to workplace missed neutron dose.
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6.3.2 Dosimetry Technology

PGDP dosimetry methods evolved with the development of improved technology and better 

understanding of complex radiation fields. The adequacy of dosimetry methods to measure radiation 

dose accurately is determined from radiation type, energy, exposure geometry, and other factors 

described in this section. The dosimeter exchange frequency gradually lengthened, corresponding in 

general to the period of regulatory dose controls.

6.3.2.1 Beta/Photon Dosimeters

PGDP has historically used personnel dosimetry services from ORNL or the Y-12 Plant laboratory. 

(UCND 1980) In 1945, ORNL implemented the beta/gamma film dosimeter design, which was 

developed originally at the Metallurgical Laboratory at the University of Chicago (Pardue, Goldstein, 

and Wollan 1944). ORNL followed a research and development process that led to gradual upgrades 

in dosimetry capabilities for complex radiation fields (Thornton, Davis, and Gupton 1961). Other DOE 

sites followed this evolution in dosimetry capabilities, which led to site-specific multielement film and 

TLD systems.

Figure 6-1 shows the energy response characteristics of the PGDP beta/gamma dosimeters based on 

the essentially identical two-element film dosimeter that was designed at the University of Chicago 

and used at the Hanford Site (as well as ORNL, Los Alamos National Laboratory, and probably other 

MED sites). 

Figure 6-1. Estimated dosimeter photon response characteristics (Wilson, et al. 1990). 

In addition, Figure 6-1 shows the Hp(10) response. Further, the figure shows the energy response of 

Hanford multielement film and TLDs (Wilson et al. 1990). The curve labeled “Two-Element Film 

Shield” represents ORNL dosimeters from 1945 through 1978. ORNL used a multielement film 

dosimeter after 1953 (Thornton, Davis, and Gupton 1961), but processed photon response as it did 

for the two-element dosimeter and used the same shielding as that in the two-element dosimeter. 

The figure shows that the two-element dosimeter overresponded in relation to Hp(10) from 0.05 to 

0.3 MeV, followed Hp(10) for higher energies, and underresponded for lower energies. It also shows 
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that TLDs are capable of following Hp(10) over the energy range of interest. The majority of PGDP 

worker photon dose has come from handling uranium of low enrichment. Therefore, the photon 

energy spectrum has been almost entirely in the range from 30 to 250 keV (Shleien, Slaback, and 

Birky 1998).

The nonpenetrating response of the two-element dosimeter was calculated as the difference between 

the open window (OW) and shielded (S) portions of the film based on a uranium calibration. The twoelement dosimeter workplace nonpenetrating (i.e., beta or shallow) dose response based on the 

uranium calibration should adequately represent Hp(0.07) or at least be favorable to claimants 

because of the significant overresponse of the OW portion of the film to lower energy photons that 

could have been present (Wilson et al. 1990). The multielement film dosimeters and TLDs, which 

were also calibrated to uranium slabs, had the ability to correct more accurately for mixed photon and 

beta radiation (Wilson et al. 1990).

6.3.2.2 Neutron Dosimeters

Dosimeters at PGDP historically had a neutron-sensitive element that was processed on request. 

From 1989 to 1998, this capability was provided by a TLD that contained a 6

LiF chip, which is very 

responsive to low-energy neutrons. There is no indication of recorded neutron doses for PGDP 

workers wearing either of these dosimeters [1]. The use of commercial Harshaw thermoluminescent 

neutron dosimeters (TLNDs) to assess neutron dose routinely (along with deep and shallow dose) 

began in 1998. ORNL has provided the dosimeters and associated services. The albedo dosimeter 

has been worn with a belt to minimize distance from the worker’s body, which optimizes the albedo 

effect for which the dosimeter is calibrated [2]. Before 1998, the beta/photon badge assembly 

contained a neutron-sensitive element (NTA; Eastman Kodak Type 2 film) with an energy threshold of 

about 0.5 MeV. This element was processed only when requested. A review of data does not 

indicate the assignment of neutron dose before 1998.

The quality factors (QFs) for neutrons have changed significantly over time. In current regulations, 

QFs that are used to convert radiation dose (millirad) to dose equivalent (millirem) are based on 

International Commission on Radiological Protection (ICRP) Publication 38 (ICRP 1983). The most 

current QFs from ICRP Publication 60 (1991) are about 2 times higher than the ICRP Publication 38 

(ICRP 1983) values: therefore, an adjustment of a factor of 2 is necessary. 

Average neutron energy has been less than about 1 MeV (510 keV for 2% 235U, 770 keV for 5% 235U, 

and 860 keV for 97% 235U) (Cardarelli 1997, p. 9). QF equals 10 for ICRP (1983), or about 20 for the 

ICRP (1991) revision. The average neutron energy from depleted and natural uranium cylinders has 

ranged from 210 to 360 keV (Cardarelli 1997, p. 9). Process-created unmoderated and deuterium 

(water) 252Cf neutrons were between 1,306 and 1,403 keV. Therefore, assuming that 100% of the 

neutron doses at PGDP were delivered by neutrons in the 0.1-to-2-MeV energy range is favorable to 

claimants. 

6.3.3 Calibration

Potential error in recorded dose is dependent on: 

• Dosimetry technology response characteristics to each radiation type, energy, and geometry;

• Calibration methodology; and

• Extent of similarity between the radiation fields for calibration and those in the workplace.
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The potential error is much greater for dosimeters with significant variations in response, such as film 

dosimeters for low-energy photon radiation and the nuclear track emulsion (NTA) and TLNDs for 

neutron radiation [4].

6.3.3.1 Beta/Photon Dosimeters

The beta/photon film dosimeters at PGDP were calibrated to 226Ra until 1980 when the calibration 

source changed to 137Cs (ORAUT 2007a). The calibration to both 226Ra and 137Cs was free in air (no 

phantom) until the DOELAP procedures adopted in 1986 required phantoms (DOE 1986). Hp(10) is 

defined with a phantom, in particular the ICRU slab phantom, which is a conservative practical 

definition of anterior-posterior whole-body dose to the standard ICRU spherical phantom (ICRU 1998).

Introduction of on-phantom calibration of film dosimeters, and the replacement of 226Ra by 137Cs as 

the calibration source, changed the relationship between recorded dose and Hp(10). In addition to 

registration of the additional backscattered radiation, the generally lower energy photon spectrum 

from 226Ra in comparison with that of 137Cs (662 keV) gave a greater optical density for the same dose 

during calibration (Figure 6-1). In contrast, the effect of backscatter is to overestimate dose, and 

calibration with 226Ra tends to underestimate the dose in relation to calibration with 137Cs. 

In the 1980s, studies at a number of laboratories assessed changes from the on-phantom calibration 

mandated by the DOELAP testing criteria (Fix et al. 1982; Wilson 1987; Wilson et al. 1990; Taylor et 

al. 1995). While not exactly the same at all sites, most film dosimeters, like those at PGDP, had 

common features due to their evolution from the original work of Pardue, Goldstein, and Wollan 

(1944). The early badges were calibrated to exposure in free air. Laboratory tests at the Hanford Site 

showed 8% and 4% increases in dosimeter response for on-phantom exposures using 226Ra and 

137Cs, respectively (Fix et al. 1982). With free-air calibration, the exposure to the wearer tends to be 

overestimated by this amount, which is assumed to be similar for Paducah. Tests at the Savannah 

River Site, on the other hand, indicated that film badge doses underestimated Hp(10) by 11.9% before 

1986 and by 3.9% in 1986 (Taylor et al. 1995). Lacking site-specific data for PGDP, this TBD 

recommends the use of exposure-to-organ dose conversion factors (DCFs) in Appendix B of OCASIG-001 (NIOSH 2007) for dose reconstruction at PGDP with no numerical adjustment to the recorded 

doses; this procedure should be favorable to claimants (Fix et al. 1982). It allows for an overestimate 

of exposure, as assessed in the Hanford studies, that should be sufficient to offset effects due to the 

calibration source if they are in the opposite direction.

For a number of years, ORNL used uranium beta as well as 226Ra gamma calibration curves to 

interpret film densities (Thornton, Davis, and Gupton 1961). The ratio of beta-to-gamma responses 

was tested in several ways. Films wrapped in a 7-mg/cm2 absorber were placed in contact with a slab 

of natural uranium. The densities per rad were nearly the same as those produced from 226Ra gamma 

rays measured behind a cadmium filter. In addition, stacks of film were exposed on a uranium 

surface, and the densities at various depths were used to extrapolate to the value for a depth of 

7 mg/cm2

. This value was nearly equal to that produced by the same dose from 226Ra photons behind 

the cadmium filter. Therefore, for beta radiation from natural uranium, the density produced per rad in 

film was equal to the density produced per rad behind the cadmium filter by 226Ra gamma rays. 

Analysts concluded that, for routine personnel dosimetry, film was equally sensitive to beta and 

gamma radiations. Because the film badge had a minimum absorber thickness of 80 mg/cm2

between the film and the source, the effective beta energy is necessary to interpret the film density in 

terms of Hp(0.07). The radiation was routinely treated as 1.7-MeV beta particles from uranium, which 

are about 40% absorbed in 80 mg/cm2 (Thornton, Davis, and Gupton 1961). Therefore, this 

determination of beta dose was specific to uranium.
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6.3.3.2 Neutron Dosimeters

Calibration of neutron dosimeters for PGDP was appropriate for the work locations (MMES 1994). 

