ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 59

Document Title:

Portsmouth Gaseous Diffusion Plant –

Occupational External Dose

Document Number: ORAUT-TKBS-0015-6

Revision: 01

Effective Date: 11/07/2006

Type of Document: TBD

Supersedes: Revision 00

Subject Expert: Paul J. Demopoulos

Approval: Signature on File Approval Date: 10/26/2006

Paul J. Demopoulos, Document Owner

Approval: Signature on File Approval Date: 10/30/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 10/26/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 11/02/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 11/07/2006

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-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 2 of 59

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

01/18/2005 00 Technical Basis Document for Portsmouth Gaseous Diffusion Plant –

Occupational External Dose. First approved issue. Initiated by Mark 

D. Notich.

11/07/2006 01 Approved Revision 01 to incorporate worker outreach comments

addressed in CT-0189. Amends Section 6.3.2.2 based on worker

outreach comments. Adds required language to the Introduction. 

Adds purpose and scope subsections to the Introduction. Adds four 

references. Incorporates formal internal, NIOSH and DOL review 

comments. Table 6-37 has been changed to include footnote 

citations. Constitutes a total rewrite of the document. Sections 6.5, 

6.5.1, and 6.5.2 and the reference section were changed to include 

ORAUT-OTIB-0040, Rev. 00, “External Coworker Dosimetry Data for 

the Portsmouth Gaseous Diffusion Plant” coworker dosimetry data 

information for dose reconstruction. ORAUT-OTIB-0017, Rev. 01, 

and ORAUT-OTIB-0040, Rev. 00 citations were also added. This 

revision results in an increase in assigned dose and a PER is 

required. Initiated by Paul J. Demopoulos. Training required: As 

determined by the Task Manger. Initiated by Paul J. Demopoulos.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 3 of 59

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ...............................................................................................................7

6.1 Introduction ..............................................................................................................................9

6.1.1 Purpose......................................................................................................................10

6.1.2 Scope.........................................................................................................................11

6.2 Basis of Comparison..............................................................................................................11

6.3 Dose Reconstruction Parameters...........................................................................................11

6.3.1 PORTS Historic Administrative Practices.................................................................... 12

6.3.2 PORTS Dosimetry Technology................................................................................... 13

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

6.3.2.2 Neutron Dosimeters ........................................................................................ 18

6.3.3 Calibration ..................................................................................................................22

6.3.3.1 Film Badges Beta/Photon ............................................................................... 22

6.3.3.2 TLD Badges Beta/Gamma .............................................................................. 24

6.3.3.3 TLD Badges Albedo Neutron .......................................................................... 24

6.3.4 Workplace Radiation Fields ........................................................................................ 25

6.3.4.1 Workplace Beta/Photon Dosimeter Response ................................................ 28

6.3.4.2 PORTS Workplace Neutron Response ........................................................... 28

6.3.4.3 PORTS Workplace Neutron Dosimeter Response .......................................... 34

6.3.5 PORTS Workplace Dose Uncertainty ......................................................................... 34

6.4 Adjustments to Recorded Dose..............................................................................................35

6.4.1 Beta Dose Adjustments .............................................................................................. 35

6.4.2 Photon Dose Adjustments .......................................................................................... 35

6.4.3 Neutron Dose Adjustments......................................................................................... 35

6.5 Missed and Unmonitored Dose ..............................................................................................37

6.5.1 Missed and Unmonitored Shallow Dose ..................................................................... 37

6.5.2 Missed and Unmonitored Photon Dose ...................................................................... 43

6.5.3 Missed and Unmonitored Neutron Dose ..................................................................... 47

6.6 Organ Dose............................................................................................................................47

6.6.1 Organ Dose Conversion Factors ................................................................................ 47

6.7 Dose Reconstruction..............................................................................................................48

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

Glossary ............................................................................................................................................53
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LIST OF TABLES

TABLE TITLE PAGE

6-1 Film badge meter calendar......................................................................................... 13

6-2 Security badge letter prefix ......................................................................................... 13

6-3 PORTS dosimeter type, period of use, exchange frequency, LOD, and 

potential annual missed DE (rem)............................................................................... 13

6-4 PORTS historical dosimetry events ............................................................................ 15

6-5 IARC testing results for US beta/photon dosimeters ................................................... 17

6-6 Testing results for Hanford two-element and multielement film 

dosimeters for energy and angular response.............................................................. 17

6-7 Harshaw Type L card element and filter description ................................................... 18

6-8 Personal neutron dosimetry results November 1996 to February 1997....................... 19

6-9 Area neutron dosimetry results November 1996 to February 1997 ............................. 20

6-10 Neutron and gamma results for six full and four empty 5-in. product 

cylinders stored in the X-326 Product Withdrawal Vault, February 14, 

1986. ..........................................................................................................................20

6-11 Neutron survey results for full 5-in., 2.5- and 10-ton product cylinders 

stored in various locations, March 26, 1987................................................................ 20

6-12 Roentgen-to-rem conversion factors used for TLD dosimeter calibration.................... 24

6-13 Laboratory sources of uncertainty for beta/photon dosimeter calibration 

parameters ................................................................................................................. 25

6-14 Laboratory sources of uncertainty for neutron dosimeter calibration 

parameters ................................................................................................................. 26

6-15 PORTS workplace beta/photon dosimeter response .................................................. 26

6-16 Major facilities at PORTS where 99Tc might have accumulated .................................. 27

6-17 Reactor returns fed to cascade................................................................................... 28

6-18 Properties of radionuclides that might be found at uranium facilities........................... 29

6-19 Common workplace beta/photon dosimeter Hp(10) performance ............................... 30

6-20 Dose fractions for PORTS calibration facility .............................................................. 31

6-21 Dose fractions for PORTS nondestructive assay laboratory facility in 

Building 710................................................................................................................32

6-22 Dose fractions for PORTS HEU storage vault in Building 345..................................... 33

6-23 PORTS neutron-to-photon DE ratios .......................................................................... 34

6-24 Typical workplace neutron dosimeter Hp(10) performance......................................... 34

6-25 Estimates of uncertainty ............................................................................................. 35

6-26 Adjustments to reported PORTS deep photon dose ................................................... 35

6-27 Historical neutron quality or weighting factors............................................................. 36

6-28 PORTS facility neutron dose fractions and associated ICRP (1991) 

correction factors........................................................................................................ 37

6-29 Missed beta dose according to dosimeter type ........................................................... 38

6-30 Reported SDE dose by year....................................................................................... 39

6-31 Departmental SDE dose ratios ................................................................................... 40

6-32 Beta contamination plant limits for PORTS................................................................. 42

6-33 SDE rates for 99Tc ...................................................................................................... 43

6-34 Missed photon dose adjustments to recorded deep dose according to 

PORTS facility ............................................................................................................ 44

6-35 Reported gamma, photon, or DDE dose by year ........................................................ 45

6-35 Reported gamma, photon, or DDE dose by year ........................................................ 46

6-36 Departmental DDE dose ratios ................................................................................... 46
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6-37 PORTS dosimeter type, period of use, exchange frequency, LOD, and 

potential annual dose missed ..................................................................................... 47
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LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Measured Hanford two-element dosimeter photon response 

characteristics ............................................................................................................ 23

6-2 Results of neutron spectrum measurements made at 1 m from bare 

252Cf fission neutron source ........................................................................................ 31

6-3 Results of neutron spectrum measurements made in the front of shuffler 

unit .............................................................................................................................32

6-4 Results of neutron spectrum measurements made about 24 in. in front 

of 93% to 96% HEU cylinders..................................................................................... 33

6-5 PORTS reported SDE average and maximum dose by year ...................................... 40

6-6 PORTS reported DDE average and maximum dose by year ...................................... 44
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AP anterior-posterior (X-ray view)

BJC Bechtel Jacobs Company

cm centimeter

cpm counts per minute

DDE deep dose equivalent

DE dose equivalent

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

EU enriched uranium

ft foot

g gram

GAT Goodyear Atomic Corporation

GM Geiger-Mueller detector

HEU highly enriched uranium

HP health physics

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

hr hour

IARC International Agency for Research on Cancer

ICN International Chemical and Nuclear Corporation

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

in. inch

keV kiloelectron-volt, 1,000 electron-volts

kV kilovolt

LAW low-assay withdrawal

LOD limit of detection

LODR limit of dose rate

mCi millicurie

MDL minimum detectable level

MED Manhattan Engineer District

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mm millimeter

mo month

mR milliroentgen
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mrem millirem

MTU metric tons of uranium

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

NVLAP National Laboratory Accreditation Program

ORNL Oak Ridge National Laboratory

OW open window (i.e., no filter) nonpenetrating dose

PGDP Paducah Gaseous Diffusion Plant

POC probability of causation

PORTS Portsmouth Gaseous Diffusion Plant

QF quality factor

R roentgen

rem radiation equivalent man

rep radiation equivalent physical

RN radionuclide

RU recycled uranium

SDE shallow dose equivalent

SRS Savannah River Site

TBD technical basis document

TEPC Tissue Equivalent Proportional Counter

TLD thermoluminescent dosimeter

TLND thermoluminescent neutron dosimeter

U.S.C. United States Code

USEC United States Enrichment Corporation

wk week

yr year

α alpha particle

μCi microcurie

μg microgram

§ section or sections
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6.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

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

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

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

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

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

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

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

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

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

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

derived:

] guidelines established under 

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

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

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

• Radiation from naturally occurring radon present in conventional structures

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

 1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 10 of 59

Radiation dosimetry practices at the Portsmouth Gaseous Diffusion Plant (PORTS) were based 

initially on experience gained during several decades of radium and X-ray medical diagnostic and 

therapy applications. These methods were well advanced at the start of the Manhattan Engineer 

District (MED) program to develop nuclear weapons in about 1940. The primary new challenges 

encountered by MED, and later the U.S. Atomic Energy Commission (AEC), operations were to 

measure worker dose to external radiation involved:

• Comparatively large quantities of high-level radioactivity

• Mixed radiation fields involving beta, photon (gamma and X-ray), and neutron radiation with 

low, intermediate, and high energies

• Neutron radiation

From 1954 until 1986, Goodyear Atomic Corporation (GAT) operated the PORTS site. In 1986, 

Martin Marietta Energy Systems assumed responsibility for PORTS operations. The Energy Policy 

Act of 1992 transferred responsibility for the site from the U.S. Department of Energy (DOE) to a 

newly created entity, the United States Enrichment Corporation (USEC), which leased the 

nonoperational portion of PORTS to Bechtel Jacobs Company (BJC), the DOE primary contractor 

(USEC 2003). On July 1, 1993, the operational side of PORTS officially transferred to USEC (USEC 

2003), which is regulated by the U.S. Nuclear Regulatory Commission (NRC). From 1993 to 1999, 

Martin Marietta Utility Services, which became Lockheed Martin Utility Services, operated the USEC 

portion of the site as the primary contractor. In May 1999, USEC assumed control without a primary 

contractor. The remaining portion of PORTS (including cylinder storage and legacy wastes) is under 

the operation of DOE; BJC has been the primary contractor since April 1998.

PORTS employees have been exposed to gamma, beta, neutron, and X-ray radiation. Early 

employees were exposed to higher levels of radiation due to (1) recycled uranium that could have 

included transuranic elements and 99Tc, (2) highly enriched uranium, or (3) greater amounts of 

processed materials. From the beginning of operations in 1954 to 1980, PORTS monitored its 

employees for external exposures beginning with film dosimetry and, from 1981 to the present, with 

various forms of thermoluminescent dosimetry. 

Records of surveys, investigations, procedures, and facility controls are available. The types of 

instrumentation and procedures used over the history of the PORTS site have varied. The External 

Dose Reconstruction Implementation Guideline (NIOSH 2002) recognizes that personnel external 

dosimetry issued to workers is the best way to determine external dose. Methods of calibration and 

limitations of dosimetry systems have been documented with on- and offsite references, as indicated 

in this TBD. 

