ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the K-25 Site – Occupational 

External Dose

Document Number:

ORAUT-TKBS-0009-6 

Effective Date: 11/24/2004

Revision No.: 00

Controlled Copy No.: ________

Page 1 of 24

Subject Expert: Christopher J. Miles

Document Owner

Approval: Signature on File Date: 10/25/2004

Jay J. Maisler, Team Leader

Approval: Signature on File Date: 10/25/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 11/01/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 11/24/2004 James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

3

Acronyms and Abbreviations ...............................................................................................................4

6.0 Occupational External Dosimetry ................................................................................................5

6.1 Introduction ...........................................................................................................................5

6.2 Basis of Comparison .............................................................................................................5

6.3 Dose Reconstruction Parameters..........................................................................................6

6.3.1 K-25 Historical Administrative Practices ....................................................................... 6

6.3.2 K-25 Dosimetry Technology .........................................................................................8

6.3.3 Calibration ....................................................................................................................9

6.3.4 Workplace Radiation Fields ........................................................................................ 10

6.4 Adjustments to Recorded Dose........................................................................................... 11

6.5 Missed and unmonitored dose ............................................................................................ 13

6.5.1 Estimating Missed and Unmonitored Deep Photon Dose............................................ 13

6.5.2 Estimating Missed and Unmonitored Shallow Dose.................................................... 15

External, 30 cm air ................................................................................................................. 17

6.5.3 Estimating Missed and Unmonitored Neutron Dose.................................................... 17

6.6 Uncertainty..........................................................................................................................17

6.7 Dose Reconstruction........................................................................................................... 18

6.8 Organ Dose.........................................................................................................................19

References ........................................................................................................................................20

Glossary ............................................................................................................................................22
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LIST OF TABLES

Table

6-1 K-25 dosimeter type, period of use, exchange frequency, MDL, and potential 

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

Page

7

6-2 99Tc recovery operations ...........................................................................................................10

6-3 Major radiation emissions at K-25 ............................................................................................. 12

6-4 Recorded deep doses for workers by year at K-25.................................................................... 14

6-5 Recorded shallow doses for workers by year at K-25................................................................ 15

6-6 Shallow dose rates for 99Tc ....................................................................................................... 17

LIST OF FIGURES

Figure

6-1 K-25 film and TLD dosimeters compared to Hp(10) ....................................................................

Page

9
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 11/03/2003 00-A New Technical Basis Document for the K-25 Site 

for Occupational External Dose. Initiated by 

Joseph L. Alvarez.

Draft 01/26/2004 00-B Incorporates NIOSH and internal comments. 

Initiated by Joseph L. Alvarez.

Draft 05/06/2004 00-C Incorporates changes for consistency with other 

gaseous diffusion plants and completes comment 

resolution. Initiated by Jay J. Maisler.

Draft 07/26/2004 00-D Incorporates additional changes for consistency 

with other gaseous diffusion plants and completes 

comment resolution. Initiated by Jay J. Maisler.

Draft 09/02/2004 00-E Provides additional data for recorded deep doses 

and recorded shallow doses to workers. Initiated 

by Jay J. Maisler.

Draft 10/11/2004 00-F Incorporates responses to NIOSH comments. 

Initiated by Jay J. Maisler.

11/24/2004 11/24/2004 00 First approved issue. Initiated by Jay J. Maisler.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 4 of 24

ACRONYMS AND ABBREVIATIONS

CEDR Comprehensive Epidemiologic Data Resource

DCF Dose Conversion Factor 

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

EEOICPA Energy Employees Occupational Illness Compensation Program Act

GM geometric mean

GSD geometric standard deviation

Hp(d) Personal Dose Equivalent at tissue depth d (d = 10 mm or 0.07 mm)

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

Kg kilogram

MDL minimum detection limit

MED Manhattan Engineering District

MeV million electron volts

mm millimeter

mrem millirem

n neutron

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

NTA nuclear track, type A emulsion (also, NTB and NTB-2)

NVLAP National Voluntary Laboratory Accreditation Program

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

PIC Pocket Ionization Chamber dosimeters

RU recycled uranium

s second

Sv Sievert

TBD technical basis document

TEPC tissue-equivalent proportional counter

TLD thermoluminescent dosimeter

TLND thermoluminescent neutron dosimeter

U.S.C. United States Code

WB whole body
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6.0

6.1 INTRODUCTION

OCCUPATIONAL EXTERNAL DOSIMETRY

Site Profile documents, which include technical basis documents (TBDs), 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 the NIOSH staff in the completion of the individual work required for 

each dose reconstruction.

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

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

facility” or a “Department of Energy facility” as defined in the Energy Employees Occupational Illness 

Compensation Program Act of 2000 (EEOICPA; 42 U.S.C. § 7384l (5) and (12)).”

Workers at the K-25 site, especially those employed during peak production in the 1950s, 1960s, and 

1970s, were exposed to radiation types and energies associated with natural and recycled uranium 

enrichment processes. In addition, K-25 workers occasionally worked at the X-10 and Y-12 facilities 

and might have received exposures to other sources of radiation. Personnel dosimeter records are 

available for K-25 workers beginning in 1945. The operations and radiation safety staff routinely 

reviewed dosimeter results for compliance with radiation control limits and investigated doses 

approaching annual or quarterly dose limits. The number of individuals assigned dosimeters varied 

over the years. Dosimeter monitoring intervals also varied, depending on the potential to receive 

significant dose. During the early years, only workers entering controlled areas and likely to receive 

measurable dose received dosimeters. Beginning in 1951, dosimeters were issued to the entire work 

force as part of the security badge, although only those likely to have received measurable dose were 

processed. Beginning in 1980, all dosimeters were processed. The External Dose Reconstruction 

Implementation Guide (NIOSH 2002) identified dosimeter records as the highest quality records for a 

retrospective dose assessment.

Radiation dosimetry practices were initially based on experience gained during several decades of 

radium and X-ray medical diagnostic and therapy applications. These methods were generally well 

advanced at the start of the Manhattan Engineering District (MED) program to develop nuclear 

weapons beginning about 1940.

The EEOICPA gives responsibility to NIOSH for producing radiation dose estimates that the U.S. 