Dosimeter response was characterized in a manner that would represent the workplace (MMES

1994). Reference dosimetry for these measurements was evaluated with tissue-equivalent 

proportional counters (TEPCs). TEPCs provide an absolute measure of absorbed dose in a tissuelike material and, with an appropriate algorithm, an estimate of the neutron QF (Scherpelz and 

Murphy 1995). The basis for the calibration factor was developed using data from a room at the Y-12 

Plant that was used to store an array of small canisters of UF4. Measurements were made with 

Bonner spheres at the same location. The average QF was 11, and the average energy range was 

0.6 to 1.4 MeV (Soldat et al. 1990).

In 1989, field measurements for neutron flux were made by Pacific Northwest National Laboratory 

representatives at the end row of the cylinder yard at the K-25 Plant. The measurements were 

completed with a TEPC and a phantom with TLDs approximately 4 ft from the outside of a cylinder; 

the phantom was near the center of the cylinder’s length. The results were evaluated qualitatively 

because the dose rate was low and an appropriate power supply was not available. The calibration 

factors were similar to those in Y-12 Building 9212 in the UF4 storage area container array and 

confirmed the appropriateness of these values (Soldat et al. 1990). These calibration factors apply to 

the PGDP TLNDs (MMES 1994). 

6.3.4 Workplace Radiation Fields

6.3.4.1 Beta/Photon Fields

PGDP operations have been characterized by the relatively low-level external beta and photon 

radiation fields associated with uranium in feed materials, products, wastes, and contaminated 

equipment and systems. Processed RU was present with natural, depleted, and enriched (up to 2% 

235U by weight) abundances. (Section 6.3.4.3 describes potential sources for neutron exposure.)

Table 6-2 summarizes the major sources of external radiation throughout PGDP operations (PACE 

and University of Utah 2000). The photon energy range of principal interest is 30 to 250 keV. 

Handling uranium material of these types did not, in general, produce areas with significantly elevated 

photon radiation. 

The major facilities and associated activities at PGDP have been (BJC 2000):

• C-331, C-333, C-335, and C-337 – gaseous diffusion process buildings, 

• C-410/420 – UF6 Feed Plant, 

• C-310 – Purge and Product Withdrawal Building, 

• C-315 – Surge and Tails Withdrawal Building, 

• C-340 – Metals Plant, 

• C-400 – Decontamination and Cleaning Building, and

• C-720 – Maintenance Building. 

The buildings with the greatest potential for elevated direct radiation levels were C-340, C-410, C-420, 

and the cascade buildings (PACE and University of Utah 2000). From 1952 to approximately 1980, 

the major sites of potential exposure to radioactive material were buildings involved in the conversion 

of UO3 powder to enriched UF6 in solid or gaseous form, UF4 and uranium metals recovery operations, 

and the decontamination building. Feed and enrichment operations were in Buildings C-410, C-420, 

C-331, C-333, C-335, C-337, C-310, and C-315, while UF4 and uranium recovery were in Building 

C-340 (UCNC 1957b). The decontamination operation was in Building C-400. The oxide conversion 

building, C-420, was where UO3 powder (clean or recycled) was received and converted to UF4. 
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From Building C-420, material went to Building C-410, the feed plant, for conversion to UF6. Last, UF6

was processed through the cascade buildings (C-331, C-333, C-335, and C-337). Enriched UF6 was 

withdrawn in Building C-310, the product withdrawal building, while depleted UF6 was removed in 

Building C-315, the tails withdrawal building. Radiation surveys were performed near the UF6

cylinders to evaluate the potential for exposure to personnel working adjacent to the shipping 

containers and area exposure rates in the cylinder yards (McDougal 1980; Frazee 1982; Mason 

1986). Table 6-3 lists the principal buildings, sources for external dose, and periods of operation.

Table 6-2. Major radiation sources, half-lives, energies (MeV), and abundances (%). 

Nuclide Source Half-life Alpha

Beta 

(maximum) Gamma

U-238 Primary U isotope 4.51E+09 yr 4.15 (21%), 

4.20 (79%)

Not applicable Not applicable

U-235 Primary U isotope 7.1E+08 yr 4.21 (6%), 

4.37 (17%), 

4.40 (55%), 

4.60 (5%)

Not applicable 0.144 (11%), 

0.163 (5%), 

0.186 (57%), 

0.205 (5%)

U-234 Primary U isotope 2.47E+05 yr 4.72 (28%), 

4.77 (72%)

Not applicable 0.053 (0.12%)

Th-234 Decay product 24.1 d Not applicable 0.103 (21%), 

0.193 (79%)

0.013 (9.8%), 

0.063 (3.5%), 

0.092 (3%), 

0.093 (4%)

Pa-234m Decay product 1.17 min Not applicable 2.29 (98%) 0.765 (0.3%), 

1.001 (0.60%)

Th-231 Decay product 25.5 hr Not applicable 0.206 (13%), 

0.287 (12%), 

0.288 (37%), 

0.305 (35%)

0.026 (2%), 

0.084 (10%)

Tc-99 Impurities from RU 2.12E+05 yr Not applicable 0.294 (100%) 0

Table 6-3. Buildings and periods of operation. 

Site facilities Source for external dose Period

C-310 Purge and Product 

Withdrawal

UF6 process equipment and cylinders 1953–present

C-315 Surge and Tails Withdrawal UF6 process equipment and cylinders 1953–present

C-331, C-333 Gaseous Diffusion 

Process Buildings

UF6 process equipment and cylinders 1952–1964

1969–1970

1972–1976

C-335, C-337 Gaseous Diffusion 

Process Buildings

UF6 process equipment and cylinders 1954–1964

1969–1970

1972–1976

C-340 Reduction and Metals 

Facility

Process equipment, contaminated floors 1957–1962

1967–1977

C-400 Decontamination and 

Cleaning Buildings

UF6 process equipment and cylinders 1952–1990

C-410 UF6 Feed Plant and C-420 

Oxide Conversion Plant

Process equipment, contaminated floors 1953–1964

1968–1977

C-415 Feed Plant Storage Building Radioactive source storage area 1953–1977

C-745 A-V Cylinder Yards UF6 cylinders 1953 (estimated)–present

PGDP also processed RU. The feed material contained trace amounts of radioactive impurities not 

present in natural uranium feed material. Because these impurities were present in such minute 

concentrations, their radiological impact was usually negligible (PACE and University of Utah 2000). 
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However, some routine chemical processes would concentrate them (PACE and University of Utah 

2000). From an external dose standpoint, the most significant impurity in RU is the pure beta emitter 

99Tc, which tends to deposit in enrichment equipment and “pocket” in the higher sections of the 

diffusion cascade (DOE 2001). Technetium-99 was also concentrated for recovery and removal. The 

relatively low-energy beta particles (maximum 294 keV) from 99Tc pose minimal external exposure 

potential because of their limited range. Neither film badges nor TLDs efficiently detect them, 

particularly in the presence of uranium. Clothing and gloves provide adequate shielding. Skin 

contamination is the only credible scenario in which significant shallow dose could occur from 99Tc. 

6.3.4.2 Workplace Beta/Photon Dosimeter Response

PGDP has historically used personnel dosimeter services from ORNL or the Y-12 Plant laboratory. 

(UCND 1980) Essentially all radiological work areas involved photon and beta radiation characteristic 

of operations with low-enrichment uranium. 

All personnel dosimetry PGDP (from ORNL) used over its history included suitably filtered elements 

for the determination of deep dose. The ORNL film dosimeters, which were calibrated in terms of 

exposure from 226Ra photons, show either good agreement with or an overresponse to Hp(10) (see 

Figure 6-1). However, because dosimeters were calibrated in terms of exposure over most of ORNL 

history, dose reconstructors should apply exposure to organ DCFs to the recorded deep dose results 

for the period through 1988 (i.e., for the period before DOELAP accreditation). From 1989 to the 

present, reported deep dose values are considered equivalent to Hp(10) (ORAUT 2007a, 2007b) 

6.3.4.3 Neutron Fields

While neutron radiation has occurred in some areas at PGDP, measured levels have been low. There 

are no identified locations where measurable neutron dose was encountered (MMES 1994). This is 

confirmed by studies of neutron fields at gaseous diffusion plants (Scherpelz and Murphy 1995; 

Cardarelli 1997). Cylinder yards, feed and withdrawal areas, and locations where uranium forms 

deposits in the cascade have been investigated (Cardarelli 1997). These studies identified the 

storage cylinders, which contained either depleted UF6 (tails) or enriched UF6 (product), as areas 

where neutron fields could represent an exposure hazard. Estimates of dose equivalent rates ranged 

from 0.007 to 0.34 mrem/hr with QFs from 7 to 10. Radiation measurements indicated that the 

neutron flux increased as a function of uranium enrichment; neutron flux increased from 0.2 mrem/hr 

for cylinders with as much as 5% enrichment to 4 mrem/hr on contact with 97% enrichment (DOE 

2001). A representative average value is 0.2 mrem/hr based on a QF of about 10 (Scherpelz and 

Murphy 1995; Cardarelli 1997). Estimates of average neutron energies ranged from 0.25 to 0.56 MeV 

(Scherpelz and Murphy 1995). Neutron monitoring of individuals during a UF6 cylinder-painting 

project (BJC 1999) indicated a neutron-to-photon dose equivalent ratio of approximately 1 to 5 based 

on a QF of 10 [5].