6.1.1

This TBD represents a specific support mechanism concerning documentation of external dosimetry 

historical practices at the PORTS plant. This external dose TBD can be used to evaluate external 

dosimetry data for monitored workers and can serve as a supplement to individual monitoring data. 

For unmonitored workers, this document presents information that can provide for estimations of 

external doses. This document provides a site profile of PORTS that contains technical basis 

information to be used to evaluate the total occupational external radiation dose for EEOICPA 

claimants.

Purpose
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6.1.2

This document provides supporting technical data to evaluate the total PORTS plant occupational 

radiation dose that can reasonably be associated with the worker’s radiation exposure. This dose 

results from exposure to external radiation sources in PORTS facilities that would be added to 

PORTS occupationally required diagnostic X-ray examinations, and to onsite environmental releases, 

if applicable, in order to determine the total external dose. Also included are techniques to estimate 

the doses that could have occurred while an employee was not monitored, was inadequately 

monitored, and whose monitoring records are incomplete or missing (i.e., missed dose), as well as 

dose that could have been missed due to analytical detection limits. Over the years new and more 

reliable scientific methods and protection measures have been deployed. The methods needed to 

account for these changes are also identified in this document.

Scope

This Occupational External Dosimetry TBD describes the external dosimetry program at PORTS. It 

discusses dose reconstruction, practices and policies at PORTS, and dosimeter types and 

technologies for measuring dose from the different types of radiation present in the work environment. 

It also discusses the specific details of the evaluation of doses measured from exposure to beta, 

gamma, and neutron radiation; sources of bias; workplace radiation field characteristics; responses to 

different beta/gamma and neutron dosimeters in the workplace fields; and adjustments to the 

recorded dose measured by these dosimeters during specific years.

6.2 BASIS OF COMPARISON

A basis of comparison for dose reconstruction 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 skin dose is noted as 

Hp(0.07). For penetrating radiation of significance to whole-body dose d = 10 mm, and whole-body 

dose is noted as Hp(10). The International Commission on Radiological Units and Measurements 

(ICRU) recommends the use of both Hp(0.07) and Hp(10) as the operational quantities to be recorded 

for radiological protection (ICRU 1993). In addition, Hp(0.07) and Hp(10) are the quantities used in 

the DOE Laboratory Accreditation Program (DOELAP), which the Department has used to accredit 

personnel dosimetry systems since the 1980s. The National Laboratory Accreditation Program 

(NVLAP), which is the NRC equivalent to DOELAP, uses the same operational quantities. 

PORTS has monitored photon (gamma) and beta radiation since 1954. Therefore, comparisons with 

similar dosimetry systems monitoring nearly equivalent radiation work environments are possible. 

PORTS operated its own film dosimetry system from 1954 to 1980 and its own thermoluminescent 

dosimetry system from 1981 to January 1, 1999. The Plant did not monitor neutron dose with 

personal dosimetry until 1992, then operated an onsite neutron dosimetry system from 1992 to 1994. 

By 1995, vendors provided dosimetry for USEC [with the International Chemical and Nuclear 

Corporation (ICN) as the vendor] and BJC [Oak Ridge National Laboratory (ORNL) with Y-12 

laboratories as the vendor]. As early as 1963, PORTS recognized that the health physics (HP) group 

did not have adequate instrumentation to monitor for neutrons. The policy was that neutron 

monitoring was not necessary (Cardarelli 1997). Pacific Northwest Laboratories conducted a neutron 

survey in 1992 (Soldat and Tanner 1992). PORTS did not routinely monitor for neutrons until 1997. 

6.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of the beta, photon (X-ray and gamma ray), and neutron radiation types, energies, and 

geometries of exposure in the workplace, and the characteristics of the PORTS dosimeter responses,

are crucial to the assessment of bias and uncertainty of the original recorded dose in relation to the 
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 12 of 59

radiation quantity Hp(10). Earlier dosimetry systems can be compared to current systems to evaluate 

their performance based on the premise that current systems have more stringent criteria as indicated 

in 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 available original recorded doses 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

Monitoring at PORTS included radiation level monitoring with portable and area instrumentation, use 

of pocket ionization chambers when necessary, establishment of radiation areas and high radiation 

zones, and criticality monitoring with personnel and area instrumentation along with personnel 

dosimeters. 

PORTS Historic Administrative Practices

PORTS based the frequency of personnel dosimetry on safety policy. When operations began in 

1954, personnel received dosimetry based on job assignments. In 1960, personnel dosimetry was 

assigned to all employees, contractors, and visitors as a picture identification and personnel badge.

PORTS analyzed all badges assigned to radiation workers, but not all visitor or contractor badges 

were analyzed. In 1981, personnel dosimetry changed from film dosimeters to thermoluminescent 

dosimeters (TLDs) with exchange frequencies varying from monthly to quarterly based on work 

assignments. As indicated below, a number of frequency and monitoring selection changes have 

occurred over time at PORTS (GAT 1964). 

One administrative tool PORTS used to maintain control of film badge and later thermoluminescent 

dosimeter (TLD) badges was a color code. Badges for selected departments used the color codes 

listed in Table 6-1. This was first listed in the film badge procedure dated May 3, 1963 (GAT 1963), 

and continued until PORTS used outside personal dosimetry vendors (January 1, 1999, for USEC 

employees and January 1, 1997, for BJC employees for neutron monitoring).

Table 6-2 lists letter prefixes used for badge inserts (GAT 1971); these codes might occur in 

dosimetry records or databases.

In another administrative method, PORTS places badges that cannot be assigned to personnel, 

visitors, or contractors in the “bucket” file. In addition, since at least December 29, 1969, PORTS has 
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processed badges at random for employees not among the normally selected employees at a rate of 

100 per quarter for spot checks; the Plant records positive results only for readings greater than the 

Table 6-1. Film badge meter 

calendar (GAT 1971).

Insert colors Quarter

Black – Red 1

Blue – Green 2

Black – Red 3

Blue – Green 4

Table 6-2. Security badge letter prefixa.

Badge HP insert Personnel

A A Akron GT&R officials

PM AO AEC personnel

SO SO Federal Bureau of Investigation personnel

B B Nationwide and all food vendors

C C OVEC

CC CO Construction contractors

F F All temporary visitors and employee lost badges, etc.

J J Miscellaneous (IBM, General Telephone, etc.)

H H Hurst dosimeters (Criticality Dosimeters)

E E Equipment and emergency badges

a. No prefix was used for PORTS employees.

limit of detection (LOD) of 30 mrem (GAT 1971). Selected employees were determined by 

department, job category, or job assignments. Anyone likely to exceed 10% of the contemporary 

regulatory limits probably received a permanent badge. This policy continued from the film dosimetry 

program though the TLD program until it ended on January 1, 1999.

On occasion, PORTS has assigned extremity dosimeters to workers based on their job categories 

and potential to exceed 10% of the DOE/NRC regulatory limit. [The LOD is about 30 mrem. This is 

the same as the minimum detectable level (MDL).] Table 6-3 lists the dosimeters used at PORTS.

6.3.2

PORTS maintained onsite personnel dosimetry from 1954 to 1998. The dosimetry section followed 

operational and technical guidelines, as indicated in the available procedural information. As listed in 

Table 6-4, which summarizes major events in the PORTS personnel dosimetry program, the program 

was dynamic. Changes occurred due to changes in dosimetry technology, regulatory guidance, and 

plant operations. 

PORTS Dosimetry Technology

In the mid-1990s, an internal investigation indicated that PORTS did not keep dosimetry records in 

accordance with procedures, which resulted in improper assignment of doses. From 1993 to 1995, 

some employee exposures were recorded as zero for exposed damaged TLDs. The HP group 

reconstructed these records, which resulted in minor adjustments. All doses were much less than 

DOE or NRC limits (DOE 2002a, p. 37).
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Table 6-3. PORTS dosimeter type, period of use, exchange frequency, LOD, and potential annual missed DE (rem).

Dosimeter Period of use Exchange frequency

Laboratory 

LOD

Maximum 

annual 

missed dosea

Beta/photon dosimeters

PORTS film 2-elementb 9/22/54–7/16/57 Weekly (n=52) {selected groups} 0.03c 0.78

7/17/57–9/30/58 Biweekly (n=26) {selected groups} 0.03c 0.39

10/01/58–4/8/59 Weekly (n=52) {chemical operators and material handlers}

Monthly (n=12) {remainder of selected groups}

0.03c

0.03c

0.78

0.18

4/9/59–7/31/60 Every 4 wk (n=13) {all selected groups} 0.03c 0.195

8/1/60–7/5/64 Monthly (n=12) {all selected groups}

Quarterly (n=4) {all other employees}

0.03c

0.03c

0.18

0.06

7/6/64–12/28/69 Quarterly (n=4) {all employees} 0.03c 0.06

12/29/69–12/30/73 Quarterly (n=4) {selected employees} 0.03c 0.06

12/31/73–6/29/75 Quarterly (n=4) {selected employees}

Semiannual (n=2) {unselected employees}

0.03c

0.03c

0.06

0.06

6/30/75–12/31/80 Quarterly (n=4) {selected employees}

Monthly (n=12) {selected female employees only}

0.03c

0.03c

0.06

0.18

PORTS Harshaw 2276 4-element TLD 

without window

1/1/81–12/31/82 Monthly (n=12) {all monitored}

Quarterly (n=4) {all monitored}

0.015d

0.015

0.09

0.03

PORTS Harshaw 2276, 8000, 8800 

4-element TLD with window

1/1/83–12/31/98 {1/1/93–12/31/96 for BJC 

employees}

Quarterly (n=4) 0.010e

(0.04 SDE)

0.04

(0.08 SDE)

ICN TLD 760 1/1/99–present {USEC employees} Quarterly (n=4) 0.01f

(0.03 SDE)f

0.02

(0.06 SDE)

ORNL Panasonic 8805/8806 

4-element TLD with window

1/1/99–present {BJC employees} Quarterly (n=4) 0.01h

(.03 SDE) h

0.02

(0.06 SDE)

Neutron dosimeters

PORTS TLD albedo dosimeter {USEC 

and BJC}

1/1/1992–12/31/94 {unmoderated Cf-252 

calibrated}

Quarterly (n=4) 0.02i 0.04

1/1/95–12/31/96 {moderated Cf-252 calibrated} Quarterly (n=4) 0.02i 0.04

ICN TLD 760 {USEC} 1/1/97–present {moderated Cf-252 calibrated} Quarterly (n=4) 0.01f 0.02

Y-12 Panasonic TLND {BJC 

employees}

1/1/97–12/31/98 Quarterly (n=4) 0.01g 0.02

ORNL Panasonic TLND 8806 

4-element TLD {BJC employees}

1/1/1999–present Quarterly (n=4) 0.01h 0.02

a. Maximum annual missed dose (NIOSH 2002): For photon/beta missed dose = LOD/2 × n (frequency (p. 18); for neutron missed dose = LOD/2 × n(frequency) (p. 29).

b. Kodak personnel type 2 film with gold sandwiched with cadmium for high-energy gamma, OW with aluminum for low-energy gamma and beta. LOD for SDE and DDE are the same (0.03 rem), as the 

reporting level.

c. GAT (1963), 30 mrem reporting level for gamma and beta.

d. (Wagner 2003).

e. Bassett (1986a, p. 3).

f. ICN (2003).

g. Souleyrette (2003).

h. McMahon (2003).

i. Cardarelli (1997).
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 15 of 59

Table 6-4. PORTS historical dosimetry eventsa.

Date Description

9/22/1954 PORTS Film 2-element system started.

8/1/1960 Film badge combined with security badge (picture).

7/6/1964–

12/28/1969

All employees monitored.

1/1/1981 Harshaw 2276 TLD system/4-element enacted (without window), Sr-90 

calibration; depleted uranium beta, radium gamma (1985). In addition, 

company photo was placed over beta chip during 1981 preventing beta 

(skin) determinations for 1981.