Department of Labor (DOL) will use in adjudicating certain cancer claims under the Act. The doses 

are those incurred by employees in the performance of duty at U.S. Department of Energy (DOE) and 

predecessor agencies and atomic-weapons employer facilities. EEOICPA covers two cancer claimant 

groups: (1) a Special Exposure Cohort and (2) all others. Employees in group 1 who have a specified

cancer and meet certain other criteria qualify for compensation under EEOICPA. Unlike group 2, their 

claims do not require NIOSH dose reconstruction for DOL adjudication. Oak Ridge Associated 

Universities (ORAU) has developed this TBD for the second group, which excludes the special cohort 

established for K-25 and the other gaseous diffusion plants.

6.2 BASIS OF COMPARISON 

Various radiation dose concepts and quantities have been in use to measure and record occupational 

dose since the initiation of the MED in the early 1940s. 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. Hp(d) is an operational quantity used for the 

purpose of determining the worker’s actual dose equivalent from a dosimeter reading. It is defined as 

a practical method for calibration of instruments and dosimeters to dose equivalent in International 

Commission on Radiological Protection (ICRP) Report 74 (ICRP 1996) and in International 

Commission on Radiation Units and Measurements (ICRU) Report 57 (ICRU 1998). Weakly 

penetrating radiation is significant to shallow dose equivalent, defined at d = 0.07 mm or Hp(0.07). 

Penetrating radiation is significant to deep dose equivalent and is defined at d = 10 mm or Hp(10). 

Hp(0.07) and Hp(10) are the radiation quantities used in the DOE Laboratory Accreditation Program 

(DOELAP) used to accredit DOE personnel dosimetry systems since the 1980s (DOE 1986). The 

National Voluntary Laboratory Accreditation Program (NVLAP), which is the U.S. Nuclear Regulatory 

Commission (NRC) equivalent to DOELAP, uses the same operational quantities. This TBD uses 

Hp(10) and Hp(0.07) as deep and shallow dose equivalents, respectively.

6.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of the beta, photon (X-ray and gamma ray), and neutron radiation type, energy, and 

geometry of exposure in the workplace, and the characteristics of the K-25 dosimeter responses are 

crucial to the assessment of bias and uncertainty of the original recorded dose in relation to the 

radiation quantity Hp(d). Dose reconstructors can compare earlier dosimetry systems to current 

systems to evaluate their performance, based on the premise that current systems have more 

stringent criteria, as indicated in the DOELAP and NVLAP programs.

Accuracy and precision of the recorded individual worker doses depend on (Fix et al. 1997):

• Administrative practices that facilities adopt to calculate and record personnel dose based 

on technical, administrative, and statutory compliance considerations.

• Dosimetry technology, which includes the 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 the methods of calibration to 

sources of exposure in the workplace.

• Workplace radiation fields, which can include mixed types of radiation, variations in 

exposure geometries, and environmental conditions.

An evaluation of the original recorded doses, as available, combined with detailed examinations of 

workplace radiation fields and dosimeter responses to those fields is the recommended option to 

provide the best estimate of Hp(d) for individual workers.

6.3.1

The K-25 Site used personnel dosimeters to measure and record doses from external radiation to 

designated workers throughout the history of its operations. These dosimeters include one or more of 

the following:

K-25 Historical Administrative Practices

• Personnel whole-body (WB) beta/photon and neutron dosimeters 

• Pocket Ionization Chamber (PIC) dosimeters 

• Personnel extremity dosimeters
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• Personnel neutron dosimeters.

K-25 began operations in 1945 using dosimeter and processing technical support provided by the Oak 

Ridge National Laboratory (ORNL). ORNL, then the Clinton Laboratory, had implemented its 

dosimetry methods based on the personnel beta/photon dosimeter design developed at the 

Metallurgical Laboratory at the University of Chicago (Pardue, Goldstein, and Wollan 1944). ORNL 

provided K-25 with beta/photon film dosimeters and neutron nuclear track, type A (NTA) emulsion. 

Table 6-1 summarizes the personnel dosimeters used at K-25 over the years, along with their periods 

of use, exchange frequencies, Minimum Detection Levels (MDLs), and estimated annual missed 

doses. (Note: The dosimeter exchange frequency discussed in this TBD is undergoing additional 

research to verify its accuracy. In Table 6-1, when the actual frequency is questionable, weekly is 

assumed (n=50), which provides a claimant-favorable estimate for the potential annual missed dose.) 

The listed detection levels were measured against calibrated sources and documented in laboratory 

reports (Gupton 1978; Thornton, Davis, and Gupton 1961). The ORNL external dosimetry technical 

basis document (Martin Marietta 1994) reported the current Site thermoluminescent dosimeter (TLD) 

system MDLs, based on DOELAP protocol (DOE 1986).

Table 6-1. K-25 dosimeter type, period of use, exchange frequency, MDL, and potential annual 

missed dose. 

Dosimeter Period of use

Exchange 

frequency

Reported 

laboratory 

MDL (rem)

Potential 

annual missed 

dose (rem)

Beta/photon dosimeters

ORNL two-element film dosimeter 1945-1951 Weekly (n = 50)

Controlled areas

0.03 0.75

ORNL two-element film dosimeter Nov. 1951-1953 Weekly (n = 50)

All employees for ID

0.03 0.75

ORNL multi-element film dosimeter 1953-1960 Weekly (n=50) 0.03 0.75

ORNL multi-element film dosimeter II 1960-1974 Weekly (n=50) 0.03 0.75

1975-1979 Quarterly (n=4) 0.03 0.06

ORNL TLD Class 1 1980-1987 Quarterly (n=4) 0.03 0.06

Harshaw TLD 1988 Monthly (n=12) 0.005 0.06

Neutron dosimeters

NTA 1951-1955 Weekly (n = 50) ~0.05 1.25

NTB 1956-1958 Weekly (n = 50) ~0.05 1.25

NTB-2 1958 Weekly (n = 50) ~0.05 1.25

NTA 1958-1976 Weekly (n = 50) ~0.05 1.25

NTA 1976-1978 Quarterly (n=4) ~0.05 0.1

Eastman Type 2 1978-1988 Monthly (n=12) ~0.05 0.3

Harshaw TLD 1988 Monthly (n=12) 0.015 0.09

Dose reconstruction parameters concerning K-25 administrative practices significant to dose 

reconstruction involve policies to:

• Assign dosimeters to workers

• Exchange dosimeters

• Estimate dose for missing or damaged dosimeters

• Replace any destroyed or missing records

• Evaluate and record dose for incidents

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

K-25 policies were in place for all of these parameters. Routine practices appear to have required 

assigning dosimeters to all workers who entered a controlled radiation area until November 1951, 
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 8 of 24

when all workers wore dosimeters in their photo-identification badges (Thornton, Davis, and Gupton 

1961). Dosimeters were exchanged on a routine schedule: weekly until 1976, then quarterly, and 

finally monthly after 1983. Some individual workers or designated groups of workers were on other 

exchange frequencies, up to annually, depending on exposure potential. From 1945 to 1979, ORNL 

processed dosimeters only on request. Universal processing began in 1980. ORNL recorded all 

measured results and used them to estimate dose. Doses below the detection limit were recorded as 

zero.