Cylinders of highly enriched (93% to 96%) uranium (HEU) were measured with a TEPC on a phantom 

about 24 in. from the cylinders (Soldat and Tanner 1992). The dose equivalent from the cylinders was 

about 0.8 mrem/hr with a total dose equivalent of 14 mrem. The multisphere measurement at the 

same location as the phantom resulted in an average neutron energy of 0.53 MeV and a dose 

equivalent rate of 0.5 mrem/hr.

The solid lines in Figure 6-2 show the calculated energy spectrum from the multisphere detectors 

(Bonner spheres). Table 6-4 lists dose fractions for the neutron energy groups (indicated by the 

dashed lines in Figure 6-2). The dose fractions for the lower (less-than-10-keV) and intermediate 

(10- to 100-keV) energy neutron groups were about 47% of the total dose from the measurements 

(ORAUT 2007a).
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Figure 6-2. Results of neutron spectrum measurements about 24 in. in front of 93%–96% HEU 

cylinders (Soldat and Tanner 1992).

Table 6-4. Dose fractions for PORTS HEU storage vault in Building 

345 near an unshielded 252Cf source. 

Neutron

energy group

Actual 

fractions

Fractions 

favorable to claimants

<10 keV 0.300 0.300

10–100 keV 0.172 0.610

0.1–2 MeV 0.447 0.610

2–20 MeV 0.081 0.081

Exposure to low-enriched UF6 (less than 5%) results in a lower neutron flux than that from highly 

enriched UF6 (greater than 97%) as surveyed at the Portsmouth Gaseous Diffusion Plant (PORTS) by 

Soldat and Tanner (1992). The dose fractions in Table 6-4 are favorable to claimants (Soldat and 

Tanner 1992).

A neutron study in 1990 at ORNL and the Y-12 Plant was the only definitive study of neutron energy 

spectra documented over the history of PGDP (Soldat et al. 1990). The energy spectra are assumed 

valid for the earlier years due to the presence of enriched uranium.

6.3.4.4 Workplace Neutron Dosimeter Response

Quantitative monitoring for neutron dose began at PGDP in 1998. TLNDs were used in conjunction 

with appropriate work field calibration factors. Before 1998, the beta/photon badge assembly 

contained a neutron-sensitive element (NTA; Eastman Kodak Type 2 film) with an energy threshold of 
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about 0.5 MeV. This element was processed only when requested. A review of data does not 

indicate the assignment of neutron dose before 1998.

6.4 ADJUSTMENTS TO RECORDED DOSE

6.4.1 Photon Dose

Recorded doses varied in reporting units depending on regulatory requirements and dose definitions 

(national and international). The DOE reporting unit is the millirem, a unit of dose equivalent. The 

international unit of dose equivalent is the millisievert, which is equivalent to 100 mrem. Dose 

reconstructors should apply exposure to organ DCFs to deep dose results for PGDP workers for the 

period through 1988. From 1989 to the present, reported deep dose values are considered equivalent 

to Hp(10) (ORAUT 2007b). 

6.4.2 Nonpenetrating Dose

The early film dosimeters were calibrated to uranium for nonpenetrating radiation. No numerical 

adjustment of recorded shallow doses is recommended. Incident reports are a possible source that 

dose reconstructors can consult for investigations of nonroutine beta exposures and dose 

assessment. 

6.4.3 Neutron Dose

Measured neutron energies at PGDP are between 0.1 and 2.0 MeV, for which the ICRP Publication 

60 radiation weighting factor is 20 (ICRP 1991). Therefore, dose reconstructors should multiply the 

reported neutron dose equivalent by the appropriate ICRP (1991) correction factor to be used for 

reconstruction (NIOSH 2007). Apply this factor to measured, missed, and unmonitored neutron 

doses.

6.5 MISSED DOSE

Missed deep and shallow doses have been examined for three groups of PGDP workers as follows:

• Group 1. A zero dose was recorded but the worker was not monitored (most workers from 

1953 to July 1960).

• Group 2. A zero dose was recorded for the dosimeter system for any response less than the 

MDL.

• Group 3. There was no recorded dose because workers were not monitored or the dosimetry 

record is not available. 

Neutron dose rates at PGDP were low (MMES 1994). Neutron dosimeters were not routinely 

assigned and doses were not recorded until about 1998. Neutron doses from before 1998 were 

based on a conservative calibration associated with a neutron-sensitive element in the beta/gamma 

dosimeter. Application of a neutron-to-gamma dose equivalent ratio of 1 to 5 appears to be a 

satisfactory option that is favorable to claimants because the photon dose is reliably measured. This 

ratio can be applied to selected work activities [6].
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6.5.1 Estimating Missed and Unmonitored Photon Deep Dose

6.5.1.1 Non-Construction Trade Workers

Watson et al. (1994) examined methods to be considered when there is no recorded dose for a period 

during a working career (Group 1). In general, estimates of unmonitored dose can be made by using 

dose results for coworkers or the recorded dose before and after the period when they were not 

monitored. However, these situations require careful examination. The dose reconstructor should 

consider all reasonable methods and assign the most appropriate dose based on employee job 

descriptions and work locations. NIOSH (2007) cites several different models.

For Group 2, the missed dose for dosimeter results less than the MDL is particularly important for 

earlier years when MDLs were higher and dosimeter exchange was more frequent. NIOSH (2007) 

describes an acceptable estimate that is favorable to claimants of the maximum potential missed dose 

as one-half the MDL (or limit of detection, LOD) multiplied by the number of zero dose results. The 

right-hand column in Table 6-1 lists estimates of the annual missed dose for Group 2 at PGDP.

If it is definite that the employee was not a radiation worker, the unmonitored deep dose for that 

period can be assigned as the onsite ambient dose.

Otherwise, dose reconstructors should treat an individual in Group 1 or 3 as a radiation worker, then 

approach the unmonitored deep dose in two ways. First, consider the same assignment of missed 

dose as that for Group 2 (from the right-hand column of Table 6-1). However, for 1953 through July 

1960, with the frequent (weekly) dosimeter exchange and relatively large MDL, the resulting implied 

annual missed dose of 1 rem is probably unrealistically large for many unmonitored workers in Groups 

1 and 3. Figure 6-3 shows the distribution of individual annual deep dose equivalent for monitored 

workers from 1953 to 1974 (Baker ca. 1995). Few of these individuals received as much as 1 rem in 

a year.

Figure 6-3. Historical distribution of deep dose equivalent (Baker ca. 1995). 

An alternative approach for Group 1 or 3 is to base the unmonitored dose estimate on exposure data 

compiled for monitored PGDP workers. ORAUT-OTIB-0020, Use of Coworker Dosimetry Data for 

External Dose Assignment (ORAUT 2011), provides general instructions to evaluate the measured 

and missed doses for monitored PGDP workers to arrive at a dose that is favorable to claimants to be 
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assigned to unmonitored workers. Attachment B contains the details of the evaluation of PGDP 

coworker dose to be assigned to unmonitored workers. These measured doses include an analysis of

the missed dose, which is particularly significant for the earlier years with higher LODs and frequent 

dosimeter exchanges. Table B-2 provides the 50th- and 95th-percentile coworker doses.

6.5.1.2 Construction Trade Workers

Measured doses to construction trade workers (CTWs) have been increased to account for 

uncertainty for reasons described in ORAUT-OTIB-0052, Technical Information Bulletin: Parameters 

to Consider When Processing Claims for Construction Trade Workers (ORAUT 2014). For extended 

employment periods without a measured dose, consideration about whether to assign an unmonitored 

dose using the coworker doses in Attachment B is necessary. In this case, the measured coworker 

penetrating annual dose has been multiplied by a factor of 1.4 (ORAUT 2014), and the missed dose 

was determined using NIOSH (2007) guidance. Table B-3 lists the 50th- and 95th-percentile doses 

for CTWs. 

6.5.2 Estimating Missed and Unmonitored Shallow Dose

The procedure for assessing missed and unmonitored shallow dose is similar to that for missed deep 

dose. 

For Group 2, the last column of Table 6-1 lists the missed annual shallow dose equivalent in keeping 

with the MDL/2 method of evaluation. Attachment B contains guidance on determining the 

reconstructed skin dose. Figure 6-4 shows the historical data for the distribution of shallow dose 

equivalent for monitored workers (Baker ca. 1995). In comparison with Figure 6-4, the Table 6-1 

values for annual missed shallow dose for Group 2 is favorable to claimants.

For nonradiological workers in Groups 1 and 3, the unmonitored shallow dose can be assigned as the 

environmental dose. Dose reconstructors should regard other individuals in these groups as radiation 

workers and consider the same estimate as that for Group 2. As an alternative, use Attachment B, 

Table B-2. Significant nonroutine beta doses, such as from skin contamination events (particularly 

during 99Tc recovery and removal), could be addressed in specific incidence reports. Attachment A 

provides guidance.
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Figure 6-4. Historical distribution of shallow dose equivalent (Baker ca. 1995).

