ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 34

Document Title:

Paducah Gaseous Diffusion Plant – Occupational 

External Dose

Document Number: ORAUT-TKBS-0019-6

Revision: 03

Effective Date: 04/04/2007

Type of Document TBD

Supersedes: Revision 02

Subject Expert(s): James E. Turner and Bill R. Thomas

Site Expert(s): N/A

Approval: Signature on File Approval Date: 03/26/2007

Daniel S. Mantooth, Document Owner

Concurrence: Signature on File Concurrence Date: 03/27/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 03/26/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 03/27/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 04/04/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 2 of 34

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.
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 3 of 34

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

6.1 Introduction ..............................................................................................................................7

6.1.1 Purpose........................................................................................................................7

6.1.2 Scope...........................................................................................................................8

6.2 Basis of Comparison................................................................................................................8

6.3 Dose Reconstruction Parameters.............................................................................................8

6.3.1 Administrative Practices ...............................................................................................9

6.3.2 Dosimetry Technology................................................................................................11

6.3.2.1 Beta/Photon Dosimeters .............................................................................. 12

6.3.2.2 Neutron Dosimeters ..................................................................................... 13

6.3.3 Calibration ..................................................................................................................13

6.3.3.1 Beta/Photon Dosimeters .............................................................................. 13

6.3.3.2 Neutron Dosimeters ..................................................................................... 14

6.3.4 Workplace Radiation Fields ........................................................................................15

6.3.4.1 Beta/Photon Fields....................................................................................... 15

6.3.4.2 Workplace Beta/Photon Dosimeter Response.............................................. 16

6.3.4.3 Neutron Fields.............................................................................................. 17

6.4 Adjustments to Recorded Dose.............................................................................................. 18

6.4.1 Photon Dose...............................................................................................................18

6.4.2 Nonpenetrating Dose..................................................................................................18

6.4.3 Neutron Dose .............................................................................................................19

6.5 Missed Dose ..........................................................................................................................19

6.5.1 Estimating Missed and Unmonitored Photon Deep Dose............................................ 19

6.5.2 Estimating Missed and Unmonitored Shallow Dose....................................................20

6.5.3 Estimating Missed Neutron Dose................................................................................23

6.6 Uncertainty.............................................................................................................................23

6.7 Dose Reconstruction..............................................................................................................24

6.8 Organ Dose............................................................................................................................24

6.9 Attributions and Annotations .................................................................................................. 24

References ........................................................................................................................................26

Glossary ............................................................................................................................................30
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LIST OF TABLES

TABLE TITLE

6-1 Dosimeter type, period of use, exchange frequency, MDL, and potential 

annual missed dose ...............................................................................................................

PAGE

10

6-2 Major radiation sources.......................................................................................................... 15

6-3 Buildings and periods of operation ......................................................................................... 16

6-4 Technetium-99 recovery operations ....................................................................................... 16

6-5 Dose fractions for PORTS HEU storage vault in Building 345 ................................................ 18

6-6 Average recorded deep dose and maximum for any single worker by year............................ 21

6-7 Average recorded shallow dose and maximum for any worker by year .................................. 22

6-8 Shallow dose equivalent rates for 99Tc ................................................................................... 23

LIST OF FIGURES

FIGURE TITLE

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

PAGE

12

6-2 Results of neutron spectrum measurements made about 24 in. in front of 

93% – 96% HEU cylinders ..................................................................................................... 18

6-3 Historical distribution of deep dose equivalent........................................................................ 20

6-4 Historical distribution of shallow dose equivalent.................................................................... 22
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 5 of 34

ACRONYMS AND ABBREVIATIONS

cm centimeter

d day

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

GM geometric mean

GSD geometric standard deviation

HEU highly enriched uranium

Hp(d) personal dose equivalent at tissue depth d (d = 10 mm or 0.07 mm)

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

kg kilogram

MDL minimum detection level

MED Manhattan Engineering District (a DOE predecessor agency)

MeV megaelectron-volt, 1 million electron-volts

mg milligram

min minute

mm millimeter

mrem millirem

NIOSH National Institute for Occupational Safety and Health

NTA nuclear track emulsion, type A (film)

ORNL Oak Ridge National Laboratory

PGDP Paducah Gaseous Diffusion Plant

POC probability of causation

QF quality factor

RU recycled uranium

TBD technical basis document

TLND thermoluminescent neutron dosimeter
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 6 of 34

TEPC tissue-equivalent proportional counter

TLD thermoluminescent dosimeter

U.S.C. United States Code

yr year

§ section
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6.1 INTRODUCTION

6.1.1

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

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

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

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

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

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

Purpose

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 

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

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

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

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

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

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

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

derived:

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

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

• Radiation from naturally occurring radon present in conventional structures

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

 1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 8 of 34

6.1.2

Paducah Gaseous Diffusion Plant (PGDP) workers, especially those employed 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 (PGDP 1976). 