6.3.2

The dosimeters provided to the K-25 site by ORNL evolved with dosimetry technology. The purpose 

of dosimeter designs was to accommodate the various radiation fields that workers might encounter 

throughout the ORNL complex. The dosimeters provided to the K-25 site were generally simple, 

designed primarily to measure uranium beta/gamma fields with relatively low potential for significant 

neutron doses. The history and development was similar for the entire ORNL area because the same 

contractor was responsible for all sites. Dosimetry practices varied somewhat between sites due to 

differing radionuclide mixes.

K-25 Dosimetry Technology

6.3.2.1 Beta/Gamma Dosimeters

In 1945, ORNL implemented the beta/gamma film dosimeter design originally developed at the 

Metallurgical Laboratory at the University of Chicago (Pardue, Goldstein, and Wollan 1944). ORNL 

followed a process of dosimetry research and development leading to gradual upgrades in dosimetry 

capabilities for the complex radiation fields encountered at the Site. Other DOE sites followed this 

evolution in dosimetry capabilities, leading to site-specific multielement film and thermoluminescent 

dosimetry systems. Figure 6-1 shows the energy response characteristics of the K-25 beta/gamma 

dosimeters, based on the essentially identical two-element film dosimeter designed at the University 

of Chicago (Pardue, Goldstein, and Wollan 1944). Figure 6-1 also shows the Hp(10) response, and 

the energy response of the thermoluminescent dosimeters (Wilson et al. 1990). The curve labeled 

“Two element film” is representative of the ORNL dosimeters used from 1945 through 1978. ORNL 

used a multielement film dosimeter after 1953 (Thornton, Davis, and Gupton 1961), but processed the 

photon response as it did the two-element dosimeter, using identical shielding. Figure 6-1 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 most of the energy range.

The dosimeter response for photon energies greater than 0.06 MeV is of primary interest for K-25 

workers. Because lower energy photons (<0.06 MeV) would be unlikely to contribute significantly to 

dose at K-25, the under-response in this region is of no consequence. In most areas at K-25, the 

majority of the photon dose is attributable to photons in the 0.06- to 0.25-MeV range, energies at 

which dosimeters will overestimate exposure, reported in units of roentgen. Under some 

circumstances (see Section 6.3.4), particularly when recycled uranium is involved or where 238U 

progeny tend to concentrate, the photon dose from higher energy photons can be significant. For 

these higher energy photons, the dosimeter response would be likely to represent exposure and 

Hp(10) accurately. Overall, dosimeters used at K-25 tended to overestimate photon exposure.

The beta response of the dosimeters was calculated as the difference between the “unshielded” and 

“shielded” portions of the film with adjustment for the contribution of low-energy photon radiation to the 

unshielded region of the dosimeter. The dosimeter shallow dose was calibrated to a uranium metal 

slab, making the adjustment appropriate for K-25. The multielement film dosimeters and TLDs were 
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0.1

1

10

0 0.2 0.4 0.6 0.8 1

Energy (MeV)

Relative Response

Hp(10)

Two element film

TLD

Figure 6-1. K-25 film and TLD dosimeters compared to Hp(10). 

also calibrated to uranium slabs, and had the ability to correct more accurately for the beta energy if 

known. 

6.3.2.2 K-25 Neutron Dosimeters 

There was minimal potential for significant neutron dose at K-25. Neutron dose rates ranged from 

zero to less than 0.05 mSv/hr at K-25 locations (Martin Marietta 1994; DOE 2000). K-25 used two 

types of neutron dosimeters over the years, NTA film and TLD.

NTA Film - The nuclear track emulsion, NTA types A, B, and B-2, neutron dosimeter was an NTA film 

packet enclosed in the beta/gamma dosimeters, both film and TLD, until the adoption of the Harshaw 

TLD dosimeter in 1989.

TLD - The Harshaw TLD neutron dosimeter is an integral part of the dosimeter card placed in the 

identification badge. Processing for neutron dose is on an as-needed basis. TLDs with higher 

neutron sensitivity might be used in some cases.

6.3.3

Potential error in recorded dose is dependent on the dosimeter 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 the field 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 both the NTA film and TLDs for neutron 

radiation.

Calibration

6.3.3.1 Beta/Photon Dosimeters

Dosimeters provided to K-25 were calibrated using 226Ra or 137Cs in-air (i.e., no phantom) until the 

1986 adoption of DOELAP procedures requiring calibration with phantoms. Photon calibrations with 

the use of a phantom tended to cause overestimated doses, because backscatter from the body (the 

intent of a phantom) results in additional dose to the dosimeter for any given exposure in air.
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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). Beta calibrations were routinely to a slab 

of uranium. Thus, the determination of beta dose was specific for uranium.

6.3.3.2 Neutron Dosimeters

Calibration of neutron dosimeters for use at K-25 was appropriate for the work locations in which 

these dosimeters were worn. 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 quality factor (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 quality factor was 11, and the 

average energy range was 0.6 to 1.4 MeV (PNL 1990). 

In 1993, field measurements were made by ORNL representatives at the end row of the K-25 cylinder 

yard with a TEPC and a phantom with TLDs about 4 feet from the outside of a cylinder at about the 

middle of its length. The results were evaluated qualitatively because the dose rate was very low and 

an appropriate power supply was not available. The correction factors were similar to those in the 

Y-12 UF4 storage area and confirmed the appropriateness of these values. These correction factors 

apply to the K-25 thermoluminescent neutron dosimeters (TLNDs). 

6.3.4

Beta Fields

Workplace Radiation Fields

During normal operations at K-25, beta radiation fields were the dominant external radiation hazard. 

The source of these beta fields were the 238U daughters 234mPa and, to a lesser extent, 234Th. The 

contact beta dose rate from depleted uranium metal is 240 mrem/hr. This dose rate decreases with 

enrichment. During melting and casting operations, the 238U progeny can concentrate and produce 

beta radiation fields as high as 20 rads per hour (DOE 2000). 