6.5.3 Estimating Missed and Unmonitored Neutron Dose

Dose reconstructors should add a neutron component to annual doses prior to 1998. Table 6-5 lists 

the criteria for assigning missed and unmonitored neutron dose. The affected radiological areas 

include, but are not limited to, cascade facilities, feed and production withdrawal areas, oxide 

conversion facilities, feed manufacturing facilities, decontamination or cleaning facilities, cylinder 

yards, and neutron source storage areas. 

Table 6-5. Missed and unmonitored neutron dose assignment (BJC 1999). 

Description Neutron dose assignment

Unmonitored employee – all radiological areas Assign neutron dose based on coworker dose. Apply a 

neutron-to-photon ratio of 0.2 for dose equivalent.

Monitored (photon and electron) employee Assign neutron dose if positive photon results are 

recorded. Apply a neutron-to-photon ratio of 0.2 for 

dose equivalent.

Monitored (photon and electron) employee having 

no dosimetry results greater than half the LOD

Assign neutron dose based on the missed dose. Apply 

a neutron-to-photon ratio of 0.2 for dose equivalent.

Beginning in 1998, employees with neutron exposure were monitored, and these dosimetry records 

should be used for dose assignment. The neutron dose equivalent should be multiplied by a factor of 

2 as previously discussed in Section 6.4.3.

6.6 UNCERTAINTY

PGDP has historically used ORNL personnel dosimeter services. ORNL has assessed the standard 

error in the recorded film-badge dose as ±30% for photons of all energies (ORAUT 2007b). The 

standard error for beta dose is the same (or somewhat larger for unknown mixtures of beta/gamma 

dose). Therefore, the film badge dose uncertainty is 1.3. The uncertainty in the TLD dose is 1.15 

(ORAUT 2007b), which is consistent with NIOSH (2007).
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6.7 DOSE RECONSTRUCTION

As much as possible, dose reconstructors should base dose to individuals on dosimetry records. It is 

important to distinguish between the recorded nonpenetrating and penetrating doses and the actual 

Hp(0.07) and Hp(10). The following list summarizes appropriate actions for dose reconstructors:

• Consider dosimetry records that provide nonzero beta/photon values for Hp(10) and Hp(0.07)

to be adequate. No numerical adjustment of the doses is required. Beta energies are greater 

than 15 keV and photon energies range from 30 to 250 keV.

• Assign missed dose to workers for whom dosimetry records provide zero beta/photon values 

for Hp(10) and Hp(0.07) based on MDL/2 times the number of zero results, as described in 

Sections 6.5.1 and 6.5.2 (NIOSH 2007).

• Assign only the missed environmental dose from the environmental TBD (ORAUT 2012b) if it 

is certain that the individual was not a radiation worker. However, individuals with no recorded 

dose might or might not have been radiological workers, so if there is uncertainty, estimate the

missed dose as described in Section 6.5. No numerical adjustments to the missed dose are 

necessary.

• Multiply reported, missed, and unmonitored neutron dose equivalents by the appropriate ICRP 

(1991) correction factors. 

• Base the assignment of missed and unmonitored neutron dose equivalent on a neutron-tophoton ratio of 0.2 for dose equivalent (BJC 1999) for years before 1998. Beginning in 1998, 

base the neutron assignment on the dosimetry records. Assign missed neutron dose using 

Table 6-1. Multiply the estimated neutron dose equivalent by 2 to adjust for ICRP (1991).

• Pay special attention to the possibility of skin contamination incidents for workers involved with 

99Tc recovery operations (Attachment A).

• See Section 6.6 for a discussion of uncertainty.

6.8 ORGAN DOSE

NIOSH (2007) discusses the conversion of measured doses to organ dose equivalent, and 

Appendix B of that document contains the appropriate DCFs for each organ, radiation type, and 

energy range based on the type of monitoring performed. In some cases, simplifying assumptions are 

appropriate [7].

6.9 ATTRIBUTIONS AND ANNOTATIONS

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

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

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

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

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

for review on the Project’s Site Research Database (SRDB).

[1] Turner, James E. Oak Ridge Associated Universities (ORAU) Team. Consultant. 2003. 

The reviewed records did not reveal recorded neutron doses for either dosimeter.
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[2] Turner, James E. ORAU Team. Consultant. 2003. 

Proximity of the albedo dosimeter is important for its response; standard practice ensured this.

[3] Turner, James E. ORAU Team. Consultant. 2003. 

ICRP (1991) recommends a weighting factor of 20 for neutron energies between 0.1 and 

2 MeV. Doses of record used a QF of 10; therefore, a factor of 2 correction is indicated.

[4] Turner, James E. ORAU Team. Consultant. 2003. 

The importance of energy response to accurate measurement of dose equivalent is well 

known, and the response of the historical dosimeters is shown in Figure 6-1.

[5] Turner, James E. ORAU Team. Consultant. 2003. 

The determination of a neutron-to-photon ratio for absorbed dose was based on a dose 

equivalent ratio that can be used to estimate neutron dose from photon measurements. 

[6] Turner, James E. ORAU Team. Consultant. 2003. 

Empirical neutron and photon worker dose equivalent data provide a basis from which neutron 

dose equivalent can be inferred from a better known photon dose (based on interpretation of 

BJC 1999). 

[7] Turner, James E. ORAU Team. Consultant. 2003. 

Appendix B of NIOSH (2007) contains tables for numerous organs. Some professional 

judgment is needed to fit particular conditions.
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GLOSSARY

absorbed dose

Amount of energy (ergs or joules) deposited in a substance by ionizing radiation per unit mass 

(grams or kilograms) of the substance and measured in units of rads or grays. See dose.

albedo dosimeter

Thermoluminescent dosimeter that measures the thermal, intermediate, and fast neutrons 

scattered and moderated by the body or a phantom from an incident fast neutron flux.

albedo effect

In relation to health physics, dosimeter response caused by the moderating and 

backscattering of neutron radiation by a human chest or a phantom.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

attenuation

Process by which absorption and scattering reduces the number of particles or photons 

passing through a body of matter.

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

Bonner sphere

See multi-sphere neutron spectrometer.

cascade

At PGDP, series of compressor, heat exchanger, control valve and motor, converter stages, 

and supporting piping arranged in stages, cells, and units that progressively increase the 

concentration of 235U in a uranium hexafluoride (UF6) feed. Enrichment occurs as UF6 passes 

through semiporous barriers in the converter stage. These barriers allow the lighter 235U 

molecules to pass through more easily, which results in a gas with a slightly higher percentage 

of 235U (enriched) on one side of the barrier and a slightly lower percentage (depleted) on the 

other side. The enriched UF6 gas flows toward the top of the cascade while the depleted UF6

gas travels toward the bottom of the cascade.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.
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deep dose equivalent [DDE, Hd, Hp(10))]

Dose equivalent in units of rem or sievert for a 1-centimeter depth in tissue (1,000 milligrams 

per square centimeter). See dose.

DOE Laboratory Accreditation Program (DOELAP)

Program for accreditation by DOE of DOE site personnel dosimetry and radiobioassay 

programs based on performance testing and the evaluation of associated quality assurance, 

records, and calibration programs.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rad, rep, or grays.

dose conversion factor (DCF)

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

interest (e.g., liver or colon). In relation to radiography, ratio of dose equivalent in tissue or 

organ to entrance kerma in air at the surface of the person being radiographed.

dose equivalent (DE, H))

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect

from the absorbed dose. See dose.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See albedo dosimeter, film dosimeter, neutron film dosimeter, 

pocket ionization chamber, thermoluminescent dosimeter, and track-etch dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

enrichment

Isotopic separation process that increases the percentage of a radionuclide in a given amount 

of material above natural levels. For uranium, enrichment increases the amount of 235U in 

relation to 238U. Along with the enriched uranium, this process results in uranium depleted in 

235U. At PGDP this involves a process that occurs as UF6 passes through barriers in 

converters that allow isotopes of lower molecular weight to pass through.

external dose

Dose received from radiation emitted by sources outside the body.

film

(1) In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. (2) X-ray film.
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film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

gaseous diffusion plant

Facility where uranium hexafluoride (UF6) gas is filtered to enrich the 235U and separate it from 

238U. The process requires enormous amounts of electric power and results in an increase in 

235U enrichment from 1% to about 3%.

gray (Gy)

International System unit of absorbed radiation dose, which is the amount of energy from any 

type of ionizing radiation deposited in any medium; 1 Gy equals 1 joule per kilogram or 

100 rads.

highly enriched uranium (HEU)

Uranium enriched to at least 20% 235U for use as fissile material in nuclear weapons 

components and some reactor fuels. Also called high-enriched uranium.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray 

radiation.

limit of detection (LOD)

Minimum level at which a particular device can detect and quantify exposure or radiation. Also 

called lower limit of detection and detection limit or level. See minimum detectable level.

minimum detectable activity or amount (MDA)

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

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

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

(Type I error). See action level, decision level, and minimum reporting level.

minimum detectable level (MDL)

See minimum detectable activity.

multi-sphere neutron spectrometer

Spectrometer that consists of a series of neutron-moderating spheres of tissue-equivalent 

material with a neutron detector in the middle of the respective spheres. Algorithms are used 

to calculate the neutron spectra.
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neutron (n)

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen. See element.

neutron film dosimeter

Film dosimeter with a nuclear track emulsion, type A, film packet.

neutron radiation

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron.

nonpenetrating dose (NP, NPEN)

Dose from beta and lower energy photon (X-ray and gamma) radiation that does not penetrate 

the skin. It is often determined from the open window dose minus the shielded window dose. 