Before about July 1960, personnel dosimeters were not assigned to all workers (PGDP 1957a). 

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. External Dose Reconstruction Implementation Guideline

(NIOSH 2006) indicates that these records represent the highest quality records for assessment and 

reconstruction of doses.

Scope

Initial radiation dosimetry practices were based on experience gained during 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 Engineering District (MED) program to develop nuclear 

weapons, which began in about 1940.

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 in the 1940s, various radiation dose concepts and quantities have been 

used to measure and record occupational dose. The basis of comparison for reconstruction of 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 whole-body 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 1993). 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 (Martin Marietta 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.
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 9 of 34

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

comparability to be considered in using NIOSH (2006) 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

• 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/University of 

Indiana 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

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

controls, and personnel dosimeters to measure exposure in the workplace (Harris 1957; PGDP 

1957a,b, 1964, 1976; UCND 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:

Administrative Practices

• Personnel whole-body beta/photon dosimeters

• Pocket ionization chamber dosimeters

• 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/University of Indiana 2000).

In 1953, PGDP began using dosimeter and processing technical support provided by the Oak Ridge 

National Laboratory (ORNL) (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 used at ORNL and other major sites at that time (PGDP 1957a). 

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

Goldstein, and Wollan 1944). ORNL has provided PGDP with dosimeters from early in the operations 

period through the present. 
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Table 6-1. Dosimeter type, period of use, exchange frequency, MDL, and potential annual missed dose.

Dosimeter Period of use Monitored population

Exchange 

frequency

Laboratory 

MDL 

(rem)(a)

Maximum annual 

missed dose 

equivalent (rem)(b)

Hp(10) beta/photon dosimeters

Four-element film 1953 through 7/1960 Selected workers based on activities performed Weekly (n = 50) 0.04 1.0

Four-element film After 7/1960 through 1980 Workers in C-340, C-400, and C-410 Monthly (n =12) 0.04 0.24

After 7/1960 through 1980 Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n =4) 0.04 0.08 

After 7/1960 through 1980 Workers and visitors not likely to exceed 0.1 of 

applicable guidelines

Annual (n = 1) 0.04 0.02

Harshaw two-chip TLD Beginning 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 Beginning 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

Beginning 1989 through present Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n =4) 0.02 0.04

Hp(0.07) beta/photon dosimeters

Four-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 1980 Workers in C-340, C-400, and C-410 Monthly (n =12) 0.12 0.72

After 7/1960 through 1980 Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n =4) 0.12 0.24

After 7/1960 through 1980 Workers and visitors not likely to exceed 0.1 of 

applicable guidelines

Annual (n = 1) 0.12 0.06

Harshaw two-chip TLD Beginning 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 Beginning 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

Beginning 1989 through present Workers and visitors with potential to exceed 0.1 of 

applicable guidelines

Quarterly (n =4) 0.02 0.04

Neutron dosimetersc

Harshaw TLND Beginning 1998 to 2003 (ongoing) Selected workers based on activities performed Quarterly (n =4) 0.015 0.03

a. Estimated film dosimeter detection levels based on NIOSH (1993), NRC (1989), and Wilson et al. (1990). TLD detection levels from Martin Marietta (1994) and personal communication with site 

personnel.

b. Maximum annual missed dose (NIOSH 2006).

c. The potential annual missed dose based on laboratory irradiations is not applicable to workplace missed neutron dose.
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 11 of 34

The precise detection levels listed 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 (Martin Marietta 1994) based on a DOELAP-accredited laboratory testing protocol 

(DOE 1986). During earlier years, MDLs were subject to additional uncertainty because factors 

involving radiation field and film type, as well as 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), NRC (1989), Wilson et al. (1990), and site 

personnel. Examination of older records, where 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 

(Martin Marietta 1994). The film badge was replaced by the TLD in 1980 (PGDP 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

• 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/University of Utah 2000). This 

population of monitored individuals represents those with the highest exposure potential. After July 

1960, PGDP routine practices required the assignment of dosimeters to all workers who entered a 

controlled radiation area (BJC 2000). Dosimeters were exchanged on a routine schedule (PGDP 

1957a, 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 2000a). All dosimeters were processed, and measured results were recorded and 

used to estimate dose.

Current administrative practices are generally available (Martin Marietta 1994), as is detailed 

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

provide information on historical practices at PGDP (PACE/University of Utah 2000; BJC 2000; PGDP 

1957a, 1980; Baker ca. 1995).

6.3.2

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.

Dosimetry Technology
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6.3.2.1 Beta/Photon Dosimeters

PGDP has historically used personnel dosimeter services from ORNL. 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 thermoluminescent dosimetry systems.

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

the essentially identical two-element film dosimeter 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). 

0.1

1

10

100

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Relative Response

Photon Energy, MeV

Hp(10)

Two-Element Film 

Shield

Multi-Element Film

TLD

Figure 6-1. Estimated dosimeter photon response characteristics.