1/1/1983 Window added to TLD badge.

4/20/1986 DOELAP accreditation process begins.

1/1/1997 Neutron monitoring begins with ICN TLD 760; beta/gamma monitoring 

continues. 

1/1/1997 BJC splits with USEC, utilizes Y-12 Panasonic TLND dosimetry.

1/1/1999 USEC ends PORTS TLD program, utilizes ICN TLD 760 or equivalent.

1/1/1999 BJC utilizes ORNL Panasonic 8805 (beta/gamma)/8806 (neutron).

See Table 6-3 for references. Based on GAT and USEC/BJC procedures.

6.3.2.1 Beta/Photon Dosimeters

6.3.2.1.1 Film Dosimetry, Two-Element, 1954-1980

The film dosimetry program began in 1954. The dosimeter description from the documentation 

obtained is cryptic. A Description of Co-Operative Work Assignments in Industrial Hygiene and 

Health Physics (Wooldridge 1964) describes the dosimeter as a “film badge with Kodak Type-2 

personal monitoring film combined with aluminum, cadmium, and gold filters for beta-gamma, low 

energy gamma, and high energy gamma radiation. There are also sulfur and gold filters for neutron 

exposures.” This document indicates a detection range from 30 to 2,000 mrem. None of the 

documentation indicated any analysis for neutron or extremity exposure during the film badge era 

(Wooldridge 1964, p. 3).

The film badge had two elements with an open window (OW) and a shield of gold sandwiched 

between cadmium to admit mostly high-energy gamma. To measure high-energy gamma deep dose 

equivalent (DDE), the shield should have been at least 1 mm thick (Wooldridge 1964, p. 5).

There is some description of film processing. Kodak Type 2 film had two emulsions. The first was 

fast and, when developed and monitored for density, would yield results for an exposure range from 

30 to 2,000 mrem. The second, which was a fast emulsion that would measure gamma radiation from 

5 to 300 R, was used for an extended exposure range (Wooldridge 1964, p. 3).

Series of films, including a control for every 150 film badges, were developed for each batch. The 

control film was taken from the same emulsion as the series. Controls were used to determine the 

amount of natural film darkening from the normal wear cycle and to enable differences in emulsions 

and developing. 

A densitometer was used to measure the density of the control badge and then zeroed. The zeroed 

densitometer was used to read the badges in that series. If the measurement in the OW was less 

than or equal to zero, the beta, gamma, and combined beta/gamma measurements were recorded as 

zero. In addition, total exposures less than 30 mR were recorded as zero. 
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 16 of 59

If the shielded portion of the film read zero, the cause of OW darkening was believed to be beta 

exposure, so gamma exposure was recorded as zero. A calibration graph generated twice a year 

was used to determine the beta dose. 

If the shielded portion of the film read the same as the OW, or within the ratio of about 1.3 (OW to 

shielded portion) on the graph of the OW reading, all darkening was caused by gamma exposure. 

Beta exposure was recorded as zero. The densitometer reading from the shielded portion was found 

on the vertical axis of the graph and the corresponding gamma exposure was derived from the 

gamma or shielded labeled curve.

If the shielded portion of the film read greater than zero, but the ratio of the OW reading to the 

shielded portion reading was greater than the ratio on the calibration graph, the following formula was 

used to determine the darkening due to the gamma exposure:

OW – S(R) = beta (6-1)

where

OW = OW densitometer reading

S = shielded portion densitometer reading

R = ratio given on the graph

The corresponding beta exposure was derived from the graph and recorded. The gamma exposure 

was obtained by using the shielded portion densitometer reading and the curve labeled gamma. Total 

beta/gamma exposure was obtained by adding these exposure readings (GAT 1964, 1971, p. 2-6). 

PORTS used emergency and equipment badges in potentially high radiation areas to determine the 

cause of high exposure readings in personnel badges because those badges remained in the same 

locations and were exposed to the same conditions for known periods. 

One way to determine the effectiveness of an external dosimetry system is to compare similar 

systems used in comparable environments. The International Agency for Research on Cancer (IARC) 

evaluated a two-element film dosimeter used at the Hanford Site to monitor several photon energies 

in exposure orientations as a combination of anterior-posterior (AP), rotational, and isotropic 

geometries (Thierry-Chef et al. 2002). As indicated in Table 6-5, the two-element dosimetry system 

overestimated the Hp(10) for most exposure orientations, especially for lower energies. The listed 

118-keV energy is the most applicable energy for PORTS exposure potential. (Table 6-16 in Section 

6.3.4 lists radionuclide information.) 

Wilson et al. (1990) conducted another study on dosimetry at the Hanford Site. That study used only 

the AP orientation for low energies for film dosimeters (see Table 6-6). This is the only orientation 

that PORTS used for film dosimeter calibration. The lower energies of 16 and 59 keV that represent 

plutonium photons are close to the 13, 30, 53, 63, and 68 keV representative of low-energy uranium, 

uranium progeny, and contaminant photon energies present at PORTS. There are also many 

intermediate energy photons, most of which fall in the 30- to 250-keV energy bin (see Table 6-16).

6.3.2.1.2 Thermoluminescent Dosimeters, 1981-present 
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 17 of 59

PORTS Four-Element TLD, January 1, 1981, to December 31, 1998

The TLD program began in 1981 under GAT. Most of the following information is from 

Thermoluminescent Dosimeter Response and Calibration (Bassett 1986a). 

Table 6-5. IARC testing results for U.S. beta/photon dosimetersb,c.

Geometry Phantom

118 keV 208 keV 662 keV

Meana SD/Mean Meana SD/Mean Meana SD/Mean

US-2 (Hanford two-element film dosimeter)

AP Slab 3.0 2.1 1.3 1 1.0 0.8

AP Anthropomorphic 3.0 4.2 1.2 1.9 1.0 1.8

Rotational Anthropomorphic 2.2 2 1.4 3 1.2 3.2

Isotropic Anthropomorphic 1.5 4.4 1.1 1.6 1.0 2.7

US-8 (Hanford multielement film dosimeter)

AP Slab 1.0 1.5 1.0 0.8 0.8 1.7

AP Anthropomorphic 0.8 9.5 0.9 6 0.8 1.8

Rotational Anthropomorphic 1.2 1.9 1.2 17 1.1 1.8

Isotropic Anthropomorphic 1.0 3 1.2 9 1.0 2.3

US-22 (SRS multielement thermoluminescent dosimeter)

AP Slab 0.9 4.4 0.9 3.9 0.9 3.5

AP Anthropomorphic 0.8 3.1 0.9 2.1 0.9 3.9

Rotational Anthropomorphic 1.1 3.1 1.2 1.5 1.0 4.1

Isotropic Anthropomorphic 0.9 0.3 1.0 2.5 0.9 1.6

a. Ratio of recorded dose to Hp(10)

b. (Thierry-Chef et al. 2002).

c. (ORAUT 2003a or ORAUT 2003b)

Table 6-6. Testing results for Hanford two-element and multielement film dosimeters for energy and 

angular response.

a,b 

Beam

(energy, keV)

AP exposure Rotational exposure

Film dosimeters

TLD

1972–present

Film dosimeters

TLD 

1972–93

Two-element

1944–1956

Multielement

1957–1971

Two-element 

1944–1956

Multielement

1957–1971

16 0.1 0.9 N/Ac N/A N/A N/A

59 0.5 1.1 N/A N/A N/A N/A

M150 (70) 0.7 0.70 0.95 1.31 1.31 1.77

H150 (120) 1.6 0.64 0.87 3.00 1.20 1.64 

137Cs (662) 1.0 1.0 1.0 1.46 1.46 1.46

a. Divide recorded dose by table value to estimate Hp(10).

b. Based on Wilson et al (1990).

c. N/A = not applicable

PORTS used the Harshaw Type L card with three TLD-700 chips (7Li) and one TLD-600 chip (6Li). In 

general, chips 1 and 2 were used for skin dose, chip 3 for deep dose, and chip 4 for lens dose. 

Although not indicated in the calibration procedure, the TLDs were probably irradiated in the AP 

geometry. Table 6-7 lists the TLD shielding configurations.

Manufacturer performance specifications for the Harshaw Type L card for response are as follows:

10 mrem ±15% at 90% confidence for 60Co and 40 mrad ±20% at 90% confidence for natural uranium 

beta particles.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 18 of 59

The Panasonic 802D four-element TLD system, similar to that used at the Hanford and Savannah 

River Site (SRS), was evaluated. The TLD was irradiated by several photon energies in exposure 

orientations as a combination of AP, rotational, and isotropic geometries. As indicated in Table 6-5, 

the TLD dosimetry system closely estimated Hp(10) for all exposure orientations within ±20% of the 

Table 6-7. Harshaw Type L card element and filter description.

a

Chip 

position TLD type Shield density thickness (mg/cm2)

Total density 

thickness (mg/cm2)

1 TLD-700 Mylar/8, Teflon/8 16

2 TLD-700 Laminated Photo/75, Polyethylene/84, Teflon/8 167

3 TLD-600 Laminated Photo/75, Polyethylene/52, Cadmium/790, 

Gold/245, Tape and Teflon/15

1,177

4 TLD-700 Laminated Photo/75, Polyethylene/65, Aluminum/281, Tape 

and Teflon/15

436

a. (Bassett 1986a, p. 3).

expected exposure. The 118-keV listed energy (AP and rotational orientation) is the most applicable 

energy for PORTS exposure potential. A similar French Harshaw TLD with a plastic filter of 

1,000 mg/cm2 was within ±10% of the U.S. Panasonic TLD. Although the Panasonic system might 

have different filters, the overall response was similar (Thierry-Chef et al. 2002, Table 2, p. 106).

ORNL (Y-12) Four-Element TLD, January 1999 to present – BJC Employees

PORTS used TLDs in a few configurations, including this configuration with a Panasonic 8805/8806 

four-element card. It was designed to monitor beta, photon, and neutron radiation (McMahon 2003).

Commercial ICN Four-Element TLD System, January 1999 to Present – USEC Employees

PORTS implemented a commercial ICN TLD system on January 1, 1999. This system includes a 

four-chip beta/photon dosimeter and a separate neutron dosimeter. Technical characteristics are 

described at the ICN Internet site (www.ICN.com). This dosimetry is NVLAP-accredited.

6.3.2.2 Neutron Dosimeters

PORTS Four-Element TLD, January 1, 1992, to December 31, 1998

PORTS used the Harshaw Type L card with three TLD-700 chips (7

Li) and one TLD-600 chip (6Li) 

(thermal neutron sensitive) with a 235-mg/cm2 density thickness. The Plant performed neutron 

monitoring with this dosimeter but only of selected groups. From January 1, 1992, to December 31, 

1994, the PORTS thermoluminescent neutron dosimeter (TLND) system was calibrated against an 

unmoderated 252Cf neutron source. From January 1, 1995, to December 31, 1996, a moderated 252Cf 

source was used. The change was in response to a perceived over-response of area neutron 

dosimeters. 

ORNL (Y-12) Four-Element TLD, January 1999 to present – BJC Employees

PORTS used TLDs in a few configurations. This configuration has a Panasonic 8805/8806 fourelement card. It was designed to monitor beta, photon, and neutron radiation (McMahon 2003).

Commercial ICN Four-Element TLD System, January 1999 to Present – USEC Employees

PORTS implemented a commercial ICN TLD system on January 1, 1999. This system includes a 

four-chip beta/photon dosimeter and a separate neutron dosimeter. Technical characteristics are 

described at the ICN Internet site (www.ICN.com). This dosimetry is NVLAP-accredited.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 19 of 59

Despite changes in neutron dosimetry systems at PORTS, neutron sensitivities have remained 

essentially the same. 