See dose.

nuclear track emulsion, type A (NTA)

Film made by the Eastman Kodak Company that is sensitive to fast neutrons. The developed 

image has tracks caused by neutrons that become visible under oil immersion with about 

1,000-power magnification. The number of tracks in a given area is a measure of the dose 

from that radiation.

occupational dose

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

on-phantom

Exposure of a dosimeter on a phantom to simulate the dosimeter’s response when worn on a 

person.

open window (OW)

Area of a film dosimeter that has little to no radiation shielding (e.g., only a holder and visible 

light protection). The open window measures nonpenetrating as well as penetrating dose, 

which minimizes the potential for beta radiation to contribute to the interpreted penetrating 

dose. See film dosimeter.

penetrating dose (PEN)

Dose from moderate to higher energy photons and neutrons that penetrates the outer layers of 

the skin. See dose.

personal dose equivalent [Hp(d)]

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeters 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. These are 

noted as Hp(0.07) and Hp(10), respectively. In 1993 the International Commission on 

Radiological Measurement and Units recommended Hp(d) as the dose quantity for radiological 

protection.
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phantom

Any structure that contains one or more tissue substitutes (any material that simulates a body 

of tissue in its interaction with ionizing radiation) and is used to simulate radiation interactions 

in the human body. Phantoms are primarily used in the calibration of in vivo counters and 

dosimeters. See slab phantom.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.

pocket ionization chamber (PIC)

Cylindrical monitoring device commonly clipped to the outer clothing of an individual to 

measure ionizing radiation. A PIC may be self-reading or require the use of an outside device 

to be able to read the dosimeter. Also called pencil, pocket pencil, pencil dosimeter, and 

pocket dosimeter.

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Program Act of 2000, the percent likelihood, at the 99th percentile, that a 

worker incurred a particular cancer from occupational exposure to radiation.

proton

Basic nuclear particle with a positive electrical charge and mass slightly less than that of a 

neutron. There are protons in the nuclei of every atom, and the number of protons is the 

atomic number, which determines the chemical element. See element.

quality factor (Q, QF)

Principal modifying factor (which depends on the collision stopping power for charged 

particles) that is employed to derive dose equivalent from absorbed dose. The quality factor 

multiplied by the absorbed dose yields the dose equivalent. See dose, relative biological 

effectiveness, and weighting factor.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. See radionuclide.
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recycled uranium (RU)

Uranium first irradiated in a reactor, then recovered through chemical separation and 

purification. RU contains minor amounts of transuranic material (e.g., plutonium and 

neptunium) and fission products (e.g., technetium) or uranium products (e.g., 236U) after 

purification. PGDP lists the isotopic activity ratios as:

Isotope Activity fraction

234U 0.8489

235U 0.0120

236U 0.1388

238U 0.0003 or 0.0004

(both listed)

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

rep

Historical quantity of radiation (usually other than X-ray or gamma radiation) originally defined 

as 93 ergs absorbed per gram in the body and redefined in the 1940s or early 1950s as the 

amount that would liberate the same amount of energy (93 ergs per gram) as 1 roentgen of 

X- or gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 9.3 milligray. The word derives from roentgen equivalent physical; rep 

is also the plural.

roentgen (R, sometimes r)

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0 degrees Celsius and standard atmospheric 

pressure. An exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft 

tissue for higher energy photons (generally greater than 100 kiloelectron-volts).

shallow absorbed dose (Ds)

Absorbed dose at a depth of 0.07 millimeters (7 milligrams per square centimeter) in a material 

of specified geometry and composition.

shallow dose equivalent [SDE, Hs, Hp(0.07)]

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

sievert (Sv)

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 sievert

equals 100 rem.

skin dose

See shallow dose equivalent.

thermoluminescence

Property that causes a material to emit light as a result of heat.
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 39 of 55

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose.

thermoluminescent neutron dosimeter (TLND)

Thermoluminescent dosimeter for measurement of neutron dose.

tissue-equivalent proportional counter (TEPC)

Device that measures absorbed dose from neutron radiation in materials nearly equivalent to 

tissue. Analysis of the counter data determines the effective weighting factor and the dose 

equivalent for that radiation.

whole-body (WB) dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose.

X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
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ATTACHMENT A 

EXTERNAL EXPOSURE TO TECHNETIUM-99

TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Introduction ............................................................................................................................ 41

A.2 Background............................................................................................................................ 41

A.3 Potential for Exposure ............................................................................................................ 41

A.4 Magnitude of Exposure .......................................................................................................... 42

A.5 Assignment of External Dose from Technetium-99................................................................. 43

A.6 Assignment of Skin Contamination Dose from Technetium-99............................................... 43

LIST OF TABLES

TABLE TITLE PAGE

A-1 Facilities with 99Tc exposure potential .................................................................................... 41

A-2 Job titles for workers with possible 99Tc exposure .................................................................. 42
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ATTACHMENT A

EXTERNAL EXPOSURE TO TECHNETIUM-99 (continued)

A.1 INTRODUCTION

This attachment provides guidance for the assignment of external dose from 99Tc for employees of

PGDP. Due to its nonpenetrating characteristics, combined with the routine use of personal 

protective equipment (PPE) by affected employees, the dose potential from 99Tc is low. Information 

on the assignment of 99Tc dose in this attachment is based on information about work location, job 

title, and job description together with shallow dose data that can be used by dose reconstructors to 

identify employees who could have been exposed to 99Tc. 

A.2 BACKGROUND

Technetium-99 is present at PGDP as a contaminant from the introduction of RU into the cascade at 

various times throughout site operations. It is a long-lived fission product with a radiological half-life of 

213,000 years and is a pure beta emitter with average and maximum energies of 84.6 keV and 

293.6 keV, respectively. Although it is difficult to detect due to its low-energy beta emission, this 

characteristic results in minimum potential for external dose. Studies have shown that the outer layer 

of skin affords significant protection to the germinal skin layers from 99Tc, and the wearing of PPE 

such as coveralls and gloves provides further skin protection. In this way, the vast majority of 

radiation from 99Tc is attenuated before it can interact with the body (ORAUT 2012b).

A.3 POTENTIAL FOR EXPOSURE

Operations procedures at PGDP required special precautions for working around significant quantities 

of 99Tc, especially during plant maintenance and repairs on the upper cascade equipment. These 

included engineering controls, administrative controls, and the use of PPE. The control measures 

routinely used to protect workers from exposure to uranium and its progeny provided an even greater 

protection factor for exposure to 99Tc (Saraceno 1981).

Specific work activities that could have resulted in exposure to 99Tc included:

• Technetium recovery operations,

• Removal of equipment from the cascade for routine maintenance, and

• Removal and replacement of cascade equipment during the Cascade Improvement Program 

and Cascade Upgrade Program.

Table A-1 lists the facilities at PGDP with 99Tc exposure potential.

Table A-1. Facilities with 99Tc exposure potential. 

Facility Description

C-409 Stabilization Building

C-410 Feed Plant

C-420 Oxide Conversion Plant

C-331 Gaseous Diffusion Process Building

C-333 Gaseous Diffusion Process Building

C-335 Gaseous Diffusion Process Building

C-337 Gaseous Diffusion Process Building

C-310 Purge and Product Withdrawal Building

C-710 Analytical Laboratory

C-400 Decontamination and Cleaning Building

C-720 Maintenance Building
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ATTACHMENT A

EXTERNAL EXPOSURE TO TECHNETIUM-99 (continued)

Employees with any of the job titles listed in Table A-2 could have had exposure to 99Tc while working 

in the facilities listed in Table A-1. The highest exposure potential would have been to maintenance 

workers in the top purge cells and to those doing change-outs of trapping media near the top purge 

cells. 

Table A-2. Job titles for workers with possible 

99Tc exposure. 

Job title

Cascade worker/operator

Chemical operator

CTW

Decontamination and decommissioning worker

Feed plant operator

Maintenance mechanic

Radiological worker

A.4 MAGNITUDE OF EXPOSURE 

Based on a review of the properties of 99Tc and the routine controls that were in place, automatic 

assignment of 99Tc dose due to skin contamination is not warranted. Guidance for assignment of 99Tc 

skin contamination dose on a case-by-case basis is provided below. It is apparent, however, that 

external exposure to 99Tc was unlikely to be measured by dosimetry due to its low-energy electron 

characteristics. In certain cases, as described below, an annual external dose assignment from99Tc 

should be included in the dose estimate under EEOICPA. 