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 used in the two-element 

dosimeter. The figure shows that the two-element dosimeter over-responded in relation to Hp(10)

from 0.05 to 0.3 MeV, followed Hp(10) for higher energies, and under-responded for lower energies. 

It also shows that TLDs are capable of following Hp(10) over the energy range of interest. The 

majority of PGDP worker photon dose comes from handling uranium of low enrichment. The photon 

energy spectrum is almost entirely in the range from 30 to 250 keV (Schleien et al. 1998).

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

the unshielded and shielded portions of the film based on a uranium calibration. The two-element 

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 over-response of the unshielded portion of the film to any 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).
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6.3.2.2 Neutron Dosimeters

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

request. After 1989, this capability has been provided with 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].

The quality factors (QFs) used historically for neutrons have changed significantly. 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 (1991) are about 2 times higher than the ICRP (1983) values. 

Because a QF of 10 was used for the referenced radiation measurements, the PGDP personnel 

dosimetry, an adjustment to ICRP (1991) of at most a factor of 2 times higher would be necessary [3].

Average neutron energy is 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 neutrons from depleted and natural uranium cylinders ranged from 210 

to 360 keV (Cardarelli 1997, p. 9). Unmoderated and deuterium (water) 252Cf neutrons created were 

between 1,306 and 1,403 keV. This means the dose as monitored at PGDP since 1998 was 

overestimated and, therefore, is favorable to the claimant.

6.3.3

Potential error in recorded dose is dependent on dosimetry technology response characteristics to 

each radiation type, energy, and geometry; the methodology used to calibrate the dosimetry system; 

and the extent of similarity between the radiation fields used for calibration and that present in the 

workplace. 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 and TLND for 

neutron radiation [4].

Calibration

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 2004b). 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 1993).

Introduction of on-phantom calibration of film dosimeters and 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 from 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 
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 14 of 34

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 in NIOSH (2006, Appendix B) 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 for 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 needed 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). Thus, the determination 

of beta dose was specific to uranium.

6.3.3.2 Neutron Dosimeters

Calibration of neutron dosimeters for use at PGDP was appropriate for the work locations in which 

those dosimeters were worn (Martin Marietta 1994). Dosimeter response was characterized in a 

manner that would represent the workplace (Martin Marietta 1994). Reference dosimetry for these 

measurements was evaluated with tissue-equivalent proportional counters (TEPCs). TEPCs provide 

an absolute measure of absorbed dose in a tissue-like material and, with an appropriate algorithm, an 

estimate of the neutron QF (PNL 1995). The basis for the calibration factor was developed using data 

obtained at the Y-12 Plant in a room 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 (PNL 1990).

In 1989, field measurements for neutron flux were made by PNL 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 (PNL 1990, Measurement M23 in Table 5.9; ORNL 2005). These calibration factors apply to 

the PGDP TLNDs (Martin Marietta 1994).
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 15 of 34

6.3.4

6.3.4.1 Beta/Photon Fields

Workplace Radiation Fields

PGDP operations are 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/

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. 

Table 6-2. Major radiation sources.

Nuclide Source Half-life

Energies (MeV) and abundances 

of major radiations

Alpha Beta (max) Gamma

U-238 Primary U isotope 4.51E9 yr 4.15 (21%)

4.20 (79%)

U-235 Primary U isotope 7.1E8 yr 4.21 (6%) 0.144 (11%)

4.37 (17%) 0.163 (5%)

4.40 (55%) 0.186 (57%)

4.60 (5%) 0.205 (5%)

U-234 Primary U isotope 2.47E5 yr 4.72 (28%) 0.053(0.12%)

4.77 (72%)

Th-234 Decay product 24.1 d 0.013 (9.8%)

0.103 (21%) 0.063 (3.5%)

0.193 (79%) 0.092 (3%)

0.093 (4%)

Pa-234m Decay product 1.17 min 2.29 (98%) 0.765 (0.3%)

1.001 (0.60%)

Th-231 Decay product 25.5 hr 0.206 (13%)

0.287 (12%) 0.026 (2%)

0.288 (37%) 0.084 (10%)

0.305 (35%)

Tc-99 Impurities from RU 2.12E5 yr 0.294 (100%) None

The major facilities and associated activities at PGDP are (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

• 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/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, 
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 16 of 34

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 (PGDP 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. 

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 6-3. Buildings and periods of operation.

Site facilities Source for external dose

Operation

Begin End

C-310 Purge and Product Withdrawal UF6 process equipment and cylinders 1953 1999

C-315 Surge and Tails Withdrawal UF6 process equipment and cylinders 1953 1999

C-331, C-333, C-335, C-337 Gaseous Diffusion 

Process Buildings

UF6 process equipment and cylinders 1953 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) Ongoing

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/University of Utah 2000). 