6.3.2.2.1 PORTS UF6 Neutron Surveys

Neutron dosimetry was not used at PORTS until January 1, 1992. In 1996, at the request of several 

union representatives, NIOSH prepared a hazard evaluation report (Cardarelli 1997). This report 

referred to a survey of neutron radiation levels around 5-in. cylinders of highly enriched uranium 

(HEU) in storage in 1985. This survey resulted in measurements of 3 mrem/hr at the surface and 

0.5 mrem/hr at 1 m. Lower enrichment 10-ton storage cylinders produced radiation levels of 

0.5 mrem/hr neutron dose equivalent (DE) at the surface. Based on this information, a worker who 

spent about 3 hr/wk exposed to 0.5 mrem/hr would receive 75 mrem of neutron dose per year. 

Because this is 1.5% of the regulatory limit of 5,000 mrem/yr, the HP group deemed neutron 

personnel monitoring to be unnecessary (Cardarelli 1997). [Cylinder lot surveys measured an 

average gamma DE rate of about 22 mrem/hr at the surface and 3 to 12 mrem/hr at 1 m (Author 

unknown 1998a, p. 1; 1998b, p. 8)]. 

Actual personnel exposures monitored from November 1996 to February 1997 indicated no 

measurable neutron dose above the 20-mrem/quarter LOD or MDL for the dosimeter used 

(Table 6-8). This indicates possibly 80 mrem of neutron dose missed per year, a value close to that 

estimated by surveys conducted about 10 yr earlier, as mentioned above.

Table 6-8. Personal neutron dosimetry results November 1996 to February 1997a

.

Job title Department Buildings Comments Doseb (mrem)

Process operator 720, 730, 

740

X-333, X-330, 

X-326

Cold recovery, tails, product, lowassay withdrawal, unit operator, 

extended range product station

<=20 mrem

Security guard 151, 152 X-326, X-705, 

X-345 

Product withdrawal, rotation P-12 <=20 mrem

Chemical operator 721, 771, 

791

X-344, X-345, 

X-744G, X-326 

Recovery, cylinder lots, small parts, 

tunnel

<=20 mrem

Health physics 

technician

300 X-342, X-343, 

X-344, X-705 

HP coverage <=20 mrem

Laborer 147 Cylinder yards Paint and scrap in yards (20 hr/wk) <=20 mrem

Uranium material 

handler

791 X-344, X-745C, 

X-745E, X-744G, 

X-326

Cylinder lots, L-cage, warehouse 

storage, autoclave, shipping and 

receiving, vault

<=20 mrem

a. Cardarelli (1997).

b. LOD or MDL = 20 mrem.

PORTS determined from a study of dosimeters in the eight work areas that neutron dose would be 

12.5% of the photon DE (Cardarelli 1997, p. 8). A Hanford study yielded a 26% average neutron-tophoton dose ratio (reactor average; Fix, Wilson, and Baumgartner 1996). SRS experience indicates a 

somewhat lower ratio of 10% neutron-to-photon dose (reactor average; ORAUT 2003a, Section 

5.3.4.2.3.2). From the Guide of Good Practices for Occupational Radiological Protection in Uranium 

Facilities (DOE 2000, p. 2-19), a 0.01- to 0.2-mrem/hr neutron DE rate for natural to 5% enriched 

cylinders in cold storage is likely. A neutron DE rate of 2 to 5 mrem/hr at contact and 1 to 2 mrem/hr 

at 3 ft is likely for 97% enriched cylinders. Table 6-9 lists neutron dosimetry results for the work 

areas.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 20 of 59

Additional neutron radiation level surveys by PORTS health physics are presented in Tables 6-10 and 

6-11. One survey was conducted in 1986 and was used to estimate the potential exposure to workers 

who transported highly enriched uranium cylinders (Bassett 1986b). It was assumed that each worker 

would take 3 hr/wk, for 50 wk in a year to transport cylinders (Bassett 1986c). The 0.5 mrem/hr DE

Table 6-9. Area neutron dosimetry results November 1996 to February 1997a

.

Buildings Locations/comments Dose (mrem)

X-326 Extended range product {Dynamics St. 2)

Product withdrawal {T-57-8-3 Bed 3}

80 mrem (3 mo)

60 mrem (3 mo)

X-330 Tails

Deposit;dd-4;29AB-1

G-33;29-3-7-7 {randomly selected}

Low-assay withdrawal (LAW) 

<= 20 mremb

< =20 mremb

<=20 mremb

<=20 mremb

X-343 Typical movement (300 cylinders/mo) <= 20 mremb

X-345 Behind Phantom 5 on wall <= 20 mremb

X-705 Portal 

DOE Lot 11,200;2-3%, row 22-23 Sec 44

<= 20 mremb

420 mremb

210 mremb

710 mremb

X-745G DOE Lot; 11,200;2-3%, row 20-21 Sec 3, heel 210 mremb

510 mremb

320 mremb

a. Cardarelli (1997, pp. 21–22).

b. Measurement period of 1 mo.

Table 6-10. Neutron and gamma results for six full and four empty 5-in. product cylinders stored in 

the X-326 Product Withdrawal Vault, February 14, 1986.a

Survey Point and Radiation Type Empty Full

Surface/Neutron <= 0.5 mrem/hr <= 3.0 mrem/hr

Surface/Gamma <= 1.0 mR/hr <= 2.4 mR/hr

Surface/Neutron and Gamma <= 1.5 mrem/hr <= 5.4 mrem/hr

1 Meter/Neutron and Gamma <= 0.2 mrem/hr <= 0.5 mrem/hr

a. Bassett (1986b).

Table 6-11. Neutron survey results for full 5-in 2.5- and 10-ton product cylinders stored in various 

locations, March 26, 1987.a

Location Description

Fast 

Neutrons 

(mrem/hr)

Slow 

Neutrons 

(mrem/hr) Comments

X-342 North 

Cylinder Lot

Full 10-ton cylinder 0.47 0.34 Readings taken between 2 full 

cylinders, 1 ft from each

X-342 North 

Cylinder Lot

Full 2.5-ton cylinder 11.3 7.8 Readings taken approximately 1 ft 

from each cylinder

X-326 Full 5-in. cylinder 

HEU

- 2.5–3.0 Readings taken at contact with 

cylinder

X-345 Center Vault Full 5-in. cylinder 

HEU in overpack

- 1.0–1.5 Readings taken at contact with 

overpack

a. AHJ (1987).
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 21 of 59

rate measurement from the one meter reading of six full 5-in. cylinders was used in the calculation. 

This resulted in a 75-mrem/yr DE:

DE in mrem = 0.5 mrem/hr x 50 wk/yr x 3 hr/wk = 75 mrem/yr.

From sharing of the work, it was expected that the average dose was more likely 25 mrem/yr.

A neutron survey result was also conducted March 26, 1987, of 5-in., 2.5- and 10-ton full cylinders. 

The measurements distinguished fast from thermal neutrons as listed in Table 6-11.

Guards could have been more routinely exposed workers than previously assessed by PORTS health 

physics. Three estimates of DE (without the applied correction as discussed in the next section) are 

presented below.

Using the 11.3mrem/hr fast and 7.8-mrem/hr thermal neutrons at 1 ft from full 2.5-ton cylinders from 

Table 6-11 as a maximizing approach, one would obtain the following DE estimate: 

19.1 mrem/hr x 50 wk/yr x 3 hr/wk = 2,865 mrem/yr

Using 3.0-mrem/hr full surface neutron values from Table 6-10 as a best estimate, one would obtain 

the following DE estimate:

3.0 mrem/hr x 50 wk/yr x 3 hr/wk = 450 mrem/yr

Using 0.47 mrem/hr fast and 0.34 mrem/hr thermal neutrons at 1 ft from full 10-ton cylinders from 

Table 6-11 full as an underestimate, one would obtain the following DE estimate:

0.81 mrem/hr x 50 wk/yr x 3 hr/wk = 121.5 mrem/yr

These can be considered potential neutron doses to guards because it has been suggested that 

guards were often told to guard the cylinders for entire shifts. 

Note the above calculations are still likely to be conservative in all three cases based on:

1. Transport of cylinders per year estimate is probably high.

2. Cylinders are assumed full.

3. Surface readings or 1-ft readings were used.

As a comparison, if photon doses from ORAUT-OTIB-0040 (ORAUT 2005a) are used:

• 1954, 95th percentile: 1,736 mrem x 0.2 (neutron/photon ratio on measured personnel dose) 

= 347.2 mrem routine or high-level exposure. 

• 1954, 50th percentile: 780 mrem x 0.2 (neutron/photon ratio on measured personnel dose) 

= 156 mrem average or intermittent low-level exposure.

If photon dose is used from this TBD:

• 1954: 750 mrem (adjusted by 0.5 or LOD/2 method) x 0.2 (neutron/photon ratio on measured 

personnel dose) 

= 150 mrem
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 22 of 59

As a comparison using the PORTS Occupational Environmental Dose TBD (ORAUT 2004b):

• For cylinder yard workers, a DDE of 1,056 mrem for both neutron and (beta/gamma) for 2001 

• 267 mrem DDE neutron and beta/gamma or less before 2001, a DDE of 178 mrem/yr neutron 

and beta/gamma ambient radiation level to be assigned to unmonitored personnel 

6.3.2.2.2 Quality Factors for Neutrons

The quality factors (QFs) used historically for neutrons have changed significantly. In current 

regulations, QFs that are used to convert radiation dose (mrad) to DE (mrem) are based on 

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

current QFs from Publication 60 (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 personnel dose results 

(Table 6-8), and PORTS personnel dosimetry, an adjustment to ICRP (1991) of at most a factor of 

2 times higher would be necessary.

Average neutron energy is 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 PORTS since 1992 (for 

calibration facility personnel) and 1994 (others included) was overestimated and therefore favorable to 

claimants.

From November 1996 to February 1997, specific process areas and personnel were monitored using 

the Landauer/Neutrak ER badge (Landauer 2003). This badge combines an albedo dosimeter with a 

Columbia Resin No. 39 chip, which monitors neutrons from >30 to 50 keV up to 35 MeV. The TLD 

albedo chip monitors neutrons in the energy range of 0.5 to 100 keV. Thermal neutrons were not 

monitored due to the assumption that most neutron exposures at PORTS are from intermediate to fast 

neutrons (Cardarelli 1997).

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 similarity between the radiation fields used for calibration and those in the workplace. 

Calibration

6.3.3.1 Film Badges Beta/Photon

For PORTS film dosimeter calibration, eight badges were exposed to a 22.5-mCi 226Ra source in an

AP geometry at various distances free in air to yield zero (control), 30, 60, 100, 200, 500, 750, and 

1,000 mR of gamma exposure (GAT 1964). To correct for this free-in-air exposure [a conversion from 

roentgen (87.6 ergs/g) to rad (95 ergs/g)], ratios of the differences of rad to roentgen were taken, as 

follows: 

(87.6 – 95)/87.6 = 8.45%

In other words, if the badges had been calibrated with an anthropomorphic phantom, they would have 

received 95 ergs/g (plus scatter) rather than the 87 ergs/g they actually received. Assuming a QF of 1 
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 23 of 59

for rad-to-rem conversion, the correction factor due to not using a phantom would be –8.45% relative 

to using a phantom. 

Because 226Ra or 137Cs was used for calibration, an under-response of 8% or 3.9%, respectively,

would result due to higher energy calibration energies, in comparison to the actual lower exposure 

energies encountered in the workplace (ORAUT 2003a). A correction of 16.5% for film dosimetry 

(1954 to 1980), 12.5% for TLD dosimetry (1981 to 1986), and 4% for TLD dosimetry (1987 to present) 

is recommended.