Site evaluations at PGDP assessed the potential for an external exposure problem from 99Tc recovery 

operations and found that the likelihood of high exposure was low due to the following reasons (Baker 

et al. 1978):

• Gloves were worn routinely for all operations involving the handling of containers. 

• All material was transferred remotely from point to point, with one exception. For movement 

from one container to another, the transfer was done by pumping; the container was never 

dumped by hand. 

• The solutions were dilute. 

• Less than 20% of employee work time was spent at jobs with the potential to generate 99Tc 

contamination. 

Information about the magnitude of 99Tc exposure is available in Tc-99 Contamination (Swinth 2004). 

Measured contamination exposure levels at PGDP ranging from 10,000 to 335,849 cpm/100 cm2 were

considered, which resulted in average dose rates to the skin (calculated using the VARSKIN program) 

that ranged from 0.212 mrem/hr on contact to 0.013 mrem/hr at a distance of 10 cm in air. These 

dose rates account for the use of coveralls with a density thickness of 28 mg/cm2

. To estimate the 

skin dose from a contamination event, a contamination level of 25,000 dpm/100 cm2 (250 dpm/cm2

) 

was assumed based on the action limit for 99Tc contamination on work surfaces and hand tools 

(GAT 1963). 
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ATTACHMENT A

EXTERNAL EXPOSURE TO TECHNETIUM-99 (continued)

The dose from a contamination event is calculated as follows (Swinth 2004):

25,000 dpm/100 cm (A-1) 2 × 0.081 mrem per dpm/cm2 = 20 mrem

The assumed contamination value is greater than the average contamination level of 

13,540 cpm/100 cm2 identified by Swinth (2004). The value of 0.081 mrem/dpm/cm2 is derived from a 

value of 1.6 × 10-3 mrem/dpm/cm2 multiplied by a residence half-time of 1.5 days. This half-time is 

assumed because 99Tc can be difficult to remove from the skin. 

Because the low-energy 99Tc electrons would not have been detected by dosimetry, the potential 

unmeasured external electron dose can be estimated by assuming an ambient dose rate level of 

0.2 mrem/hr, a technetium-to-uranium progeny ratio of 0.4, and a 2,000-hour work year (Bassett 

1986):

0.2 mrem/hr (maximum ambient level) × 0.4 (Tc:U progeny ratio) × 2,000 hr/yr (A-2)

= 160 mrem/yr

Because the facilities, processes, and contaminants were similar at all three gaseous diffusion plants, 

the magnitude of exposure discussed here should be valid for PGDP.

A.5 ASSIGNMENT OF EXTERNAL DOSE FROM TECHNETIUM-99

The assignment of external dose due to the presence of 99Tc is warranted under certain 

circumstances for cancer sites on the hand. Dose assignment is limited to the hand because the 99Tc 

dose rate at distances beyond 30 cm is less than 0.08 mrem/hr and drops off rapidly at greater 

distances. The following conditions must be met to assign an external dose from 99Tc:

1. Claimant has skin cancer on the hand(s); and

2. Claimant worked in a facility where 99Tc was present (Table A-1); and

3. Claimant performed a job function that could have involved 99Tc exposure (Table A-2); and

4. Claimant dosimetry indicates a relatively high ratio (more than 2) of shallow to deep dose 

(NIOSH 2007). 

If, and only if, all four of the above conditions are met, the dose reconstructor should:

• Assign an external electron dose of 8 mrem/yr.

This value derives from an annual external dose of 160 mrem reduced by a protection factor of 95% 

to account for the use of PPE. The external dose should be assigned as electrons >15 keV and a 

constant distribution.

A.6 ASSIGNMENT OF SKIN CONTAMINATION DOSE FROM TECHNETIUM-99

Skin contamination dose due to 99Tc should be applied under certain circumstances for cancer sites 

where a documented skin contamination event occurred. The following conditions must be met to 

assign a skin contamination dose from 99Tc:

1. Claimant has skin cancer on a potentially uncovered area of the skin; and
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ATTACHMENT A

EXTERNAL EXPOSURE TO TECHNETIUM-99 (continued)

2. Claimant worked in a facility where 99Tc was present (Table A-1); and

3. Claimant performed a job function that could have involved 99Tc exposure (Table A-2); and

4. Claimant records indicate a contamination incident involving the area of the skin cancer site. 

If, and only if, all four of the above conditions are met, the dose reconstructor should:

• Assign a skin dose of 20 mrem per documented incident.

The skin contamination should be assigned as electrons >15 keV and a constant distribution.
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 45 of 55

 

ATTACHMENT B 

COWORKER DOSE ASSIGNMENT

TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Purpose ................................................................................................................................. 46

B.2 Background............................................................................................................................ 46

B.3 Applications and Limitations ................................................................................................... 46

B.4 Coworker Data Development ................................................................................................. 47

B.5 Coworker Annual Dose Summaries ....................................................................................... 48

B.6 Penetrating Dose Value for Selected Construction Trade Workers ........................................ 51

B.7 External Dosimetry Data Review............................................................................................ 51

LIST OF TABLES

TABLE TITLE PAGE

B-1 Missed external doses ........................................................................................................... 48

B-2 Annual external coworker doses modified to account for missed dose................................... 50

B-3 Annual external penetrating coworker doses modified in accordance with ORAUT-OTIB0052.......................................................................................................................................52
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 46 of 55

ATTACHMENT B 

COWORKER DOSE ASSIGNMENT (continued)

B.1 PURPOSE

The purpose of this attachment is to provide information, based on site coworker data, about

assignment of dose to PGDP workers who have no or limited monitoring data. The data in this 

attachment are to be used in conjunction with ORAUT-OTIB-0020, Use of Coworker Dosimetry Data 

for External Dose Assignment (ORAUT 2011).

B.2 BACKGROUND

An analysis of external coworker dose was performed to permit dose reconstructors to complete 

certain cases for which external monitoring data are unavailable or incomplete. Cases not having 

complete monitoring data can fall into one of several categories, including:

• The worker was unmonitored and, even by today’s standards, did not need to be monitored 

(e.g., a nonradiological worker).

• The worker was unmonitored, but by today’s standards would have been monitored.

• The worker might have been monitored, but the data are not available to the dose 

reconstructor.

• Partial information is available, but it is insufficient to facilitate a dose reconstruction.

As described in ORAUT-OTIB-0020 (ORAUT 2011), some cases without complete monitoring data 

can be processed based on assumptions and methodologies that do not involve coworker data. For 

example, many cases in the first category above can be processed by assigning ambient external and 

internal doses based on information in the relevant site profiles.

As described in Section 6.3.1, radiological operations at PGDP began in September 1952, and in 

1953 the site began using dosimeter and processing technical support from ORNL. Until July 1960, 

dosimeters were issued to a limited number of workers (i.e., those with the highest potential for 

exposure), and the badges were exchanged weekly. After that time, dosimeters were assigned to all 

workers who entered a controlled area, and the badges were exchanged and processed on an 

annual, monthly or quarterly schedule. There does not appear to be any significant administrative 

practice that would jeopardize the integrity of the recorded dose of record.

B.3 APPLICATIONS AND LIMITATIONS

1. Some workers might have worked at one or more other major sites within the DOE complex 

during their employment history. Therefore, this guidance must be used with caution to ensure 

that, for clearly noncompensable cases, unmonitored external doses from multiple site 

employments have been overestimated. This typically requires the availability of external 

coworker dosimetry data for all relevant sites.

2. Summary statistics based on the dosimetry data in this attachment do not extend beyond 1995 

because data beyond 1997 were not available, and the data for 1996 and 1997 included too 

few data points to be considered reliable. However, the absence of these data (and the 

subsequent development of dose distributions) should not interfere with the processing of 

most cases with a lack of external dosimetry data because well before 1995 the monitoring 

and reporting practices at the site ensured that essentially all workers with a potential for 
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ATTACHMENT B 

COWORKER DOSE ASSIGNMENT (continued)

external radiation exposure were monitored and the results are readily accessible. Coworker 

doses can be extended to later years if needed. However, the vast majority of PGDP 

employees with a potential for radiological exposure were likely to have been monitored in 

recent years. 

3. The data in this attachment address penetrating radiation from gamma radiation and 

nonpenetrating radiation from beta radiation. Neutron data are not presented. However, 

Section 6.5.3 should be used as the basis for assigning neutron doses, when relevant, in 

addition to the photon and beta doses assigned in accordance with this attachment.

4. External onsite ambient doses should not be included in addition to the coworker doses 

assigned in accordance with this attachment because such doses would have been included 

in the dosimetry results reported by the site, which were used as the basis for the coworker 

dose distributions presented below (ORAUT 2012a, 2006).

B.4 COWORKER DATA DEVELOPMENT

Dosimetry data for monitored workers from various sources were evaluated (see Section 7.0). The 

data selected for development of coworker doses were (1) a “history tape” containing annual data 

between 1953 and 1975 and quarterly data from 1976 to 1988, and (2) a database titled 

“OHIS_External” containing mostly quarterly data from 1989 to 1997 (although the data for 1996 and 

1997 were excluded from consideration, as discussed above). In all cases, the reported data 

corresponded to deep doses (i.e., penetrating gamma radiation) and shallow doses (i.e., penetrating 

plus nonpenetrating radiation). 