However, some routine chemical processes would concentrate them (PACE/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 2000b). 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 nor TLD efficiently detects 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. Table 6-4 lists the 

principal locations where and periods during which recovery operations at PGDP are believed to have 

taken place (PACE/University of Utah 2000).

Table 6-4. Technetium-99 recovery 

operations.

Building Began Terminated

C-710 Before 1959 ~1959

C-400 ~1959 ~1975

6.3.4.2 Workplace Beta/Photon Dosimeter Response

Essentially all PGDP radiological work areas involved photon and beta radiation characteristic of 

operations involving uranium at low enrichments. As discussed in Section 6.3.3.1, the recorded 
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 17 of 34

responses of the PGDP beta/photon film dosimeters are favorable to claimants and need no 

adjustment.

6.3.4.3 Neutron Fields

While neutrons occur in some areas at PGDP, measured levels are low. There are no identified 

locations where measurable neutron dose was encountered (Martin Marietta 1994). Studies have 

evaluated neutron fields at gaseous diffusion plants (PNL 1995; Cardarelli 1997); these studies 

confirm Martin Marietta (1994). 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 range from 0.007 to 0.34 mrem/hr; associated QFs range 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 2000b). A representative average value is 0.2 mrem/hr based on a QF of 

about 10 (PNL 1995; Cardarelli 1997). Estimates of average neutron energies ranged from 0.25 to 

0.56 MeV (PNL 1995). Neutron monitoring of individuals was performed during a UF6 cylinderpainting project (Meiners 1999). Results of this project 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 mounted 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 6-5 lists dose fractions for the neutron energy groups (indicated by the 

dashed lines in Figure 6-4). 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 2004a).

Exposure to low enriched UF6 (less than 5%) will result in a lower neutron flux than the neutron field 

expected from highly enriched UF6 (greater than 97%) as surveyed at the Portsmouth Gaseous 

Diffusion Plant (PORTS) by Soldat and Tanner (1992). The dose fractions listed in Table 6-5 are 

favorable to claimants (Soldat and Tanner 1992).

The neutron study performed in 1990 at X10 and Y12 was the only definitive study of neutron energy 

spectra documented over the history of PGDP. It is assumed that the energy spectra are valid for the 

earlier years, given the presence of enriched uranium. 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). This element was 

processed only when requested. (NTA film had an energy threshold of about 0.5 MeV.) A review of 

data does not indicate the assignment of neutron dose before 1998.
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 18 of 34

Figure 6-2. Results of neutron spectrum measurements 

made about 24 in. in front of 93% – 96% HEU cylinders

(Soldat and Tanner 1992).

Table 6-5. Dose fractions for PORTS HEU storage vault 

in Building 345. Neutron energy group Near unshielded Cf-252 source

< 10 keV 0.300

10-100 keV 0.172

0.1-2 MeV 0.447

2-14 MeV 0.081

Favorable to claimant dose fractions 

< 10 keV 0.300

0.01- 2 MeV 0.610

0.1-2 MeV 0.081

6.4 ADJUSTMENTS TO RECORDED DOSE

6.4.1

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

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

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

1986, deep dose equivalents at PGDP have been based on DOELAP calibration to Hp(10) and 

require no adjustment. Before 1986, TLDs were calibrated in air to 137Cs, which is nearly equivalent to 

an Hp(10) on-phantom 137Cs calibration. No adjustment to the measured TLD penetrating photon 

dose is necessary. As discussed in Section 6.3.3.1, the earlier film badge deep doses are favorable 

to claimants and require no numerical adjustment.

Photon Dose

6.4.2

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.

Nonpenetrating Dose
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 19 of 34

6.4.3

Measured neutron energies at PGDP are between 0.10 and 2.0 MeV, for which the ICRP Publication 

60 radiation weighting factor is 20 (ICRP 1990). Therefore, dose reconstructors should multiply the 

reported neutron dose equivalent by a factor of 2 to be consistent with the ICRP (1990) 

recommendations to be used for reconstruction (NIOSH 2006). Apply this factor to measured and 

missed neutron doses.

Neutron Dose

6.5 MISSED DOSE

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

1. A zero dose was recorded but the worker was not monitored (most workers from 1953 to July 

1960).

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

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 (Martin Marietta 1994). Neutron dosimeters were not routinely 

assigned and doses were not recorded until about 1998. Neutron doses reported 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, favorable to claimant option because the photon dose is reliably measured. This ratio 

can be applied to selected work activities [6].

6.5.1

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

during a working career. 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 description and 

work locations. NIOSH (2006) cites several different models.

Estimating Missed and Unmonitored Photon Deep Dose

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

for earlier years, when MDLs were higher and dosimeter exchange was more frequent. NIOSH 

(2006) describes an acceptable, favorable to claimant estimate of the maximum potential missed dose 

as one-half the MDL multiplied by the number of zero dose results (the MDL/2 method). The righthand 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 persons in 

Groups 1 and 3. Figure 6-3 shows the distribution of individual annual deep dose equivalent for 
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 20 of 34

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

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

as 1 rem in any year.