Figure 6-1 shows a comparison of a two-element film dosimetry system to the other film dosimetry 

systems. Wilson et al. (1990) measured the AP photon energy response of the Hanford systems. As 

indicated in Figure 6-1, the dosimeter OW response shows a significant over-response to lower 

energy photon radiation. Most of the photon energy spectrum and the DE at PORTS are from the 30-

to 250-keV range where there was an over-response from the two-element film-shielded portion of the 

dosimeter (silver shield). This indicates that recorded results for the PORTS two-element film would 

overall be favorable to claimants and no corrections are needed for the response of the dosimetry to 

the radiation work environment. The under-response of the two-element film shielded portion to low 

energies (less than 50 keV) should be of little consequence because the vast majority of the photon 

energy spectrum and the DE are from photons greater than 50 keV. This is because most radiation 

work environments involve shielded or self-shielded uranium sources that allow little exposure from 

less-than 30-keV photons. In the case of open systems such as processing, recovery, and 

maintenance, the exposure to low-energy photons is more probable. In these situations, because the 

low-energy photons are monitored conservatively by the OW, it might be favorable to the claimant to 

equate the shallow dose to the deep dose if the deep dose was less than the shallow dose in the 

claimant’s records.

Figure 6-1. Measured Hanford two-element dosimeter photon response characteristics

(Wilson et al. 1990).

The nonpenetrating (OW) and penetrating (gold-cadmium filter) film responses were used to estimate 

skin dose from beta and photon radiation.

For beta calibration, six badges were exposed to depleted uranium slugs (240 mR/hr) in an AP 

geometry at various times to yield zero, 60, 120, 240, 480, and 960 mR of beta radiation exposure.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 24 of 59

The slugs were placed directly on the badges (GAT 1964). This meant that the calibration closely 

matched the work environment and no corrections for beta response are necessary. 

6.3.3.2 TLD Badges Beta/Gamma

TLD badges were calibrated in free air with a 137Cs source, National Bureau of Standards (NBS)-

filtered X-rays, and depleted uranium using slab geometry. Because different energies were used for 

calibration, exposure-to-DE conversion factors were used in the calibration algorithms, as listed in 

Table 6-12. 

Table 6-12. Roentgen-to-rem conversion factors used for TLD dosimeter 

calibrationa

.

NBS-filtered X-rays

Conversion 

factors (rem/R)

shallow (0.07 mm)

Conversion 

factors (rem/R)

deep(10 mm)

M30 (20 keV) 1.08 0.45

S60 (36 keV) 1.15 1.07

M150 (70 keV) 1.41 1.47

H150 (120 keV) 1.41 1.41

Cs-137 662 (b) 1.03

a. Bassett (1986a, p. 5).

b. This value, which depends on source geometry, should be measured.

Controls were placed at the plant entrance portals: Drive gate, X-100 lobby, X-108B gate, H-Lot gate, 

and C-Lot gate. About 16 combinations of sources were used, such as 137Cs and 20-, 36-, 70-, and 

120-keV X-rays and U-slab; U-slab and 20-, 36-, 70-, 120-keV X-rays are representative of possible 

mixed fields encountered at PORTS. 

The general form of the DDE equations used in the TLD calibration system was:

DDE = Do + D1(CR3) + D2(CR4) (6-2)

SDE = S0 + S1(CR1) + S2(CR2) + S3(CR3) + S4(CR4) (6-3)

where Di and Si are the deep and shallow multiple linear regression coefficients for their respective 

chip positions and CRi is the nanocoulomb thermoluminescent response for chips 1 to 4. 

TLD-700 chips 1 and 2 (16 and 167 mg/cm2

, respectively) are not used for DDE determination. The 

basis for dose algorithm calculations is chip proportionalities, which enables discrimination of radiation 

energy types to unknown fields (Basset 1986a, pp. 22–23).

Table 6-13 summarizes laboratory sources of uncertainty parameters in beta and photon calibrations. 

If uncertainty or bias is positive, dose reconstructors should make no change. In the case of negative 

bias, dose reconstructors should make the appropriate corrections using the values in Table 6-13. 

Uncertainty was estimated using information from PORTS procedures or performance information 

from similar dosimetry systems. 

6.3.3.3 TLD Badges Albedo Neutron

In 1992, some workers in the PORTS Radiation Calibration Facility and the Applied Nuclear 

Technology Department were placed on a routine neutron dose monitoring program. In October
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 25 of 59

1994, only workers entering a Controlled or Restricted Area were monitored. From 1992 to 1994, the 

albedo dosimeter was calibrated with an unmoderated 252Cf source, which resulted in higher doses 

than expected. In 1995, the calibration procedure was modified to use a moderated 252Cf source that 

would result in lower doses than expected due to the fast-neutron environment at PORTS (210 to

860 keV) (Cardarelli 1997).

Table 6-13. Laboratory sources of uncertainty for beta/photon dosimeter calibration parameters.

Parameter Historical description Uncertaintya Comment

In-air 

calibration

Prior to 1/1/1987, calibrations were 

performed free in air. In 1997 for 

BJC and 1999 for USEC, use of 

commercial or outside vendors for 

dosimetry began. Phantom for 

calibration introducing increase of 

response from backscatter occurred 

relative to dosimetry used before 

1997 and 1999. 

±10% Before 1987, recorded dose of record 

too low (see Section 6.3.3.1).

After 1997, recorded dose of record 

too high. Backscatter radiation from 

worker’s body is highly dependent on 

dosimeter design (ORAUT 2003b, 

Table 6-2). 

Radiation 

quantity

Before 1981, PORTS used Cs-137 

and Ra-226 for beta and photon 

beam calibration.

±5% For higher energy Cs-137 and 

Ra-226 for beta and photon beam 

calibration, this caused about 4% or 

8% under-response in recorded 

dose (Basset 1986a).

Tissue depth 

of dose

Historically, PORTS used specified 

depth of 10 mm to estimate deep 

dose. 

±5% Numerical effect of this for photon 

radiation is comparatively low. 

PORTS dosimeter designs had 

filtration density thickness of about 

1,000 mg/cm2 that would relate 

closely to 1-cm depth in tissue. 

Angular 

response

PORTS dosimeter system is 

calibrated using AP laboratory 

irradiations.

~ –25%

100 keV

Recorded dose of record likely too 

low because dosimeter response is 

usually lower at non-AP angles. 

Effect is highly dependent on 

radiation type and energy 

(Tables 6-15 and 6-18)

Environmental 

effects

Workplace heat, humidity such as 

dosimeter fading impact dosimeter 

results.

±5% Fading should have been less than 

2% for TLDs. Heat effect should 

have been much less than 1%. 

Recorded dose due to these effects 

likely too low.

a. Uncertainty estimate in recorded dose compared to Hp(10) based on judgment from Hanford dosimeter laboratory studies. 

The use of the ICN TLND, the Y-12 TLND, and the ORNL TLND from 1997 to the present should not 

present much variation because all are either NVLAP- or DOELAP-certified and all use phantoms for 

calibration and similar geometries. ICN is a large commercial vendor with references to its dosimetry 

program available at its Internet site (www.ICN.com). Table 6-14 lists possible dosimeter lab 

parameter uncertainty.

6.3.4

The PORTS radiation work environment consists of a variety of combinations of complex beta, 

gamma (photon), X-ray, neutron, and bremsstrahlung (X-ray) radiations. The worker’s job, location, 

and occurrences of incidents affect the external exposure. Worker location in relation to radiation 

sources during the course of the work history is one of the most challenging parameters to establish,

Workplace Radiation Fields
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 26 of 59

as is estimation of actual radiation source levels. Assumptions related to these parameter definitions 

need to be based on available worker information and the work environment. This section describes 

the basis of the assumptions. 

For the PORTS areas listed in Table 6-15, assumptions favorable to the claimants were used. As 

suggested in NIOSH (2002), the 30- to 250-keV energy range for photons is used if a specific gamma 

Table 6-14. Laboratory sources of uncertainty for neutron dosimeter calibration parameters.

Parameter Historical description Uncertaintya Comment

Source energy 

spectra

PORTS used unmoderated and 

moderated Cf-252 source for 

calibration. Unmoderated 

source led to increased bias.

±100% Delivered dose used in calibrating 

neutron dosimeters is uncertain. 

Uncertainty listed is favorable to 

claimants.

Radiation 

quantity

QF or spectrum used for PORTS 

by outside vendors. 

±50% Recorded dose of record likely too low 

because dosimeter values based upon 

pre ICRP 60 quality factors.

Angular 

response

PORTS TLND dosimeters 

calibrated using AP laboratory 

irradiations.

±50% Recorded dose of record likely too high

because dosimeter response is often 

higher at angles other than AP. Effect is 

highly dependent on energy. 

Environmental 

stability

TLD systems are subject to 

signal fade with time, heat, 

humidity, light, etc.

±50% Recorded dose of record likely too low

because of fading; however, this effect 

depends strongly on such routine 

dosimetry practices as when calibration 

dosimeters were irradiated. 

a. Uncertainty in recorded dose compared to Hp(10) based on judgment from laboratory studies based on ORAUT (2003b, 

Table 6-3). 

Table 6-15. PORTS workplace beta/photon dosimeter response.

Process/

buildings Description

Operations Radiation

type

Energy 

Begin End selection %

X-326 High assay withdrawal station, High Assay 

Storage Area{HEU}

1954 2001 Beta

Photon

> 15 keV

30 – 250 keV

100

100

Purge system (top and side) {EU – HEU} 1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X – 330 Tails withdrawal station (depleted uranium) 1954 2001 Betaa

Photon

> 15 keV

30 – 250 keV

>250 keVb

100

85

15

X-333 Reactor grade LAW (EU) [when used as 

backup for tails withdrawal; see X-330]

1954 2001 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-342, 342A, 

& 343

Fluorine generation (342) and fixed feed 

(EU)

1954 2001 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-344 Feed manufacturing plant 1958

1962

1962

2003

Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-345 Special Nuclear Material storage and 

HASA

1978 2003 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-700 Maintenance 1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-705c,d & 

705Ec Decontamination, cleaning and recovery, oxide conversion plant

1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-710e Analytical labs 1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

>250 keV

100

75

25

X-720d Compressor shop 1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100

75

25
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 27 of 59

X-744 Smelter for aluminum recovery 1961 1983 Beta

Photon

> 15 keV

30 – 250 keV

100

100

a. Technetium-99 is expected, especially from 1977 – 1980s.

b. Uranium-238 in equilibrium with daughters 234mPa and 234Th.

c. Expect all RNs including transuranic materials and 99Tc. 

d. Beta exposures are more probable with the treatment of wastes and opening of equipment.

e. Calibration sources such as 137Cs and 226Ra and X-ray equipment from 40 to 200 kV have been used in parts of this facility.

238U (234mPa, and 234Th) have photon energies greater than 250 keV when in equilibrium with 238U, the 

energy bin of 30 to 250 keV was allocated to 85% of the photon field, and the over 250-keV range 

was allocated to 15% of the field. Buildings X-710 and X-720 use 137Cs and 226Ra calibration sources 

that yield photons of energy greater than 250 keV. Therefore, 25% of the photon field was allocated 

to these areas. This is favorable to claimants because exposure from or usage in areas other than 

calibration facilities or structural analysis areas is limited. Bremsstrahlung could be present in the 

ambient work area but would be of a lower level than the ambient gamma radiation.

The energy bin selection for beta radiation for every area at PORTS is greater than 15 keV. Of 

special note for beta areas is Building X-330 or any area of tails with potential for exposure during 

processing. Technetium-99 exposure could have occurred during cascade maintenance, removal of 

the magnesium traps, and waste processing. 

PORTS processed recycled uranium (RU) that 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. However, some routine chemical 

processes can concentrate the impurities. The most significant impurity found in RU is the pure beta 

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

the diffusion cascade (DOE 2002a). In addition, 99Tc was 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 detect them, 

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

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

Table 6-16 lists the principal locations and periods for which recovery operations at PORTS are 

believed to have occurred (DOE 2002b).

Table 6-16. Major facilities at PORTS where 99Tc could have accumulated.