In addition to what would be considered “normal” operations, K-25 processed recycled uranium (RU), 

which contained trace amounts of radioactive impurities that are not present in natural uranium feed 

material. Because these impurities were present at minute concentrations, their radiological impact 

was usually negligible. However, because many routine chemical processes would concentrate them, 

their presence is worthy of discussion. The most significant beta-emitting impurity found in RU is 99Tc, 

which tends to deposit in enrichment equipment and will “pocket” in the higher enrichment sections of 

the gaseous diffusion cascade (DOE 2000). In addition, 99Tc was concentrated at K-25 for purposes 

of recovery and removal. The relatively low-energy beta (maximum 294 keV) from 99Tc poses 

minimal external exposure potential due to its limited range. Neither film dosimeters nor TLDs will 

detect 99Tc betas efficiently, 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 6-2 lists locations and periods of recovery operations at K-25 (Bechtel 

Jacobs 2000) at which the highest concentrations of 99Tc probably existed.

Table 6-2. 99Tc recovery operations. 

Trap locations Period

K-1031 1952 - 1962

K-1410 1952 - 1962

K-1420 1960 - 1985
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Photon Fields

In many areas at K-25, the primary cause of photon doses during normal operations was gamma rays 

from 235U, with primary photon energies ranging from 144 to 205 keV. This is particularly true in areas 

where enriched uranium was present (with a higher than natural abundance of 235U). The photon 

dose rate from uranium is proportional to the level of enrichment. For 93% enriched uranium, the 

most highly enriched uranium at K-25, the contact photon dose rate is about 30 mrem/hr. The contact 

photon dose rate from depleted uranium metal is less than 10 mrem/hr. 

In some cases, especially when activities involve maintenance and decontamination of equipment 

contaminated with 238U progeny or melting and casting operations, gamma radiation and 

bremsstrahlung from 234mPa can be the most significant source of photon dose. In addition, 238U 

progeny will condense out in process equipment, causing elevated photon dose rates. Gamma dose 

rates from empty UF6 cylinders can be significantly greater than those from full cylinders because the 

UF6 in the full cylinders effectively shields high-energy photons from uranium progeny on internal 

surfaces. Bremsstrahlung, primarily from the energetic 234mPa beta, can contribute as much as 40% 

of the photon dose from uranium metal (DOE 2000). 

The impurity from RU that presents the greatest potential radiological hazard from external sources is 

232U (DOE 2000). The health hazards of 232U are due primarily to the rapid buildup of gamma activity 

of its decay products, particularly 228Th (the gamma dose from 228Th is primarily from the decay 

product 208Tl). The 232U decay products form nonvolatile fluorides that concentrate in cylinders when 

UF6 is vapor-fed. Estimates indicate that the level of gamma activity in the enrichment cascade 

equipment would increase by about a factor of 3 due to the presence of 232U (DOE 2000).

Neutron Fields

While neutrons occur in some areas at K-25, the measured levels are low. Several studies have 

evaluated neutron fields at gaseous diffusion plants (PNL 1995; Cardarelli 1996); these studies have 

shown neutron dose to be minimal in all areas. Cylinder yards, feed and withdraw areas, and 

locations where uranium forms deposits in the cascade have been investigated (Cardarelli 1996). 

These studies identified the storage cylinders, which contained either depleted UF6 (tails) or enriched 

UF6 (product), as areas where neutron fields might represent an exposure hazard. Estimates of dose 

equivalent rates range from 0.007 to 0.34 mrem/hr; associated quality factors range from 7 to 10. A 

representative average value is 0.2 mrem/hr based on a quality factor of about 10 (PNL 1995; 

Cardarelli 1996). Estimates of average neutron energies ranged from 0.25 to 0.56 MeV (PNL 1995). 

Neutron monitoring of individuals was performed at the Paducah Gaseous Diffusion plant during a 

UF6 cylinder-painting project (Meiners 1999). Results of this project indicated a neutron-to-photon 

dose equivalent ratio of approximately 20%, based on a quality factor of 10. The associated neutronto-photon absorbed dose ratio is 2%. Dose reconstructors should assume that these ratios are the 

same at K-25.

Table 6-3 summarizes the major K-25 radionuclides associated with external dose and their 

emissions.

6.4 ADJUSTMENTS TO RECORDED DOSE

Photon Dose

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

national and international. The current reporting unit used by DOE is the mrem, the unit of dose 

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

1986, deep dose equivalents at K-25 were based on DOELAP calibration to Hp(10) and require no 
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 12 of 24

Table 6-3. Major radiation emissions at K-25. 

Nuclide Half-life

Energies (MeV) and abundances of major radiations

Alpha Beta (max) Gamma

Primary uranium isotopes

U-238 4.51 × 109 yr 4.15 (25%) -- --

4.20 (75%)

U-235 7.1 × 108 yr 4.37 (18%) -- 0.144 (11%)

4.40 (57%) 0.185 (54%)

U-234 2.47 × 105 yr 4.58 ( 8%) -- 0.204 (5%)

4.72 (28%) 0.053 (0.20%)

4.77 (72%)

Decay products

Th-234 (U-238) 24.1 d -- 0.103 (21%) 0.063 (3.5%)

0.193 (79%) 0.093 (4%)

Pa-234m (U-238) 1.17 m 2.29 (98%) 0.765 (0.30%)

1.001 (0.60%)

Th-231 (U-235) 25.5 hr 0.140 (45%) 0.026 (2%)

0.220 (15%) 0.084 (10%)

0.305 (40%)

Impurities (RU)

Tc-99 2.12 × 105 yr -- 0.294 (100%) --

U-236 2.34 × 107 yr 4.49 (76%)

4.44 (24%)

U-232 72 yr 5.26 (31%)

5.32 (69%)

Th-228 (U-232) 1.9 yr 5.34 (28%)

5.43 (71%)

Pb-212 (U-232) 10.64 hr 0.346 (81%) 0.239 (47%)

0.586 (14%) 0.300 (3.2%)

Bi-212 (U-232) 60.6 min 6.05 (25%) 1.55 (5%) 0.727 (7%)

6.09 (10%) 2.26 (55%)

Tl-208 (U-232) 3.10 min 1.28 (25%)* 0.511 (23%)*

1.52 (21%)* 0.583 (86%)*

1.80 (50%)* 0.86 (12%)*

2.614 (100%)*

* Note: branching ratio of 36% from 212Bi.

adjustment. Prior to 1986, dosimeters might have been calibrated in air (i.e., no phantom). In-air 

calibration will tend to cause overestimates of photon exposures, which is claimant-favorable when 

roentgen-to-organ dose conversion factors (DCF) are applied (see Section 6.8). In general, 

dosimeters overestimated worker exposures to photons. This is particularly the case with film 

dosimeters for the most common photon energies encountered at K-25. No adjustments to recorded 

film or TLD photon exposures are necessary or recommended. Refer to Section 6.8 for converting 

reported exposures to organ dose equivalents. 