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

compiled in PACE/University of Utah (2000) for monitored PGDP workers. The first four columns in 

Table 6-6 (from Table 7.4 of the PACE report) list the number of monitored workers, their average 

recorded deep dose, and the maximum individual deep dose for each year from 1953 through 1988 

(zero doses were not included). For dose reconstruction, this TBD analysis assumed that exposure 

data for each year could be represented by a lognormal distribution with a geometric mean (GM) 

equal to the average listed in column 3 of Table 6-6 and a 99th percentile equal to the maximum in 

column 4. With these assumptions, the geometric standard deviation (GSD) of the lognormal 

distribution, shown in the last column, was computed [7]. The two parameters, GM and GSD, 

determine the dose distribution assumed for the monitored workers for each year. The values from 

columns 3 and 5 in Table 6-6 can be entered directly into the Interactive RadioEpidemiological 

Program (IREP). Dose reconstructors should consider unmonitored dose values obtained from 

Tables 6-1 and 6-6. Knowledge of specific conditions for some workers at some periods could have a 

bearing on which value would be more appropriate. In the absence of such information, use the 

larger of the two values as favorable to claimants [8].

6.5.2

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

dose.

Estimating Missed and Unmonitored Shallow 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. Figure 6-4 shows the historical data for the distribution of 

shallow dose equivalent for monitored workers at PGDP (Baker ca. 1995). When compared with 

Figure 6-4, this assessment of 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 used for Group 2. As an alternative, use Table 6-7, 
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 21 of 34

Table 6-6. Average recorded deep dose and maximum for any 

single worker by year (PACE/University of Utah 2000).

Year

Number of 

workers

Average dose, 

GM (rem)

Maximum

dose (rem) GSD (rem)

1953 223 0.1398 0.820 2.14

1954 284 0.2835 1.580 2.09

1955 417 0.2419 2.500 2.72

1956 471 0.3586 4.700 3.02

1957 669 0.2517 3.190 2.97

1958 661 0.1853 3.630 3.59

1959 570 0.2015 2.360 2.88

1960 526 0.2011 2.510 2.95

1961 1,690 0.1770 2.530 3.13

1962 1,479 0.1495 2.980 3.61

1963 1,311 0.1441 3.040 3.70

1964 1,289 0.0734 1.860 4.00

1965 1,128 0.0341 1.610 5.23

1966 1,138 0.0371 1.470 5.19

1967 1,143 0.0498 1.120 3.80

1968 1,241 0.0618 1.400 3.82

1969 1,270 0.0733 1.970 4.11

1970 1,273 0.0417 0.840 3.63

1971 1,254 0.0624 1.380 3.78

1972 1,288 0.0589 1.760 4.30

1973 1,404 0.0530 1.830 4.57

1974 1,624 0.0265 1.030 4.81

1975 2,013 0.0501 1.049 3.69

1976 2,426 0.0351 1.224 4.59

1977 2,643 0.0232 0.742 4.42

1978 2,613 0.0399 0.359 2.57

1979 2,487 0.0082 0.364 5.09

1980 2,308 0.0182 0.344 3.53

1981 1,840 0.0076 0.420 5.60

1982 1,617 0.0065 0.350 5.53

1983 1,452 0.0067 0.340 5.39

1984 1,434 0.0092 0.420 5.15

1985 1,365 0.0061 0.350 5.69

1986 1,244 0.0096 0.490 5.41

1987 1,275 0.0080 0.470 5.74

1988 1,359 0.0065 0.720 7.54

which is based on the shallow-dose data for monitored workers in Baker (ca. 1995) and shown in 

Figure 6-4. This TBD analysis assumed that the dose distribution for each year could be represented 

by a lognormal function with GM equal to the average listed in column 2 of Table 6-7 and a 95th 

percentile equal to the maximum in column 3. The resultant GSD is listed in the last column. Enter 

the GM and GSD into IREP. Compare values for unmonitored shallow dose obtained for Groups 1 

and 3 from Table 6-7 with those determined by the MDL/2 method from Table 6-1. Use knowledge of 

specific job conditions and location in judging which of the two estimates is more appropriate. In the 

absence of such information, assign the larger estimate as favorable to claimants [9].

Significant nonroutine beta doses, as from skin contamination events, could be addressed in specific 

incidence reports. In such cases, dose reconstructors should consider assessments based on 

investigations conducted at the time of the incident as the best resource for dose reconstruction.
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 22 of 34

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

Table 6-7. Average recorded shallow dose and 

maximum for any worker by year (Baker ca. 1995).