Building 

No. Name

Dates of 

operation Activities

X-326 Gaseous Diffusion Process Bldg. 1954–1991 High Assay Product 

X-330 Gaseous Diffusion Process Bldg. 1954–2001 Intermediate process & tails withdrawal

X-333 Gaseous Diffusion Process Bldg. 1954–2001 Initial enrichment & reactor product

X-344 UF6 Feed Manufacturing Plant 1958–1962 Conversion of UF4 to UF6

X-345 Special Nuclear Material Storage 1978–2003 HEU storage

X-700 Maintenance Building 1954–2003 Large component repairs

X-705 Decontamination & Cleaning Bldg. 1954–2003 Equipment wash & uranium recovery

X-705E Oxide Conversion Plant 1957–1978 Conversion of U3O8 to UF6

X-720 Compressor Shop 1954–2003 Disassembly & repair of compressors

X-744G Smelter & Aluminum Recovery 1954–1978 Recover aluminum from scrap

Source: DOE (2002b), p. 16.

Table 6-17 summarizes the RU reactor returns [also see ORAUT (2004a, Table 2.4.2-1). RU 

comprised about 1,094 MTU of the 330,000 MTU fed to the PORTS cascade (BJC 2000, p. 22). 
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 28 of 59

Table 6-17. Reactor returns fed to cascade.

Fiscal year

Amount 

fed (MTU)

Enrichment 

(% U-235) Source Remarks

1955 105.8 0.64–0.68 Paducah May–Sept. 1955

1956 54.5 0.64–0.68 Paducah

1956 293.4 0.64–0.68 Oak Ridge

1957 6.2 0.64–0.68 Paducah

1958 64.2 0.64–0.68 Paducah

1970 168.1 0.64–0.68 Paducah Oct. & Nov. 1969

1974 398.8 0.64–0.68 Paducah Jan. 1974

1974–1978 1.86 2–50 PORTS Oxide Conversion

1968–1977 0.15 78–80 Division of International Affairs

1977–1998 0.15 78–97 Babcock & Wilcox

1969–1993 0.07 78 AEC Office of Safeguards & Materials 

Management

1997–1998 1.10 56–82 France

1997–1998 0.33 80 NUMEC

TOTAL 1,094.66

Source: BJC (2000, Table 2.2.2.5-1).

6.3.4.1 Workplace Beta/Photon Dosimeter Response

All PORTS radiological work areas involve beta/photon radiation covering a wide range of energies, 

which are characterized by radionuclides in the work environment. Table 6-18 lists the properties of 

the radionuclides and the machine sources at PORTS. Information on the actual radiation 

environment can be reviewed in the Site Description TBD (ORAUT 2004a).

Table 6-19 summarizes the common beta/photon personnel dosimeter parameters important to 

Hp(10) performance in the workplace. PORTS dosimetry has made reasonable measurements of 

workplace radiation fields. The two-element dosimeter would have over-responded to the average 

photon field of about 100 keV by about 50% or more. No corrections to over-response are 

recommended. As of 1987, PORTS estimated shallow dose by adding the gamma (deep) and beta 

(shallow) doses together. This policy would be an overestimate of shallow dose. Extremity 

exposures were monitored infrequently. BJC has not monitored for extremity exposure.

6.3.4.2 PORTS Workplace Neutron Response

In general, PORTS radiation workers were exposed to ambient neutron radiation produced primarily 

from three reactions: Spontaneous fission of 235U, subcritical fission of 235U, an alpha reaction on 

fluorine from the decay of uranium [19F(α,n) 22Na] and an alpha reaction on oxygen from the decay of 

uranium [18O(α,n)21Ne]. The most likely places for neutron exposures are in storage areas or cylinder 

yards (X-345, Cylinder Lot 745), feed and withdrawal process areas (X-326, X-330, and X-333), 

calibration and laboratory assay areas where 252Cf sources were used, and areas where uranium 

deposits formed in the cascades. 

The assay laboratory in Building 710 and the storage vault for HEU in Building 345 have been 

characterized for neutron energy (Soldat and Tanner 1992). The 252Cf calibration facility in 

Building 710 energy characterizations was based on measurements of a similar facility.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 29 of 59

Table 6-18. Properties of radionuclides that might be found at uranium facilities.

a

Nuclide Half-life

Energies (MeV) and abundances 

of major radiations

Alpha Beta Gamma

Primary uranium isotopes

U-238 4.51 × 109 yr 4.15 (25%) -- 0.013 (8.8%)

4.20 (75%)

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

4.40 (57%) 0.185 (54%)

0.013 (31%)

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

4.72 (28%) 0.053 (0.20%)

4.77 (72%) 0.013 (10%)

U-236 2.34 × 107 yr 4.49 (76%) 0.013 (10%)

4.44 (24%)

U-232 72 yr 5.26 (31%) 0.013 (12%)

5.32 (69%)

Decay Products

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

0.193 (79%) 0.093 (4%)

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

1.001 (0.60%)

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

0.220 (15%) 0.084 (10%)

0.305 (40%)

Th-230 (U-234 parent) 77,000 yr 4.62 (23.4%) 0.068 (0.4%)

4.69 (76.3%) 0.012 (8.4%)

Th-228 (U-232 parent) 1.913 yr 5.34 (26.7%) 0.012 (9.6%)

5.42 (72.7%) 0.084 (1.2%)

Impurities

Tc-99 2.12 × 105 yr -- 0.294 --

Np-237 2.14 × 106 yr 4.78 (75%) 0.030 (14%)

4.65 (12%) 0.086 (14%)

0.145 (1%)

Pu-238 86.4 yr 5.50 (72%)

5.46 (28%)

Pu-239 24.4 × 104 yr 5.16 (73%)

5.14 (15%)

5.10 (12%)

Pu-240 6.6 × 103 yr 5.17 (76%)

5.12 (24%)

Pu-241 13.2 yr 0.021

Cs-137/calibration sources 30.6 yr 0.514 (95%) 0.662 (85%)

Ra-226 1,600 yr 4.60 (6%) 0.186 (3.6%)

4.78 (94%) 

Machine generated X-rays 0.07–0.2

a. In most part from Bassett (1986a).
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 30 of 59

Table 6-19. Common workplace beta/photon dosimeter Hp(10) performance.

a

 

Parameter Description Uncertaintyc Workplace biasb

Exposure 

geometry

PORTS dosimeter system 

calibrated AP lab 

irradiations. Workplace 

exposure geometries are 

highly variable.

>100 keV: 

Two–element film 

dosimeter, ~ 

+200%

Others, ±25%

<100 keV: 

Probably too low

Recorded dose probably too low because 

dosimeter response can be much lower at 

rotational and lateral angles. This effect is 

highly energy dependent. Highest doses are 

probably associated with AP geometry where 

work is performed close to radiation source. 

Effect is highly dependent on radiation 

energy.

Energy 

response

PORTS film dosimeter 

response to photon 

radiation <100 keV too low 

and >100 keV was too 

high.

Depends on 

actual field

100 keV ~ –25%

Bias in recorded dose depends on photon 

spectrum in workplace, especially for film 

dosimeter. Reasonable estimate of Hp(10)

dose is likely (Figure 6-1) (Wilson et al. 1990). 

Estimate based on 100 keV and Table 6-6.

Mixed fields PORTS dosimeters 

responded to both beta 

and photon radiation. TLD 

PORTS system was 

calibrated against number 

of different possible mixed 

fields.

Depends on 

actual field

About 16 types of mixed fields were calibrated 

under PORTS TLD system. PORTS film 

system was not calibrated against mixed 

fields. Reasonable estimate of Hp(10) dose is 

likely.

Missed dose Doses less than MDL or 

LOD are recorded as zero 

dose.

Recorded dose 

of record 

probably too low

PORTS recorded doses <MDL for all years. 

Issue is significant, primarily in earlier years 

with frequent dosimeter exchange and film 

dosimeters with higher MDLs. 

Environmental 

effects

Workplace heat, humidity 

such as dosimeter fading 

impact dosimeter results.

±5% Fading should have been less than 2% for 

TLDs. Heat effect should have been much 

less than 1%. Recorded dose due to these 

effects probably too low.

a. Judgments based on PORTS dosimeter response characteristics and workplace radiation fields.

b. Recorded dose compared to Hp(10).

c. Uncertainty based on recorded dose in comparison to Hp(10), judgment on similar dosimeter laboratory studies, and onsite procedures.

One phenomenon that occurs at gaseous diffusion plants is the formation of slow cookers, which are 

uranium deposits that can accumulate in the cascade. At a fission rate just below critical, a slight 

increase of neutron production occurs. All of these processes produce fast neutrons with energies 

usually less than 2 MeV. 

6.3.4.2.1 Calibration Laboratory in Building 710

Measurements were made of the bare 252Cf calibration source in the low-scatter room, where it was 

used with a tissue-equivalent proportional counter (TEPC) at 1 m. On the day of the measurement 

(March 12, 1992), the source was rated at 27.54 μg, corresponding to a DE rate of 64.9 mrem/hr at 

1 m. The TEPC measurements ranged from 48 to 57 mrem/hr (Soldat and Tanner 1992, p. 3.3). 

Pacific Northwest Laboratory measured the ORNL bare 252Cf calibration source for a neutron energy 

spectrum. The results are indicated by the solid lines in Figure 6-2. Table 6-20 lists the dose 

fractions for the neutron energy groups indicated by the dashed lines in Figure 6-2. The dose 

fractions for the lower (<10 keV) and intermediate (10 to 100 keV) energy neutron groups were less 

than 1% of the total dose from the measurements. Therefore, combining the lower and intermediate 

energy groups into the fast group of 0.1 to 2 MeV is a reasonable simplification of the neutron dose 

calculation.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 31 of 59

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1 10|2

Neutron Energy (MeV)

10|-7

10|-6

10|-5

10|-4

10|-3

10|-2

10|-1

Dose Equivlanet (mrem)

9.75e-5 mrem 1.21e-4 mrem

7.59e-3 mrem

2.61e-2 mrem

Figure 6-2. Results of neutron spectrum measurements made at 1 m from 

bare 252Cf fission neutron source (Soldat and Tanner 1992) (See Y-12 TBD, 

Fig. 6.3.4.2-3).

Table 6-20. Dose fractions for PORTS calibration facility.

Neutron energy group Near unshielded Cf-252 source

<10 keV 0.003

10–100 keV 0.004

0.1–2 MeV 0.224

2–20 MeV 0.769

Dose fractions favorable to claimants

0.1–2 MeV 0.23

2–20 MeV 0.77

The use of the ORNL 252Cf calibration source is reasonable because the energy spectra would be the 

same. However, neutron distribution would differ from calibration facility to facility based on scatter of 

the neutrons or room rate return. This difference should be minimal because calibration facilities are 

designed to minimize scatter and maintain consistency. This difference would probably not be more

than 20% (NCRP 1991).

6.3.4.2.2 Nondestructive Assay Laboratory in Building 710

The assay laboratory contained a californium shuffler unit and a segmented gamma scanner 

separated by a concrete wall that extended halfway down the room. Measurements of the 252Cf 

shuffler unit in its fixed-open position yielded a DE rate of 0.2 mrem/hr and an integrated dose on 

environmental dosimeters on a phantom of about 1 mrem. The multisphere measurement at the 

same location as the phantom resulted in an average neutron energy of 0.52 MeV and a DE rate of 

0.15 mrem/hr.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 32 of 59

The solid lines in Figure 6-3 show the calculated energy spectrum from the multisphere detectors 

(Bonner spheres). Table 6-21 lists the dose fractions for the neutron energy groups (indicated by the 

dashed lines in Figure 6-3). The dose fractions for the lower (<10 keV) and intermediate (10 to

100 keV) energy neutron groups were about 20% of the total dose from the measurements. 

Combining the lower and intermediate energy groups into the 0.1- to 2-MeV group is a reasonable 

simplification of the neutron dose calculation.