Shallow Dose

Beta calibrations were routinely to a slab of uranium, the primary source of shallow dose at K-25. No 

adjustment to recorded dose from dosimeter measurements is recommended.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 13 of 24

Neutron Dose 

The neutron energies at K-25 are between 0.1 and 2.0 MeV, for which the ICRP Publication 60 

radiation weighting factor is 20 (ICRP 1990). The associated dose-correction factor is 1.91, which is 

rounded to 2. Dose reconstructors should apply this factor to both measured neutron dose equivalent 

and missed neutron dose equivalent.

6.5 MISSED AND UNMONITORED DOSE

The potential for missed dose exists when workers are exposed to radiation at levels below the 

detection limit of their personnel dosimeters. In the early years of radiation monitoring, when relatively 

high detection limits are combined with short monitoring durations, missed doses can be significant. 

Methodologies for estimating missed doses are discussed in this section.

Unmonitored dose pertains to the potential dose received by workers who did not wear personnel 

dosimetry. For these cases, dose reconstructors must rely on coworker data and/or population data 

to estimate a worker’s potential unmonitored dose. These methods are also discussed in this section.

For the special case when a worker’s exposure potential has been determined to be low, the 

environmental dose should be assigned. 

6.5.1

Watson et al. (1994) examined methods analysts can consider when there is no recorded dose for a 

period during a working career. 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 (2002) describes options to calculate missed dose for this situation. The preferred 

option estimates a claimant-favorable maximum potential missed dose as (MDL)/2 multiplied by the 

number of zero-dose results. Table 6-1 lists the results of these calculations. 

Estimating Missed and Unmonitored Deep Photon Dose

Methods for estimating unmonitored doses are discussed in some detail in the External Dose 

Reconstruction Implementation Guide (NIOSH 2002). One approach involves the solicitation and use 

of co-worker data, if available. In general, the maximum reasonable co-worker dose should be 

assigned as a claimant-favorable estimate. For cases where dosimetry data is intermittent, i.e., 

workers appear to have been monitored during some time periods and unmonitored for others, the 

monitoring data prior to and after the missing data can be used to interpolate the missing data. 

Another approach to estimating unmonitored dose would be to use population data from K-25. 

It is reasonable to assume that monitored workers were more likely to receive significant dose than 

unmonitored workers. That being the case, the average reported dose for monitored workers should 

represent a claimant-favorable estimate of the dose that may have been received by an unmonitored 

worker. The first five columns in Table 6-4, show for each year from 1945 through 1988, the number 

of monitored workers receiving zero dose, the number of monitored workers receiving positive dose, 

their average recorded deep dose, and the maximum reported individual deep dose (Maisler 2004). 

(Zero doses were not included in compiling the average.) For use in dose reconstruction, when the 

number of workers receiving positive dose exceed 100, it was assumed that the exposure data for the

year could be represented by a lognormal distribution with a geometric mean (GM) equal to the 

average shown in column 4 of Table 6-4 and a 99th percentile equal to the maximum in column 5. 

With these assumptions, the geometric standard deviation (GSD) of the lognormal distribution, shown 

in column 6, and the 95th percentile dose, last column, were computed. The two parameters, GM and 

GSD, thus determine the dose distribution assumed for the monitored workers for each year. Their 

values from columns 4 and 6 in Table 6-4 can be entered directly into IREP. 
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 14 of 24

Table 6-4. Recorded deep doses for workers by year at K-25

Year Number of 

monitored 

workers 

receiving zero 

dose

Number of 

monitored 

workers 

receiving 

positive dose

Average 

Deep Dose, 

GM (rem)*

Maximum 

Deep Dose 

(rem)

GSD 

(rem)

95th

Percentile 

Dose 

(rem)

1945 0 2 0.475 0.530 ** **

1946 22 30 0.161 1.590 3.39 1.198

1947 22 22 0.086 0.490 2.59 0.412

1948 165 92 0.096 3.000 3.38 0.712

1949 377 258 0.145 1.160 2.44 0.629

1950 573 117 0.186 6.195 4.50 2.209

1951 489 232 0.132 0.605 1.93 0.386

1952 753 417 0.118 1.190 2.70 0.603

1953 594 579 0.150 1.510 2.70 0.765

1954 825 189 0.164 1.435 2.54 0.759

1955 929 129 0.093 0.825 2.55 0.435

1956 814 80 0.140 0.735 2.04 0.451

1957 596 163 0.278 3.025 2.79 1.499

1958 608 166 0.297 4.535 3.22 2.035

1959 373 443 0.192 2.475 2.99 1.168

1960 430 505 0.190 3.877 3.64 1.598

1961 549 676 0.089 2.350 4.07 0.898

1962 86 40 0.032 0.400 2.40 0.133

1963 71 42 0.036 0.151 1.92 0.106

1964 68 46 0.026 0.348 2.30 0.103

1965 79 71 0.065 0.544 2.18 0.236

1966 66 60 0.059 1.532 2.85 0.330

1967 66 59 0.068 1.398 2.69 0.345

1968 98 29 0.079 1.334 3.34 0.575

1969 79 35 0.070 0.797 3.01 0.429

1970 84 29 0.099 0.722 2.63 0.487

1971 68 46 0.066 0.622 2.73 0.343

1972 19 92 0.063*** 0.693 0.102*** 0.267

1973 93 38 0.045 0.331 2.84 0.249

1974 2 138 0.122 1.680 3.08 0.777

1975 466 4448 0.025 0.448 3.46 0.191

1976 3829 817 0.125 2.256 3.46 0.963

1977 4690 1266 0.326 1.960 2.16 1.157

1978 4554 1356 0.097 1.508 3.25 0.673

1979 5358 366 0.118 3.272 4.17 1.231

1980 1470 1653 0.126 7.932 5.92 2.346

1981 184 2921 0.070 2.545 4.68 0.884

1982 697 575 0.061 1.579 4.04 0.607

1983 866 1263 0.068 0.956 3.11 0.439

1984 887 1195 0.051 0.984 3.56 0.412

1985 889 1149 0.141 0.840 2.15 0.497

1986 803 1146 0.383 1.060 1.55 0.786

1987 988 1158 0.249 0.930 1.76 0.631

1988 1196 1166 0.243 1.080 1.90 0.696

* When the number of monitored workers receiving positive dose exceeded 100, the values in this column

represent the arithmetic mean, rather than geometric mean (GM). Since the GM is actually less than the 

arithmetic mean for these cases, treatment of these values as GMs for purposes of dose reconstruction will 

produce claimant-favorable results.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 15 of 24

** Only 2 data points exist for year 1945, therefore values for GSD and 95th percentile are not provided.

*** The data for year 1972 were normally distributed. The values for mean and GSD are the arithmetic mean 

and standard deviation.