Year

Average dose, 

GM (rem)

Maximum 

dose (rem) GSD (rem)

1953 0.539 4 2.36

1954 0.677 7 2.73

1955 0.776 9 2.86

1956 0.853 12 3.11

1957 0.834 11 3.03

1958 0.809 11 3.07

1959 0.783 10 2.98

1960 0.699 10 3.13

1961 0.734 8 2.79

1962 0.719 10 3.09

1963 0.645 8 2.95

1964 0.547 4 2.35

1965 0.511 2 1.80

1966 0.511 3 2.14

1967 0.528 6 2.84

1968 0.563 8 3.12

1969 0.616 5 2.46

1970 0.552 3 2.07

1971 0.631 7 2.81

1972 0.640 10 3.25

1973 0.679 10 3.17

1974 0.578 7 2.92

Potential doses from 99Tc skin contamination have been evaluated by using the VARSKIN computer 

code. The calculated shallow dose rate from uniform 99Tc skin contamination is 0.0016 mrem/hr per 

dpm/cm2 (Swinth 2004). Technetium-99 is difficult to remove from skin. Therefore, the integrated 
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 23 of 34

shallow dose resulting from 99Tc skin contamination could be relatively large. For example, with a 

residence half-time of 1.5 days, the dose is 0.081 mrem per dpm/cm2 of initial contamination.

In general, direct external beta dose from 99Tc is minimal. The unshielded shallow dose rate to bare 

skin (no clothing) at a distance of 10 cm in air from a uniformly contaminated surface is about 

1 × 10-4 mrem/hr per dpm/cm2

, as estimated with VARSKIN. The dose rate at 30 cm is only about 

1 × 10-6 mrem/hr per dpm/cm2

. Table 6-8 summarizes these three benchmark values for shallow 

dose equivalent rate as determined from VARSKIN for skin contamination and for external exposure 

with intervening air.

Table 6-8. Shallow dose equivalent rates for 

99Tc.

Condition

Dose-equivalent rate 

(mrem/hr per dpm/cm2)

Skin contamination 1.6 × 10-3

External, 10 cm air 1.0 × 10-4

External, 30 cm air 1.0 × 10-6

Some skin contamination events involving 99Tc could have occurred without being detected at the 

time. In some cases, therefore, it could be appropriate to consider an additional skin dose component 

for a reported shallow dose of a worker who could have had direct contact with 99Tc. In the absence 

of specific data, the dose reconstructor must make assumptions about the number of times per year 

an affected skin region could have been contaminated and the extent of each contamination. For

example, the dose reconstructor could assume a monthly contamination event at a specific location 

on the skin with an average level of 25,000 dpm/100 cm2 (the action limit for 99Tc contamination on 

work surfaces and hand tools at PGDP). With the assumed residence half-time of 1.5 days, the 

annual shallow dose equivalent would be 240 mrem (12 × 250 dpm/cm2 × 0.081 mrem per dpm/cm2

). 

The direct external dose rate at a distance of 10 cm from a surface contaminated at this level would 

be 0.025 mrem/hr (250 dpm/cm2 × 10-4 mrem/hr per dpm/cm2

). At 30 cm, the rate would be 

0.00025 mrem/hr.

6.5.3

Dose reconstructors should add a neutron component to the annual dose of individuals who worked in 

the cylinder yard before 1998. However, careful consideration should be given to work history. In 

general, only workers who were near cylinders for extended periods have the potential for neutron 

exposure. Estimates should be based on the neutron-to-photon ratio of 1 to 5 for dose equivalent, as 

determined from Meiners (1999). The neutron dose equivalent should be multiplied by the ICRP 

(1990) factor of 2.

Estimating Missed Neutron Dose

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 2004b). The 

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

dose). Thus, the film-badge dose uncertainty is 1.3. The uncertainty in the TLD dose is 1.15 (ORAUT

2004b), which is consistent with NIOSH (2006).
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 24 of 34

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 information 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 are in the 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) on the basis of MDL/2 times the number of zero results, as described 

in Sections 6.5.1 and 6.5.2 (NIOSH 2006).

• Individuals with no dose recorded might or might not have been radiological workers. If it is 

definite that the individual was not a radiation worker, the assigned missed dose is the 

environmental dose discussed in the Occupational Environmental Dose section of this PGDP 

Site Profile (ORAUT 2004c). Otherwise, estimate the missed dose as described in 

Section 6.5. No numerical adjustments to the missed dose are necessary.

• Multiply reported and missed neutron dose equivalents by 2 to adjust for ICRP (1990).

• Base the assignment of missed neutron dose equivalent estimate for cylinder yard workers for 

whom no neutron dose is recorded on a neutron-to-photon ratio of 1 to 5 for dose equivalent 

(Meiners 1999). Multiply the estimated neutron dose equivalent by 2 to adjust for ICRP 

(1990).

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

99Tc recovery operations (Section 6.5.2).

• See Section 6.6 for a discussion of uncertainty.