The annual data from 1953 to 1975 were prorated to account for partial years of employment based 

on an analysis of the length of monitored employment associated with the data (see Section B.5 for 

further discussion). The reported quarterly data from 1976 to 1988 were also prorated, but with a 

different approach (also described in Section B.5). The data from 1989 to 1995 included specific 

monitoring start and end dates, so they were prorated based on 365 d/yr. The data were prorated so 

coworker doses representing a full year of monitored employment could be derived; this permits the 

dose reconstructor to assign appropriate doses based on specific employment dates and job 

descriptions. 

The validity of the data selected for coworker dose development was confirmed by selecting a 

sampling of claimant dosimetry data submitted by the site as part of the EEOICPA Subtitle B program 

and comparing it with the data selected as described above. A review of annual data for 10 claimants 

with more than 150 worker-years of monitored employment at PGDP indicated excellent agreement 

between the two datasets. Specifically, a perfect match was found for more than 95% of the reported 

values. It was concluded that the data cited above are acceptable for the development of coworker 

doses for PGDP.

Adjustment for Missed Dose

According to OCAS-IG-001, External Dose Reconstruction Implementation Guideline (NIOSH 2007), 

missed doses should be assigned for dosimeter readings less than the LOD to account for the 

possibility that doses were received but not recorded by the dosimeter or reported by the site. Annual 

missed doses are calculated by multiplying the number of <LOD dosimeter readings by the dosimeter 

LOD and summing the results. These values are used as the 95th percentile of a lognormal 

distribution for calculating the probability of causation (POC). Therefore, in the Interactive 
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ATTACHMENT B 

COWORKER DOSE ASSIGNMENT (continued)

RadioEpidemiological Program (IREP), the calculated annual missed doses should be multiplied by 

0.5 and entered in Parameter 1 and a value of 1.52 should be entered in Parameter 2 to represent the 

geometric mean and geometric standard deviation, respectively.

The assignment of missed doses for monitored workers is particularly significant for PGDP workers

before August 1960 when they were monitored weekly. Table B-1 lists the maximum annual missed 

dose by era and type of radiation (penetrating gamma and nonpenetrating) based on information in 

Section 6.3.1 and Attachment C.

Table B-1. Missed external doses (rem) (based on Section 6.3.1 and Attachment C). 

Period

Penetrating 

LOD

Nonpenetrating 

LODa

Exchange 

frequency

Maximum annual 

penetrating

Maximum annual 

nonpenetrating

1953–1959 0.04 0.05 Weekly 2.080 2.600

1960 0.04 0.05 Variedb 1.280 1.600

1961–1980 0.04 0.05 Variedc 0.160 0.200

1981–1988 0.02 0.03 Variedd 0.080 0.120

1989–present 0.02 0.02 Quarterly 0.080 0.080

a. Attachment C provides an explanation for nonpenetrating LODs.

b. The exchange frequency was weekly through July 1960, then became less frequent (see footnote c).

c. Section 6.3.1 indicates that monthly, quarterly, or annual exchange frequencies were used during this period depending 

on work locations and the potential for exposure. A review of the data indicates that quarterly exchanges were 

predominant. Therefore, quarterly exchanges have been assumed here to calculate the maximum annual missed dose.

d. Section 6.3.1 indicates that either quarterly or annual exchange frequencies were used during this period depending on 

the potential for exposure. A review of the data indicates that quarterly exchanges were predominant. Therefore, 

quarterly exchanges have been assumed here to calculate the maximum annual missed dose.

Special Considerations

Certain aspects of the external dosimetry practices at PGDP (Section 6.3.1) were considered in the 

analysis of the site data. These include:

• In some cases, values less than the dosimeter LODs (Table B-1) were reported by the site. 

For example, values as low as a few millirem were reported even though the penetrating LOD 

was considered to be 20 or 40 mrem (depending on the era).

• As discussed above, before 1976 the data available to analyze coworker doses represent 

annual dose summaries for individual workers. Because these data include partial work years, 

the reported average annual doses tend to underestimate the received average annual doses 

by employees who worked an entire year. 

As described in Section B.5, an approach that is favorable to claimants was adopted in the 

development of coworker dose summaries; this approach is intended to account for any 

underestimate of doses to radiological workers at PGDP based on the above considerations.

B.5 COWORKER ANNUAL DOSE SUMMARIES

Based on the above-described information and approaches described above, PGDP coworker annual 

external dosimetry summaries were developed for use in the evaluation of external dose for certain 

claimants potentially exposed to workplace radiation, but with no or limited monitoring data from DOE. 

These summaries were developed using the following steps:
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ATTACHMENT B 

COWORKER DOSE ASSIGNMENT (continued)

1. As described in Section B.4, for data from 1953 to 1975 the reported deep and shallow doses, 

which represent annual summary data, were modified to account for partial years of 

employment. This adjustment was made by analyzing NIOSH-Division of Compensation and 

Analysis Claims Tracking System (NOCTS) employment data for PGDP workers and adjusting 

the reported doses upward by an appropriate multiplier corresponding to the average fraction 

of a year an employee worked at the site. For example, if in a particular calendar year the 

average employment period for all PGDP employees in NOCTS was 11 months, the reported 

annual doses were multiplied by 12/11, or 1.09. This permits the dose reconstructor to assign 

an appropriate prorated dose to account for partial years of employment or potential exposure.

2. For data from 1976 to 1988, the reported deep and shallow doses, which represent quarterly 

summary data, were modified to account for partial years of employment. Consistent with the 

guidelines in ORAUT-OTIB-0020 (ORAUT 2011), doses for individuals with less than 

4 quarters of data for a particular year were converted to annual doses by extrapolation (i.e., 

1 quarterly result was multiplied by 4; 2 quarterly results were multiplied by 2; and 3 quarterly 

results were multiplied by 1.333).

3. For data from 1989 to 1995, the reported deep and shallow doses, which represent primarily 

quarterly data, were modified to account for partial years of employment by multiplying the 

data by 365/x, where x is the number of days the employee was issued a dosimeter. This 

information is available for this period because the data includes monitoring start and end 

dates.

4. Half of the maximum annual missed doses in Table B-1 were added to the annual doses from 

steps 1 through 3 (with the exception of reported positive doses, in which case the maximum 

missed dose was reduced by the dose corresponding to one badge exchange, because it is 

not possible that all individual badge results were zero if a positive annual dose was reported).

5. The 50th- and 95th-percentile annual penetrating and shallow doses were derived from the 

doses calculated in step 4 by ranking the data into cumulative probability curves and extracting 

the 50th- and 95th-percentile doses for each year.

6. Because the reported shallow doses include penetrating and nonpenetrating radiation, the 

percentile doses pertaining to penetrating radiation in step 5 were subtracted from the 

percentile doses pertaining to the reported shallow doses to derive percentile doses pertaining 

to nonpenetrating radiation. 

7. The results are listed in Table B-2. These percentile doses should be used for PGDP workers 

with no or limited monitoring data in accordance with ORAUT-OTIB-0020 (ORAUT 2011). In 

general, the 50th-percentile dose can be used as a best estimate of a worker’s dose if 

professional judgment indicates the worker was likely to be exposed to intermittent low levels

of external radiation. The 50th-percentile dose should not be used for workers who were 

routinely exposed. For routinely exposed workers (i.e., those who were expected to have 

been monitored), the 95th-percentile dose should be applied. For workers who are unlikely to 

have been exposed, external onsite ambient dose should be used rather than coworker dose.

Doses to organs that are affected only by penetrating radiation (e.g., organs other than the skin, 

breast, and testes) are calculated based only on the Gamma columns in Table B-2 combined with the 

appropriate organ DCFs (NIOSH 2007). Doses to the skin, breast, and testes (and any other cancer 

location potentially affected by nonpenetrating radiation) are determined based on both the Gamma 
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ATTACHMENT B 

COWORKER DOSE ASSIGNMENT (continued)

and Nonpenetrating columns; gamma doses are assigned as photons with an energy range 

consistent with the information in this document, and nonpenetrating doses are assigned as electrons 

Table B-2. Annual external coworker doses modified to account for missed dose (rem). 