Dose reconstructors may use any of the methods discussed in this section to assign unmonitored 

dose or assign the estimated “missed dose” values from Table 6-1, whichever is most claimantfavorable.

6.5.2

The method used to estimate missed deep photon dose is also applicable to missed shallow dose, 

i.e., MDL/2 multiplied by the number of zero-dose results. Table 6-1 lists the results of these 

calculations.

Estimating Missed and Unmonitored Shallow Dose

All of the methods discussed above for estimating unmonitored deep dose for photons are also 

applicable to unmonitored shallow dose. If the unmonitored shallow dose is to be estimated based on 

the K-25 population data, the approach should be identical to that described above for unmonitored 

deep dose and Table 6-5 should be used (Maisler 2004). 

Table 6-5. Recorded shallow doses for workers by year at K-25

Year Number of 

monitored 

workers 

receiving 

zero dose

Number of 

monitored 

workers 

receiving 

positive 

dose

Average 

Shallow

Dose, GM 

(rem)*

Maximum 

Shallow

Dose 

(rem)

GSD 

(rem)

95th

Percentile 

Dose 

(rem)

1945 0 2 1.248 1.335 ** **

1946 15 37 0.234 4.150 4.74 3.021

1947 21 23 0.359 3.990 3.64 3.005

1948 164 93 0.252 6.000 3.43 1.913

1949 299 336 0.379 7.334 3.57 3.068

1950 492 198 0.222 6.205 4.17 2.331

1951 346 375 0.326 3.100 2.63 1.599

1952 521 649 0.563 7.970 3.12 3.656

1953 395 778 0.670 9.160 3.07 4.245

1954 594 420 0.544 7.290 3.04 3.399

1955 817 241 0.299 7.765 4.05 2.980

1956 725 169 0.512 2.985 2.13 1.777

1957 520 239 0.681 10.850 3.28 4.806

1958 508 266 0.843 10.955 3.01 5.153

1959 280 536 0.680 8.157 2.90 3.929

1960 379 556 0.565 10.569 3.52 4.467

1961 495 730 0.399 17.449 5.06 5.744

1962 51 75 0.128 1.386 3.37 0.944

1963 42 71 0.138 2.185 3.06 0.866

1964 38 76 0.121 0.720 2.60 0.581

1965 47 103 0.191 1.577 2.53 0.878

1966 58 68 0.345 2.790 2.95 2.045

1967 52 73 0.251 2.807 3.21 1.711

1968 69 58 0.177 1.774 2.90 1.023

1969 60 54 0.317 12.060 4.05 3.163
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 16 of 24

Table 6-5. Recorded shallow doses for workers by year at K-25 (continued)

Year Number of 

monitored 

workers 

receiving 

zero dose

Number of 

monitored 

workers 

receiving 

positive 

dose

Average 

Shallow

Dose, GM 

(rem)*

Maximum 

Shallow

Dose 

(rem)

GSD 

(rem)

95th

Percentile 

Dose 

(rem)

1970 64 49 0.397 18.602 4.21 4.222

1971 55 59 0.234 6.327 3.91 2.207

1972 17 94 0.185 5.135 2.26 0.706

1973 80 51 0.196 2.381 3.04 1.219

1974 2 138 0.196 3.849 3.59 1.604

1975 209 4705 0.076 1.808 3.89 0.713

1976 3438 1208 0.149 3.392 3.82 1.353

1977 4328 1627 0.282 2.588 2.59 1.348

1978 3851 2059 0.108 5.632 5.45 1.763

1979 4930 794 0.136 6.972 5.41 2.193

1980 1306 1817 0.438 23.610 5.54 7.308

1981 128 2977 0.150 6.790 5.14 2.212

1982 576 696 0.187 4.540 3.93 1.778

1983 758 1371 0.241 4.000 3.34 1.750

1984 785 1297 0.274 4.960 3.46 2.117

1985 816 1222 0.353 7.720 3.76 3.116

1986 722 1227 0.321 1.570 1.98 0.984

1987 830 1316 0.288 6.370 3.78 2.563

1988 1026 1336 0.264 6.550 3.97 2.547 * When the number of monitored workers receiving positive dose exceeded 100, the values in this column

represent the arithmetic mean, rather than geometric mean (GM). Since the GM is actually less than 

the arithmetic mean, treatment of these values as GMs for purposes of dose reconstruction will produce 

claimant-favorable results. 

** Only 2 data points exist for year 1945, therefore values for GSD and 95th percentile are not provided.

Specific incident reports might address significant nonroutine worker doses, such as skin 

contamination events. The dose assessments in such reports, based on investigations conducted at 

the time of the incident, should be the best estimates of dose received.

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

code. The calculated 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. For this reason, the integrated dose 

resulting from 99Tc skin contamination could be relatively high (e.g., 0.082 mrem per dpm/cm2 of initial 

contamination, assuming a residence half-time on skin of 1.5 days). 

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

skin (no gloves or clothing) at a distance of 10 cm from a contaminated surface is about 10-4 mrem/hr 

per dpm/cm2

, as estimated with VARSKIN. Due in large part to shielding provided by the air, the dose 

rate at a distance of 30 cm would only be about 10-6 mrem/hr per dpm/cm2

. Table 6-5 summarizes 

these three benchmark values for shallow dose rate as determined from VARSKIN for skin 

contamination and for external exposure with intervening air.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 17 of 24

Table 6-6. Shallow dose rates for 99Tc.