6.8 ORGAN DOSE

NIOSH (2006) discusses the conversion of measured doses to organ dose equivalent, and 

Appendix B of that document contains the appropriate dose conversion factors for each organ, 

radiation type, and energy range based on the type of monitoring performed. In some cases, 

simplifying assumptions are appropriate [10].

6.9 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

information, conclusions, and recommendations to assist in the process of worker dose 

reconstruction. These callouts are listed in this section with information that identifies the source and 

justification for each item. Conventional references are provided in the next section that link data, 

quotations, and other information to documents available for review on the Oak Ridge Associated 

Universities Team servers.

[1] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

The reviewed records did not reveal recorded neutron doses for either dosimeter.
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 25 of 34

[2] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

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

[3] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

ICRP (1991) recommended 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. Integrated Environmental Management. Consultant. 2003. 

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

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

[5] Turner, James E. Integrated Environmental Management. 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. Integrated Environmental Management. Consultant. 2003. 

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

dose equivalent can be inferred from better known photon dose. Interpretation of Meiners

(1999). 

[7] Smith, Matthew H. ORAU Team. Statistician. 2004. 

The computation of GSD is a standard mathematical formula and can be found in most 

statistical methods texts. 

[8] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

The choice that is favorable to the claimant is appropriate.

[9] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

The choice that is favorable to the claimant is appropriate.

[10] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

Appendix B of NIOSH (2006) contains tables for numerous organs. Some professional 

judgment is needed to fit particular conditions.
Document No. ORAUT-TKBS-0019-6 Revision No. 03 Effective Date: 04/04/2007 Page 26 of 34

REFERENCES

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Diffusion Plant Site Report, BJC/PGDP-167, Paducah, Kentucky.

Baker, R. C., ca. 1995, Occupational Radiation Exposure Experience, Paducah Gaseous Diffusion 

Plant, Union Carbide Corporation, Nuclear Division, Paducah, Kentucky.

Cardarelli, J. J., 1997, NIOSH Health Hazard Evaluation Report, HETA 96-1298-2651, prepared for 

the U.S. Department of Labor, Centers for Disease Control and Prevention, National Institute 

for Occupational Safety and Health, Cincinnati, Ohio.

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Fix, J. J., J. M. Hobbs, P. L. Roberson, D. C. Haggard, K. L. Holbrook, M. R. Thorson, and F. M. 

Cummings, 1982, Hanford Personnel Dosimeter Supporting Studies FY-1981, PNL-3736, 

Pacific Northwest Laboratory, Richland, Washington.

Fix, J. J., L. Salmon, G. Cowper, and E. Cardis, 1997, “A Retrospective Evaluation of the Dosimetry 

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Tennessee, April 16.

ICRP (International Commission on Radiological Protection), 1983, “Radionuclide Transformations: 

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England.

ICRU (International Commission on Radiation Units and Measurements), 1993, Conversion 

Coefficients for Use in Radiological Protection Against External Radiation, Report 57, 

Bethesda, Maryland.

Martin Marietta (Martin Marietta Energy Systems), 1994, Centralized External Dosimetry System 

(CEDS), Technical Basis for the Centralized Dosimetry System, Oak Ridge, Tennessee.
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Mason, D. D., 1986, Dose Levels of UF6 Cylinders, Martin Marietta Energy Systems, Paducah, 

Kentucky, June 5.

McDougal, B. E., 1980, Gamma Radiation from Freshly Emptied UF6 Cylinders, Union Carbide 

Corporation, Nuclear Division, Paducah, Kentucky, June 18.

Meiners, S., 1999, Paducah UF6 Cylinder Painting Project, Bechtel Jacobs Company, Paducah, 

Kentucky.

NIOSH (National Institute for Occupational Safety and Health), 1993, “Epidemiologic Use of 

Nondetectable Values in Radiation Exposure Measurements,” NIOSH Research Issues 

Workshop, September 9–10, 1993, Cincinnati, Ohio.

NIOSH (National Institute for Occupational Safety and Health), 2006, External Dose Reconstruction 

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NRC (National Research Council), 1989, Film Badge Dosimetry in Atmospheric Nuclear Tests, 

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Portsmouth Gaseous Diffusion Plant – Occupational External Dose, ORAUT-TKBS-0015-6, 

Oak Ridge, Tennessee.

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– Occupational External Dose, ORAUT-TKBS-0012-6, Oak Ridge, Tennessee.

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Paducah Gaseous Diffusion Plant – Occupational Environmental Dose, ORAUT-TKBS-0019-

4, Oak Ridge, Tennessee.

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Research.

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Dosimeter, CH-1553-A-2223, University of Chicago, Metallurgical Laboratory, Chicago, Illinois. 

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Program, Union Carbide Corporation, Nuclear Division, Paducah, Kentucky, September 2.