Year

Gamma

95th percentile

Gamma

50th percentile

Nonpenetrating

95th percentile

Nonpenetrating

50th percentile

1953 1.656 1.128 1.729 0.701

1954 2.218 1.183 4.386 0.970

1955 2.344 1.067 5.574 1.048

1956 2.712 1.073 4.829 1.048

1957 2.224 1.072 4.511 0.580

1958 2.019 1.040 4.021 0.466

1959 1.900 1.083 5.148 0.694

1960 1.544 0.672 3.140 0.452

1961 1.048 0.134 1.647 0.036

1962 1.024 0.080 1.422 0.059

1963 0.868 0.080 0.818 0.037

1964 0.519 0.080 0.514 0.020

1965 0.243 0.080 0.194 0.020

1966 0.225 0.080 0.242 0.020

1967 0.236 0.091 0.343 0.025

1968 0.411 0.080 0.532 0.020

1969 0.541 0.080 0.989 0.020

1970 0.349 0.080 0.763 0.020

1971 0.558 0.080 1.039 0.020

1972 0.451 0.080 1.133 0.020

1973 0.407 0.080 1.254 0.020

1974 0.217 0.080 0.854 0.020

1975 0.247 0.090 0.604 0.055

1976 0.233 0.062 0.553 0.050

1977 0.189 0.062 0.398 0.055

1978 0.193 0.089 0.150 0.031

1979 0.109 0.080 0.265 0.054

1980 0.200 0.080 0.135 0.020

1981 0.090 0.040 0.324 0.020

1982 0.053 0.040 0.712 0.020

1983 0.070 0.040 0.535 0.020

1984 0.156 0.040 0.489 0.020

1985 0.070 0.040 0.615 0.020

1986 0.130 0.040 0.755 0.020

1987 0.070 0.040 0.415 0.020

1988 0.055 0.040 0.590 0.020

1989 0.053 0.040 0.067 0.000

1990 0.040 0.040 0.052 0.000

1991 0.040 0.040 0.033 0.000

1992 0.040 0.040 0.046 0.000

1993 0.043 0.040 0.042 0.000

1994 0.040 0.040 0.037 0.000

1995 0.040 0.040 0.055 0.000
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ATTACHMENT B 

COWORKER DOSE ASSIGNMENT (continued)

>15 keV with corrections to account for clothing attenuation or other considerations. Further guidance 

is provided in ORAUT-OTIB-0017, Technical Information Bulletin: Interpretation of Dosimetry Data for 

Assignment of Shallow Dose (ORAUT 2005). 

With the methodology described above, null values for nonpenetrating dose can occur because of the 

subtraction of the reported penetrating doses from the reported shallow doses and the method 

described above, which is favorable to claimants, to establish coworker doses based on the addition 

of potential missed doses. However, a zero value in Table B-2 for nonpenetrating dose does not 

result in a dose of zero to an organ such as the skin. For example, the 50th-percentile dose to the 

skin in 1989 would be assigned entirely as 0.040 rem of photons. This approach does not result in an 

underestimation of POC (which is determined by DOL) because assigning beta dose as gamma dose 

in IREP has no negative effect (because the radiation effectiveness factors are the same for >15-keV 

electrons and >250-keV photons, and are higher for 30- to 250-keV photons). 

B.6 PENETRATING DOSE VALUE FOR SELECTED CONSTRUCTION TRADE WORKERS

Table B-3 lists penetrating dose values that have been adjusted using the guidance in Section 8.0 of 

ORAUT-OTIB-0052, Parameters to Consider When Processing Claims for Construction Trade 

Workers (ORAUT 2014). This guidance is applicable for CTWs who meet the criteria in ORAUTOTIB-0052.

B.7 EXTERNAL DOSIMETRY DATA REVIEW

PGDP Dosimetry Data database

There are many tables in this database; they contain internal and external dosimetry data. The 

external data listings are listed below, with their descriptions: 

• DRS_89_THRU_96 – External dosimetry records from 1989 to 1996, 

• DRS_97_THRU_98 – External dosimetry records from 1997 to 1998, 

• OHIS_EXTERNAL_DOSE – External dosimetry records from 1981 to 1997, 

• OHIS_EXTREMITY_DOSE – Extremity dosimetry records from 1990 to 1995, 

• OHIS_HP_SCHEDULE – Dosimetry scheduling information from 1987 to 1998, 

• OHIS_JOB_HISTORY – Personnel job history information from 1986 to 1998, 

• HISTORY_TAPE – External dosimetry records from 1953 to 1988, 

• HIS20_EDD_CALCULATED_EXPOSURE – Calculated external dosimetry exposure records 

from 1980 to 1998, and

• HIS20_EDD_INACTIVE_IRD_EXPOSURE – External dosimetry records from 1953 to 1998.
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 52 of 55

ATTACHMENT B

COWORKER DOSE ASSIGNMENT (continued)

Table B-3. Annual external penetrating coworker doses 

modified in accordance with ORAUT-OTIB-0052 (rem) 

(ORAUT 2014).

Year

Gamma

95th percentile

Gamma

50th percentile

1953 1.910 1.171

1954 2.697 1.248

1955 2.874 1.086

1956 3.389 1.094

1957 2.705 1.093

1958 2.419 1.040

1959 2.252 1.109

1960 1.913 0.693

1961 1.443 0.163

1962 1.409 0.080

1963 1.191 0.080

1964 0.702 0.080

1965 0.316 0.080

1966 0.291 0.080

1967 0.306 0.103

1968 0.552 0.080

1969 0.734 0.080

1970 0.464 0.080

1971 0.757 0.080

1972 0.607 0.080

1973 0.545 0.080

1974 0.280 0.080

1975 0.322 0.102

1976 0.302 0.063

1977 0.240 0.063

1978 0.246 0.101

1979 0.128 0.080

1980 0.256 0.080

1981 0.114 0.040

1982 0.063 0.040

1983 0.086 0.040

1984 0.206 0.040

1985 0.086 0.040

1986 0.170 0.040

1987 0.086 0.040

1988 0.065 0.040

1989 0.062 0.040

1990 0.044 0.040

1991 0.040 0.040

1992 0.040 0.040

1993 0.048 0.040

1994 0.040 0.040

1995 0.040 0.040
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 53 of 55

 

ATTACHMENT C 

SKIN DOSE ASSIGNMENT

TABLE OF CONTENTS

SECTION TITLE PAGE

C.1 GENERAL INFORMATION .................................................................................................... 54

C.2 PROCEDURE........................................................................................................................ 54

LIST OF TABLES

TABLE TITLE PAGE

C-1 Examples of skin dose assignments for badge readings in 1970............................................ 54
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 54 of 55

ATTACHMENT C 

SKIN DOSE ASSIGNMENT (continued)

C.1 GENERAL INFORMATION 

In general, the contribution to skin dose at PGDP from low-energy photons is extremely small in 

comparison with the contribution from beta particles.

All DCFs for the skin should be assumed to be 1 (ORAUT 2005).

Missed doses should be calculated based on the following LODs: 

• 1953 to 1980: 50 mrem for OW, 40 mrem for S; 

• 1981 to 1988: 30 mrem for OW, 20 mrem for S; and

• 1989 to present: 20 mrem for OW, 20 mrem for S. 

Section 6.3.1 states an OW LOD of 120 mrem for 1953 to 1980. However, this value appears to be 

speculative when compared with the LOD values for similar dosimetry systems at other sites at that 

time. As outlined in Section 6.3.1, in 1953, PGDP began using ORNL to provide and process 

dosimeters. Practices were similar to those at ORNL and other major sites. ORNL has provided 

PGDP with dosimeters from early in the operations period through the present. Based on the 

information, it appears that the reported LOD value is based on considerations involving low-energy 

beta emitters; however, this would significantly overestimate the LOD (and missed dose) when the 

principal source of exposure is uranium because the dosimeters were calibrated using uranium slabs. 

Therefore, the value has been reduced in this attachment to 10 mrem above the reported photon 

LOD. The dose reconstructor should consult Attachment A to address potential exposures to 99Tc. 

Table C-1 provides examples of skin dose assignments.

Table C-1. Examples of skin dose assignments (mrem) for badge readings in 1970 (assuming 

Paducah LODs, no clothing correction, and no 99Tc exposure).

OW reading S reading Measured dose assigned Missed dose assigned

50 0 50 (electrons) 40/2 = 20

(30- to 250-keV photons)

0 0 None 50/2 =25

(30- to 250-keV photons)

100 60 40 (electrons) AND 60 (photon energy per main 

text)

None

100 100 100 (photon energy per main text) None

0 40 40 (photon energy per main text) 50/2 = 25 (electrons)

C.2 PROCEDURE

Measured Dose

1. Subtract the reported S reading from the reported OW reading. This is the calculated 

nonpenetrating dose.

2. Assign the calculated nonpenetrating dose as electrons >15 keV. A correction factor should 

be provided for clothing, if applicable, depending on likely clothing thickness and beta energy. 

3. Assign the reported S dose as photons, partitioned by energy according to Section 6.3.4.

4. Assign the reported neutron dose (if applicable) partitioned by energy and corrected for 

neutron quality according to Section 6.4.3 (using an organ DCF of 1).
Document No. ORAUT-TKBS-0019-6 Revision No. 05 Effective Date: 02/13/2017 Page 55 of 55

ATTACHMENT C 

SKIN DOSE ASSIGNMENT (continued)

Missed Dose

1. For a badge cycle with a zero result in the OW or S reading, or both, assign a single missed 

dose.

2. If only the OW reading was reported as zero, the missed dose should be the appropriate OW 

LOD for that era (divided by 2, treated as lognormal) and considered to be electrons (corrected 

for attenuation if applicable).

3. If only the S reading was reported as zero, the missed dose should be the appropriate S LOD 

for that era (divided by 2, treated as lognormal) and considered to be 30- to 250-keV photons.

4. If both the OW and S readings were reported as zero, the missed dose should be the 

appropriate OW LOD for that era (divided by 2, treated as lognormal) and considered to be 30- 

to 250-keV photons.

5. Assign missed or unmonitored neutron dose per the direction in Section 6.5.3. 

6. If applicable, assign unmonitored 99Tc dose (as >15-keV electrons) per the direction in 

Attachment A.