Condition

Dose rate 

(mrem/hr per dpm/cm2)

Skin contamination 0.0016

External, 10 cm air 10-4

External, 30 cm air 10-6

Some skin contamination events involving 99Tc could have gone undetected. In some cases, it might 

be appropriate to add an additional skin dose component to a reported shallow dose for a worker who 

might have had direct contact with 99Tc. To estimate the annual missed dose in the absence of 

specific data, one must make assumptions regarding the number of times per year an affected skin 

region might have received contamination and the extent of each contamination. For example, one 

might assume a monthly contamination event at a specific location on the skin at an average level of 

25,000 dpm/100 cm2 (the action limit for 99Tc contamination on work surfaces and hand tools at 

Portsmouth Gaseous Diffusion Plant). The annual dose for this example would be (12 × 250 dpm/cm2

× 0.082 mrem per dpm/cm2

) = 240 mrem. The direct external dose rate at a distance of 10 cm from a 

surface contaminated at this same level would be (25,000 dpm/100 cm2 × 10-6 mrem/hr per dpm/100 

cm2

) = 0.025 mrem/hr at a distance of 10 cm and 0.00025 mrem/hr at a distance of 30 cm. 

6.5.3

Neutron dosimeters were generally insensitive to the low neutron dose rates at K-25, which resulted in 

routine recordings of “zero” dose. A neutron component should be part of the annual dose of 

individuals who worked in the cylinder yard. However, dose reconstructors should give careful 

consideration to the claimant work history. In general, only workers who were near cylinders for 

extended periods have the potential for neutron exposure. Base an estimate on the neutron-tophoton ratio of 20% for dose equivalent, as determined from the survey conducted at Paducah 

Gaseous Diffusion Plant (Meiners 1999). Multiply the dose equivalent by 2 to adjust for the ICRP 

(1990) dose factor.

Estimating Missed and Unmonitored Neutron Dose

6.6 UNCERTAINTY

A number of factors contribute to uncertainty in measured doses. Systematic errors can occur from 

calibration and processing as well as from extraneous conditions such as moisture, heat, and fading. 

Random errors also arise from variations among workers and the energy spectra and geometries of 

their exposures. No specific uncertainty assessments for K-25 dosimeter systems were found. 

However, the systems have much in common with ones used at other laboratories, such as Hanford, 

for which extensive studies have been carried out (Wilson et al. 1990; Fix, Gilbert, and Baumgartner

1994). The similarities enable reasonable comparisons for the K-25 systems, based on experience at 

Hanford and elsewhere. Descriptions of uncertainties at the Portsmouth Gaseous Diffusion Plant 

have also been based on the Hanford studies (PORTS 2004). Operations and dosimetry are similar 

at Portsmouth and K-25. 

Guidance for estimating uncertainty in external dose reconstruction is provided in NIOSH (2002). 

Under good laboratory conditions, film-badge uncertainty can be at the level of 10% to 15%. The 

absolute uncertainty at 95% confidence should not be less than the MDL, which for K-25 was 

0.03 rem for beta/gamma film (Table 6-1). Figure 2.1 of NIOSH (2002) shows the results from two 

methods of calculating the uncertainty factor. In the absence of any other site-specific data, it is 

recommended that numerical values used there be employed for film-badge dose reconstruction. 
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 18 of 24

The uncertainty for TLDs is generally smaller than that for film and somewhat less dependent on 

energy. An uncertainty factor of 1.15 is suggested as claimant-favorable for doses of 0.10 rem and 

above and a factor of 1.5 for doses below 0.10 rem. [The factor 1.5 for doses throughout the range 

between the MDL and 0.10 rem is adequate to compensate for the actual increase to about 2.0 at the 

MDL, which for K-25 is 0.03 rem (Table 6-1).] These assessments appear to be in line with the data 

reported elsewhere.

Relatively little information is available on uncertainty for shallow dose. In view of the similar 

mechanisms between photon and beta film dosimetry, NIOSH (2002) recommends applying the 

methodology just described to the beta dose.

6.7 DOSE RECONSTRUCTION

As much as possible, the basis for dose to individuals should be the 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 contains appropriate information:

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

are adequate. Dose reconstructors should consider beta energies to be greater than15 keV 

and, unless known to be otherwise, photon energies to be within the claimant-favorable energy 

range of 30 to 250 keV.

• Dose reconstructors should assign missed dose for workers for whom dosimetry records 

provide zero beta-photon values for Hp(10) and Hp(0.07), as identified in NIOSH (2002). This 

approach is conservative based on a review of historic data.

• For unmonitored workers, dose reconstructors should solicit and use coworker data. In 

general, assign the maximum reasonable coworker dose as a claimant-favorable estimate. As

an alternative, use population data.

• For unmonitored workers whose exposure potential has been determined to be low, assign the 

environmental dose.

• Dose reconstructors should multiply neutron doses assigned since 1998 by 2 to adjust for the 

ICRP (1990) dose factor.

• Dose reconstructors should assign missed neutron dose for cylinder yard workers for whom 

there is no recorded neutron dose, based on a neutron-to-photon ratio of 20% for dose 

equivalent (Meiners 1999). Multiply the estimate by 2 to adjust for the ICRP (1990) dose 

factor 0.1 to 2 MeV.

• Dose reconstructors should pay special attention to the possibility of skin contamination 

incidents for workers involved with 99Tc recovery operations (Section 6.5.2)
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 19 of 24

6.8 ORGAN DOSE

NIOSH (2002) 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. For periods when calibrations were performed in free air, prior to 1986, 

dose reconstructors should use the exposure to organ DCF. For recorded doses from 1986 to the 

present, use the Hp(10)-to-organ DCF.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 20 of 24

REFERENCES

Bechtel Jacobs (Bechtel Jacobs Company LLC), 2000, Recycled Uranium Mass Balance Project Oak 

Ridge Gaseous Diffusion Plant, BJC/OR-584, Oak Ridge, Tennessee.

Cardarelli J. J., 1996, Health Hazard Evaluation Report, 96-0198-2651, Portsmouth Gaseous 

Diffusion Plant, Piketon, Ohio.

DOE (U.S. Department of Energy), 1986, Department of Energy Standard for the Performance 

Testing of Personnel Dosimetry Systems, DOE/EH-0027, Washington, D.C.

DOE (U.S. Department of Energy), 2000, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136-2000, Washington, D.C.

Fix, J. J., E. S. Gilbert, and W. V. Baumgartner, 1994, An Assessment of Bias and Uncertainty in 

Recorded Dose from External Sources of Radiation for Workers at the Hanford Site, PNL10066, Pacific Northwest Laboratory, Richland, Washington.

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

Employed in an International Combined Epidemiologic Study,” Radiation Protection Dosimetry 

74, 39-53.

Gupton, E. D., 1978, Dosimetry with the ORNL Badge, ORNL/TM-6357, Oak Ridge National 

Laboratory, Oak Ridge, Tennessee.