PGDP (Paducah Gaseous Diffusion Plant), 1957b, Penetrating Sub-Committee Activities, Report 

Number 1, Health Physics Committee, Union Carbide Corporation, Nuclear Division, Paducah, 

Kentucky, December 2.
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PGDP (Paducah Gaseous Diffusion Plant), 1957c, Penetrating Sub-Committee Activities, Report 

Number 1, Health Physics Committee, Union Carbide Corporation, Nuclear Division, Paducah, 

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Health, third edition,” Williams & Wilkins, Baltimore, Maryland. 

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GLOSSARY

absorbed dose

Amount of energy in rads or grays deposited in a substance by ionizing radiation per unit mass 

of the substance. See dose.

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.

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.

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. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

curie (Ci)

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

approximately equal to the activity of 1 gram of pure 226Ra.

deep dose equivalent (Hd)

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

per square centimeter). See dose.

dose

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

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

dose are in units of roentgens, rads, reps, or grays. Various terms narrow the type of dose, 

and some are additive: 

• Absorbed dose is the amount of energy deposited in a substance by ionizing radiation.

• Collective dose is the sum of the doses to a specific population.

• Committed dose is the dose over time (usually 50 years for workers) to a specific organ 

or tissue from an intake of radioactive material.

• Cumulative dose is the sum of all doses to the same portion of the body or to the whole 

body over time.

• Deep dose is the dose at a 1-centimeter depth in tissue (1,000 milligrams per square 

centimeter).

• Effective dose is the sum of the equivalent doses in the principal tissues and organs of 

the body, each weighted by a tissue weighting factor that accounts for the probabilities 
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of fatal and nonfatal cancers according to severity and the average length of life lost 

due to an induced cancer. It indicates the biological effect of the radiation exposure in 

that tissue. 

• Equivalent dose or dose equivalent is the absorbed dose in a tissue or organ multiplied 

by a weighting factor for the particular type of radiation. 

• Organ dose is the dose to a specific organ.

• Penetrating dose is that from higher energy photon (gamma and X-ray) radiation and 

neutron radiation that penetrates the outer layers of the skin. Nonpenetrating dose is 

that from beta and lower energy photon radiation.

• Personal dose equivalent is the dose equivalent in soft tissue below a specified point 

on the body at a specified depth. 

• Shallow dose is the dose at a 0.07-centimeter depth in tissue (7 milligrams per square 

centimeter).

• Skin dose is the dose to the skin.

• Whole-body dose is the dose to the entire body excluding the contents of the 

gastrointestinal tract, urinary bladder, and gall bladder.

dose equivalent (H)

Product of absorbed dose in units of rem or sievert 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, and thermoluminescent 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. 

film

Radiation-sensitive photographic film in a light-tight wrapping. See film dosimeter.

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 
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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 rays are very 

penetrating, but dense materials such as lead or uranium or thick structures can stop them. 

Gamma photons are identical to X-ray photons of high energy; the difference is that X-rays do 

not originate in the nucleus. 

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.

minimum detection level (MDL)

Lowest amount (mass or activity) of a substance detectable by a specific instrument or 

process. Often assumed to be the level at which a dose is detected at the two-sigma level 

(i.e., 95% of the time). Also called minimum detectable limit and minimum detection limit or 

level.

neutron

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.

neutron film dosimeter

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

nuclear track emulsion, type A (NTA)

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

has tracks caused by neutrons that visible under oil immersion with about 1,000-power 

magnification.

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. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as dose quantity for radiological 

protection. 

photon

Basic unit of electromagnetic radiation. Photons are massless “packages” of light energy that 

range from low-energy microwave photons to high-energy gamma rays. Photons have 

energies between 10 and 100 kiloelectron-volts.
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pocket ionization chamber

Cylindrical monitoring device commonly clipped to the shirt or laboratory coat pocket to 

measure ionizing radiation. Also called pencil, pocket pencil, pencil dosimeter, and pocket 

dosimeter.

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) that travel from one point to another, 

some of which can pass through or partly through solid materials including the human body.

radioactivity

Disintegration of certain elements (e.g., radium, actinium, uranium, and thorium) accompanied 

by the emission of alpha, beta, gamma, and/or neutron radiation from unstable nuclei.

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.

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 average American receives 360 millirem a year from background radiation. 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 83 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 Xor gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 8.38 milligray. The word derives from roentgen equivalent physical.

roentgen (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°C 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 centimeters (7 milligrams per square centimeter) in a 

material of specified geometry and composition.
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shallow dose equivalent (Hs)

Dose equivalent in units of rem or sievert at a depth of 0.07 centimeters (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 Sv equals 

100 rem.

skin dose

See shallow dose equivalent.

thermoluminescence

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD)

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

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

light provides a measurement of absorbed dose. 

whole-body 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

Penetrating electromagnetic radiation (photons) of short wavelength (0.001 to 10 nanometers) 

and energy less than 250 kiloelectron-volts. X-rays usually come from excitation of the 

electron field around certain nuclei. Once formed, there is no difference between X-rays and 

gamma rays, but gamma photons originate inside the nucleus of an atom.