Figure 6-3. Results of neutron spectrum measurements made 

in the front of shuffler unit (Soldat and Tanner 1992). 

Table 6-21. Dose fractions for PORTS nondestructive assay 

laboratory facility in Building 710.

Neutron energy group Near unshielded Cf-252 source

<10 keV 0.034

10–100 keV 0.153

0.1–2 MeV 0.598

2–20 MeV 0.215

Dose fractions favorable to claimants

0.01–2 MeV 0.785

2-20 MeV 0.215

6.3.4.2.3 HEU Storage Vault in Building 345

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

phantom about 24 in. from the cylinders. The DE from the cylinders was about 0.8 mrem/hr with a 

total DE 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 DE rate of 0.5 mrem/hr. 

The solid lines in Figure 6-4 show the calculated energy spectrum from the multisphere detectors 

(Bonner spheres). Table 6-22 lists the dose fractions for the neutron energy groups (indicated by the 

dashed lines in Figure 6-4). The dose fractions for the lower (<10 keV) and intermediate (10 to 

100 keV) energy neutron groups were about 47% of the total dose from the measurements.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 33 of 59

Figure 6-4. Results of neutron spectrum measurements 

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

(Soldat and Tanner 1992).

Table 6-22. 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

Dose fractions favorable to claimants

< 10 keV 0.300

0.01–2 MeV 0.610

0.1–2 MeV 0.081

6.3.4.2.4 Neutron-to-Photon Dose Ratio

A 0.125 neutron-to-photon dose ratio was calculated in a Centers for Disease Control and Prevention 

study of PORTS from data obtained from an onsite study conducted in 1995. Eight locations at the 

site were monitored. The observed median quarterly neutron doses were at the Feed Plant (5 mrem), 

X-344 autoclave area (7 mrem), Shipping and Receiving (7 mrem), cylinder lots (7 mrem), X-345 (3 

mrem), X-744G (2 mrem), X-326 L Cage (<LOD), and the burn area (4 mrem). The mean total beta 

and gamma dose was 40 mrem. The mean neutron dose was 5 mrem; therefore, the calculated 

neutron-to-photon ratio was 0.125 (Cardarelli 1997).

A neutron dose can be calculated for all PORTS facilities with a potential for such a dose using the 

neutron-to-photon dose ratio of 0.125 applied to a fraction of the recorded photon dose. For example, 

a worker in a process facility with a recorded photon dose of 500 mrem in 1992 would be assigned a 

62.5-mrem neutron dose (500 × 0.125 = 62.5). The photon dose should be adjusted for missed dose 

before estimation of the neutron dose. Because the routine monitoring for neutron exposure began in
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 34 of 59

1997, this ratio method should be used before 1997. The LOD method can apply after 1997 for 

neutron exposures.

PORTS measurement of the process and storage areas was actually about 0.125:1. For the PORTS 

neutron-to-photon dose ratio, a default ratio of 0.2 obtained from a Paducah Gaseous Diffusion Plant 

(PGDP) cylinder survey (Meiners 1999) would be reasonable. Table 6-23 lists these values for 

PORTS facilities.

Table 6-23. PORTS neutron-to-photon DE ratios.

Facility Neutron-to-photon DE ratioa

General areas 0.20

Calibration facility 0.20

Nondestructive lab 0.20

HEU/EU storage 0.20

Process facilities 0.20

a. Multiply adjusted (i.e., for any missed dose) annual photon dose by the 

tabulated value and use this fraction times the neutron-to-photon dose ratio 

to estimate neutron dose.

6.3.4.3 PORTS Workplace Neutron Dosimeter Response

Table 6-24 summarizes typical neutron personnel dosimeter parameters important to Hp(10)

performance in the workplace. The most important parameter in relation to Hp(10) performance of 

neutron dosimeters is the difference between calibration and workplace neutron energy spectra. 

Table 6-24. Typical workplace neutron dosimeter Hp(10) performance.

a

Parameter Description Potential workplace biasb

Workplace 

neutron energy 

spectra

TLD response increases with 

decreasing neutron energy. 

Depends on workplace neutron spectra. 

Exposure 

geometry

TLD response decreases with 

increasing exposure angle.

TLD recorded dose is lower at angles other than 

AP. Effect is highly dependent on neutron 

energy. 

Missed dose Doses less than MDL recorded as zero 

dose.

Recorded dose of record is probably too low. 

Impact of missed dose is greatest in earlier years 

because of higher MDLs of neutron dosimeters. 

Environmental 

effects

Workplace environment (heat, humidity, 

etc.) fades dosimeter signal.

Recorded dose of record is probably too low.

a. Judgment based on Y-12 dosimeter response characteristics.

b. Recorded dose compared to Hp(10).

6.3.5

Uncertainty in the recorded dose is due primarily to two sources, laboratory bias and workplace 

radiation field composition and geometry. Sections 6.3.3 and 6.3.4 discuss potential effects of these 

parameters on the recorded dose. Table 6-25 lists the judged estimates of uncertainty in the recorded 

dose at PORTS based on the combination of parameters.

PORTS Workplace Dose Uncertainty
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 35 of 59

Table 6-25. Estimates of uncertainty.

Dosimeter PORTS

Laboratory 

uncertaintya

Workplace 

uncertaintyb

Beta/gamma dosimeters

Two-element film Used 1954–1980 ±20% ±50%

TLD Used 1981–2003 ±10% ±50%

Neutron dosimeters

TLND Used 1992–2003 ±25% ±50%

a. In relation to Hp(10) response of the dosimeter.

b. 95% confidence level.

6.4 ADJUSTMENTS TO RECORDED DOSE

With the exception of beta dose, adjustments to the PORTS reported doses are necessary in 

consideration of the uncertainty associated primarily with the complex workplace fields and exposure 

orientations.

6.4.1

No adjustment in recorded nonpenetrating or skin dose is recommended. PORTS incident reports 

would typically address nonroutine worker exposure to significant beta or photon radiation. The 

assessed doses in the incident reports, based on investigations conducted at the time of the incident, 

probably provide the best estimate of dose received. 

Beta Dose Adjustments 

6.4.2

Using only AP geometry, the only major dose adjustments necessary to use are corrections of 16.5% 

for film dosimetry (1954 to 1980), 12.5% for TLD dosimetry (1981 to 1986) and 4% for TLD dosimetry 

(1987 to present) as listed in Table 6-26. 

Photon Dose Adjustments

Table 6-26. Adjustments to reported PORTS deep photon dose.

Period Dosimeter Facility Adjustment to reported dose

Prior to 1/1/1981 All beta/photon 

dosimeters

All facilities Multiply reported film/TLD deep 

dose by factor of 1.165.

1/1/1981–1986 All beta/photon 

dosimeters

All facilities Multiply reported film/TLD deep 

dose by factor of 1.125.

1987–present All beta/photon 

dosimeters

All facilities Multiply reported film/TLD deep 

dose by factor of 1.04.

6.4.3

Adjustment to the neutron dose is necessary to account for the change in neutron QFs between 

historic and current scientific guidance, as described in NIOSH (2002). The QF is incorporated in the 

calibration methodology, which used flux-to-dose-rate conversion factors for varying neutron energies 

for each calibration source. Flux-to-dose-rate conversion factors were based on National Council on 

Radiation Protection (NCRP) Report 38 (NCRP 1971). This report lists both flux-to-dose-rate 

conversion factors and associated QFs that vary from 2 at energies less than 1 keV to 11 at 1 MeV. 

To convert from NCRP (1971) QFs to ICRP (1991) radiation weighting factors, a curve was fit to 

describe the neutron QFs as a function of neutron energy. The average QF for each neutron energy 

group was developed by integrating the area under the curve and dividing by the neutron energy 

range as shown in equation 6-4:

Neutron Dose Adjustments
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 36 of 59

2.0

0.1

,0.1 2.0

( )

( ) (2.0 0.1)

f

n MeV

Q E dE

Q E

Range − = −

∫

(6-4)

Table 6-27 summarizes historical changes in the QFs and the average NCRP (1971) QF for the 

neutron energy groups used in dose reconstruction. In addition, Table 6-27 lists the average QFs for 

the four neutron energy groups that encompass PORTS neutron exposures. The neutron DE

correction factor can be calculated by dividing the dose fractions from Section 6.3.4.2 for each 

neutron energy group [Df(En)] by the corresponding energy specific average (NCRP 1971) QF [ ( ) Q En ]

and then multiplying by the ICRP (1991) radiation weighting factor wR, as shown in equation 6-5. 

( ) ( ) ( )

f n

f n R

n

D E C E w

Q E

= × (6-5)

Table 6-27. Historical neutron quality or weighting factors.

Neutron energy 

(MeV)

Historical 

dosimetry 

guidelinea NCRP 38QFsb

Average QF used at 

PORTS ( ) Q En

ICRP 60 neutron 

weighting factor, 

wR

c

2.5 × 10-8 3 2 2.35 5

1 × 10-7 10 2

1 × 10-6 2

1 × 10-5 2

1 × 10-4 2

1 × 10-3 2

1 × 10-2 2.5 5.38 10

1 × 10-1 7.5 10.49 20

5 × 10-1 11

1 11

2 10 7.56 10

2.5 9

5 8

7 7

10 6.5

14 7.5

20 8 Not applicable 5

40 7

60 5.5

a. Trilateral meeting in 1949 radiation protection guidelines (Fix, Wilson, and Baumgartner 1997). 

b. Recommendations of NCRP Report 38 (NCRP 1971).

c. ICRP Publication 60 (ICRP 1991).

Table 6-28 summarizes the default neutron dose fractions by energy for PORTS work areas where 

field measurements of neutron spectra occurred, using the associated ICRP (1991) correction factors.

The neutron DE is calculated by multiplying the recorded neutron dose by the area-specific correction 

factors. For example, consider a 50-mrem recorded neutron dose to a worker at the calibration 

facility. The corrected neutron dose is 22 mrem from neutrons between 0.1 MeV and 2.0 MeV (50 × 

0.44) and 51 mrem from neutrons with energy between 2 MeV and 20 MeV (50 × 1.02). These 

adjustments should be applied to measured dose, missed dose, and dose determined based on a 

neutron-to-photon dose ratio.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 37 of 59

Table 6-28. PORTS facility neutron dose fractions and associated ICRP (1991) correction factors. 

Process Description/buildings

Operations

Neutron 

energy (MeV)

Default dose 

fraction

ICRP (1991) 

correction 

Begin End factor

Calibration Area X-710 (40 mCi Cf-252 

source)

1954 Present <0.01–2

2–20

0.23

0.77

0.44

1.02

Nondestructive 

Laboratory 

Area

X-710 (Cf-252 source) 1954 Present <0.01–2

2–20

0.78

0.22

1.50

0.28

HEU/EU Storage 

Areas

X-345 and Cylinder Yards 1954 Pre

s

e

n

t

<0.01–2

2–20

0.92

0.08

1.81

0.11

General Facilities 

including 

process areasa

Uranium enrichment, 

recovery, testing, 

maintenance, transport and 

storage areas

1954 Present <0.01–2

2–20

0.92

0.08

1.81

0.11

a. Neutron dose fraction is assumed to be that of the HEU/EU storage areas.

6.5 MISSED AND UNMONITORED DOSE

There are undoubtedly missed doses for PORTS monitored workers. Missed dose applies to workers 

who were monitored but had results below the LOD of their personal radiation monitors. For the early 

years of radiation monitoring, when relatively high detection limits were combined with short 

monitoring durations, missed doses can be significant. Methodologies for estimating missed doses 

are discussed in this section and in ORAUT-OTIB-0040, External Coworker Dosimetry Data for the 

Portsmouth Gaseous Diffusion Plant, (ORAUT 2005a).