ICRP (International Commission on Radiological Protection), 1990, “1990 Recommendations of the 

International Commission on Radiological Protection,” Publication 60, Annals of the ICRP, 

volume 21, numbers 1-3, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1996, Conversion Coefficients for Use in 

Radiological Protection Against External Radiation, Publication 74, Pergamon Press, Oxford, 

England.

ICRU (International Commission on Radiation Units and Measurements), 1998, Conversion 

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

Bethesda, Maryland.

Maisler, J. J., 2004, Memorandum to File, Integrated Environmental Management, Inc., Dayton, Ohio.

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

(CEDS), Technical Basis for the Centralized Dosimetry System.

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

Kentucky.

NIOSH (National Institute for Occupational Safety and Health), 2002, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Office of Compensation Analysis and Support, 

Cincinnati, Ohio.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 21 of 24

PACE (Paper, Allied Industrial, Chemical and Energy Workers International Union) and University of 

Utah, 2000, Exposure Assessment Project at the Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky. 

Pardue, L. A., N. Goldstein and E. O. Wollan, 1944, Photographic Film as a Pocket Radiation 

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

PNL (Pacific Northwest Laboratories), 1990, Neutron Dose Equivalent and Energy Spectra 

Measurements at the Oak Ridge National Laboratory and Y-12 Plant, Richland, Washington.

PNL (Pacific Northwest Laboratories), 1995, Enriched Uranium Cylinder Neutron Characterization at 

the Portsmouth Gaseous Diffusion Plant, Richland, Washington.

PORTS (Portsmouth Gaseous Diffusion Plant), 2004, Technical Basis Document for Portsmouth 

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

Swinth, K. L., 2004, E-mail correspondence dated 3/5/04 and 3/19/04, External Dosimetry Task 

Group..

Thornton, W. T., D. M. Davis, and E. E Gupton, 1961, The ORNL Badge Dosimeter and its Personnel 

Monitoring Applications, ORNL-3126, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Watson, Jr., J. E., J. L. Wood, W. G. Tankersley, and C. M. West, 1994, “Estimation of Radiation 

Doses for Workers Without Monitoring Data for Retrospective Epidemiologic Studies,” Health 

Phys. 67(4):402-405.

Wilson, R. H., J. J. Fix, W. V. Baumgartner, and L. L. Nichols, 1990, Description and Evaluation of the 

Hanford Personnel Dosimeter Program From 1944 Through 1989, PNL-7447, Pacific 

Northwest Laboratory, Richland, Washington.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 22 of 24

GLOSSARY

beta radiation

Radiation consisting of charged particles of very small mass (i.e., the electron) emitted 

spontaneously from the nuclei of certain radioactive elements. Most (if not all) direct fission 

products emit beta radiation. Physically, the beta particle is identical to an electron moving at 

high velocity.

curie 

A special unit of activity. One curie exactly equals 3.7 x 1010 nuclear transitions per second.

deep dose equivalent (Hd) 

The dose equivalent at the respective depth of 10 mm in tissue.

dose equivalent (H)

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The special unit is the rem. When D is expressed in Gy, H is in Sieverts (Sv). (1 Sv = 100 

rem.)

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. (See film dosimeter, neutron 

film dosimeter, thermoluminescent dosimeter.)

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external and/or internal sources of radiation.

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, 

and/or extremities. This includes the fabrication, assignment, and processing of the 

dosimeters as well as interpretation and documentation of the results.

film

In general, a "film packet" that contains one or more pieces of film in a light-tight wrapping. 

The film when developed has an image caused by radiation that can be measured using an 

optical densitometer. (See nuclear emulsion.)

film dosimeter

A small packet of film within a holder that attaches to a wearer.

flux density (n/cm2

-sec)

A measure of the intensity of neutron radiation in neutrons/cm2

-sec. It is the number of 

neutrons passing through 1 square centimeter of a given target in 1 second.

gamma rays (G or γ)

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Physically, gamma 

rays are identical to X-rays of high energy, the only essential difference being that X-rays do 

not originate in the nucleus.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 23 of 24

minimum detection level (MDL)

The minimum quantifiable dose equivalent that can be reliably measured with a given 

dosimetry system.

missed dose

The potential dose equivalent that might not have been measured due to the limitation of the 

dosimeter, even though a worker was monitored.

neutron

A basic particle that is electrically neutral, having nearly the same mass as the hydrogen atom.

neutron film dosimeter

A film dosimeter that contains a Neutron Track Emulsion, type A, film packet.

nuclear emulsion

Often referred to as “NTA” film and used to measure personnel dose from neutron radiation.

Nuclear Track Emulsion, Type A (NTA)

A film that is sensitive to fast neutrons. The developed image has tracks caused by neutrons 

that can be seen by using an appropriate imaging capability such as oil immersion and a 

1000X power microscope or a projection capability.

personal dose equivalent Hp(d)

Represents the dose equivalent in soft tissue below a specified point on the body at an 

appropriate depth d. The depths selected for personnel dosimetry are 0.07 mm and 10 mm, 

respectively, for the skin and body. These are noted as Hp(0.07) and Hp(10), respectively.

photon

A unit or "particle" of electromagnetic radiation consisting of X- and/or gamma rays.

rad

The traditional unit of absorbed dose (one rad = 100 ergs per gram of material absorbing the 

radiation energy).

radiation

Alpha, beta, neutron, and photon radiation.

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

rem

The traditional unit of dose equivalent, which is equal to the product of the absorbed dose in 

rad and the quality factor of the radiation.

roentgen (R)

A unit of exposure to gamma (or X-ray) radiation. It is defined precisely as the quantity of 

gamma (or X-) rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 kg of dry air 

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

tissue for higher (~>100 keV) energy photons.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-6 Page 24 of 24

recycled uranium (RU)

Uranium recovered from used reactor fuel

shallow absorbed dose (Ds) 

The absorbed dose at a depth of 0.07 mm in a material of specified geometry and 

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 mm in tissue.

sievert (Sv)

The SI unit for dose equivalent. (1 Sv = 100 rem.)

thermoluminescence

Property of a material that causes it to emit light as a result of being excited by heat.

thermoluminescent dosimeter (TLD)

A device used to measure radiation dose. It is comprised of a holder containing solid chips of 

material that when heated will release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose.

unmonitored dose

The potential unrecorded dose equivalent that might have resulted because an exposed 

worker was not monitored.

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1,000 mg/cm2); 

however, this term is also used to refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin or a radiograph.