Unmonitored dose could have occurred because workers could have had the potential of receiving 

less than 10% of the radiation protection guidelines or because they worked in uncontrolled areas,

where they were not considered radiation workers and therefore were not assigned dosimetry. The 

following sections discuss beta, photon, and neutron missed or unmonitored dose. Dose 

reconstructors should apply the adjustments to beta, photon, and neutron dosimetry discussed in 

Section 6.4 to the missed dose calculations discussed below. For these cases, dose reconstructors 

must rely on coworker data and/or population data to estimate a worker’s potential unmonitored dose. 

The methods to be used are discussed in ORAUT (2005a).

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

Missed shallow dose can occur if a zero dose was recorded for a dosimeter reading less than the 

LOD. Estimates of missed dose can be computed from personnel dose data from either before or 

after the missed dose period or from coworker data. NIOSH (2002) recommends calculating missed 

dose by multiplying LOD/2 by the number of zero dose results. The missed shallow dose for 

dosimeter results less than the LOD is particularly important for earlier years because LODs were 

higher and dosimeter exchange frequency was higher. The last column in Table 6-29 lists the 

resultant estimates of this annual missed dose for different years at PORTS. The LOD/2 method 

seems to be more conservative for most periods. The methods to be used are also discussed in 

ORAUT (2005a).

Missed and Unmonitored Shallow Dose
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 38 of 59

Table 6-29. Missed beta (or nonpenetrating) dose according to dosimeter type.

Dosimeter

LOD 

(rem) Period of use

Exchange 

frequency

Max. annual 

missed dose 

(rem)

Two-element 

film

0.03 9/22/54–7/16/57 Weekly (n=50) 0.75

7/17/57–9/30/58 Biweekly (n=25) 0.38

10/01/58–4/8/59 Weekly (n=50) 0.75

10/01/58–4/8/59 Monthly (n=12) 0.18

4/9/59–7/31/60 Every 4 wk (n=13) 0.20

8/1/60–7/5/64 Monthly (n=12) 0.18

8/1/60–7/5/64 Quarterly (n=4) 0.06

7/6/64–12/28/69 Quarterly (n=4) 0.06

12/29/6912/30/73 Quarterly (n=4) 0.06

12/31/73–6/29/75 Quarterly (n=4) 0.06

12/31/73–6/29/75 Semiannual (n=2) 0.03

6/30/75–12/31/80 Quarterly (n=4) 0.06

6/30/75–12/31/80 Monthly (n=12) 0.18

TLD 0.04 1/1/81–12/31/82 Monthly (n=12) 0.24

1/1/81–12/31/82 Quarterly (n=4) 0.08

0.03 1/1/83–12/31/98 Quarterly (n=4) 0.06

1/1/93–12/31/96 (BJC ) Quarterly (n=4) 0.06

0.03 1/1/99–present (USEC) Quarterly (n=4) 0.06

0.03 1/1/99–present Quarterly (n=4) 0.06

To calculate unmonitored dose, the dose reconstructor should consult ORAUT-OTIB-0040 (ORAUT

2005a) for instructions. Table 6-30, “Reported SDE dose by year in rem,” Table 6-31, “Departmental 

SDE dose ratios,” and Figure 6-5, “PORTS reported SDE average and maximum reported dose by 

year,” are for informational purposes only. 

Table 6-31 is based on the average accumulated skin dose of PGDP employees from 1953 to 1988. 

The ratios were calculated by dividing the average cumulative dose column by the Mechanical 

Inspection group cumulative total. Because PORTS and PGDP had similar operations, the relative 

dose in each departmental area should be similar. 

The beta dose reported from 1981 through the end of 1982 might not be as reliable because there 

was no OW for the PORTS TLD badge for this period. However, the average SDE for 1981 to 1982 is 

close to those recorded from subsequent years (as indicated in Table 6-30). ORAUT-OTIB-0017 

(ORAUT 2005b) and ORAUT-OTIB-0040 (ORAUT 2005a) contain instructions on reconstructing the 

beta or nonpenetrating dose during 1981 and 1982. Table 6-29 includes the potential missed beta 

dose for the respective periods of use, dosimeter types, LODs, and exchange frequencies.

If an individual has no recorded dose, and if it is definitely established that the individual was not a 

radiation worker, then the assigned dose is the environmental dose discussed in the occupational 

environmental dose portion of this PORTS Site Profile (ORAUT 2004b) and ORAUT-OTIB-0040

(ORAUT 2005a).
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 39 of 59

Table 6-30. Reported SDE dose by year (rem).a

Year

Number 

monitored Total SDE

Maximum 

SDE

Average 

SDE 

(Geometric 

mean)

Geometric 

standard 

deviation

1954 172 57.374 1.640 0.334 1.98

1955 903 128.187 0.867 0.142 2.17

1956 610 53.579 0.655 0.088 2.37

1957 482 77.845 2.095 0.162 3.00

1958 873 229.719 5.770 0.263 3.76

1959 866 442.237 10.075 0.511 3.60

1960 488 106.240 3.420 0.218 3.26

1961 549 113.291 4.890 0.206 3.89

1962 693 94.380 1.340 0.136 2.67

1963 640 78.012 1.510 0.122 2.94

1964 621 49.253 1.388 0.117 2.89

1965 356 37.018 1.150 0.104 2.80

1966 236 24.590 0.790 0.104 2.39

1967 141 18.606 1.135 0.132 2.52

1968 611 90.098 1.040 0.147 2.32

1969 901 101.743 1.890 0.113 3.35

1970 414 70.396 2.028 0.170 2.90

1971 167 30.015 2.385 0.180 3.03

1972 234 35.116 1.440 0.150 2.64

1973 207 36.700 2.140 0.177 2.91

1974 189 27.925 2.185 0.148 3.18

1975 285 59.540 2.060 0.209 2.67

1976 363 63.585 1.350 0.175 2.40

1977 523 71.895 1.410 0.137 2.72

1978 618 127.100 1.840 0.206 2.56

1979 393 58.970 1.120 0.150 2.37

1980 1,004 122.872 1.500 0.122 2.94

1981 587 26.203 0.535 0.045 2.89

1982 784 37.638 1.709 0.048 4.63

1983 1,092 61.048 1.749 0.056 4.38

1984 1,002 46.394 0.958 0.046 3.68

1985 1,428 45.647 0.452 0.032 3.12

1986 1,308 49.436 2.198 0.038 5.71

1987 1,147 51.602 2.160 0.045 5.27

1988 1,189 42.922 0.608 0.036 3.36

1989 1,491 53.851 0.593 0.036 3.33

1990 1,862 81.447 0.694 0.044 3.27

1991 1,509 61.066 0.517 0.040 3.00

1992 889 50.365 0.910 0.057 3.28

1993 667 41.737 1.551 0.063 3.95

1994 583 24.610 0.507 0.042 2.91

1995 2,112 67.011 1.996 0.032 5.89

1996 3,461 100.426 1.951 0.029 6.09

1997 3,394 116.483 1.281 0.034 4.75

1998 355 14.238 0.300 0.040 2.37

1999 189 9.423 0.247 0.050 1.98

2000 703 19.103 0.338 0.027 2.96

2001 182 5.479 0.199 0.030 2.25

2002 224 10.710 0.353 0.048 2.35

2003 385 23.498 0.569 0.061 2.61

a. Data provided by PORTS HP department. This dataset contains only positive dose workers (Litton 2004; Demopoulos 2004)
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 40 of 59

Figure 6-5. PORTS reported SDE average and maximum dose by year. Data 

provided by PORTS HP department (Demopoulos 2004).

Table 6-31. Departmental SDE dose ratios.

Description

Average cumulative 

dose (mrem)a

Ratio (using mechanical 

inspection employees 

as the basis)

Feed Plant Operators 15,834 30.99

Decontamination 12,369 24.21

Feed Plant Mechanics 9,767 19.11

Chemical Processors 3,794 7.42

Feed Plant Mechanics 1,968 3.85

Process Maintenance 1,954 3.82

Cascade Operators 1,824 3.57

Instrument 1,407 2.75

PEMU Decontamination 1,223 2.39

Electrical 987 1.93

Converter Shop 933 1.83

Material Term Serv. 931 1.82

Analytical Chemistry 877 1.72

Laundry 851 1.67

Converter Test 836 1.64

Equipment Maintenance 586 1.15

Equipment Control 572 1.12

Transportation Pool 557 1.09

Fabrication Shop 517 1.01

Mechanical Inspection 511 1.00

a. Adapted from PACE and University of Utah (2000, Table 7-3).
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 41 of 59

Energy Range

Although the different elements in a multielement dosimetry system can have specific energy 

exposure characteristics, this analysis cannot consider shielding and radiation scattering. Attempts 

were made with the PORTS TLD system to calibrate dosimeters based on a number of mixed field 

combinations. If the badge read close to the regression results of a particular mixed-field calibration 

and the person probably worked in an area with that expected mixed field, the dose calculations 

would use the mixed-field calibration factors. More detail of the procedure was not available. 

Skin Contamination

Skin 99Tc contaminations could have gone undetected because of the low-energy beta radiation. 

Although routine surveys involved monitoring for beta contamination, the uranium daughters 234Th and 

234mPa could have masked the presence of 99Tc. 

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

code (Durham 1992). The calculated shallow dose rate from uniform 99Tc skin contamination is 

1.6 × 10-3 mrem/hr per dpm/cm2 (Swinth 2004). Because the nuclide is difficult to remove from skin, 

the integrated shallow dose from 99Tc skin contamination can be relatively large. For example, with a 

residence half-time of 1.5 days, the dose is 8.1 × 10-2 mrem per dpm/cm2

. 

Because skin contamination events involving 99Tc could have gone undetected, it might be 

appropriate to consider an additional skin dose component for a reported shallow dose of a worker 

who might have had direct contact with 99Tc. To estimate an annual missed shallow dose in the 

absence of specific data, one must make assumptions about the number of times per year an affected 

skin region might have been contaminated and the extent of each contamination. For example, one 

might 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 

PORTS, see Table 6-32). With the residence half-time of 1.5 days assumed above, it follows that the 

annual shallow dose equivalent (SDE) would be 240 rem (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 

same 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.

Significant nonroutine worker doses, as can occur from skin contamination events, might 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.

The limit of dose rate (LODR) appears to be between 0.07 and 0.04 mrem/hr (see Table 6-32) based 

on the PORTS monitoring limits. One method of skin contamination monitoring was for workers to 

wrap their hands around a Geiger-Mueller (GM) tube. A typical GM tube is about 2% efficient for 99Tc 

beta particles. (A contemporary example is a Ludlum Model 44-7 end window GM.)

For the routine meter used from at least the 1960s through the 1980s, the Samson meter appears to 

have had an LOD of 60 cpm because this was the lowest count rate action limit. This would translate 

to a 3,000-dpm activity level, which corresponds to a 0.07-mR/hr 99Tc beta radiation level. Table 6-32 

points out that the lowest radiation field level used for an action level is 0.04 mR/hr. This is probably 

the LODR obtainable from a Cutie Pie ion chamber (which was used for beta and photon radiation 

field monitoring from 1965 to 1985).

External Beta Dose Rates

As a first method, dose reconstructors can estimate dose from probable external beta dose source 

terms. PORTS personnel have been exposed to some beta radiation levels, typically during
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 42 of 59

Table 6-32. Beta contamination plant limits for PORTS.

a

Area

Type of 

monitoring Action limit 1963

Action limit

1975

Action limit 1980 and 

1990 (Tc-beta)

Floor, work 

surfaces, etc.

Contamination

Radiation level

500 cpm (Samson 

meter)

0.15 mR/hr @ 1 in.

0.15 mR/hr @ 

1 in.

Surface-25,000 dpm 

500 cpm Samson

Removable-5000 dpm 

100 cpm

Hands Contamination

Radiation level

Hand and foot monitor 

as posted

Hands Contamination

Radiation level

(Hands wrapped around 

OW GM tube) 100 cpm

0.04 mR/hr 

0.04 mR/hr @ 

1 in.

Body Radiation level 0.08 mR/hr 3,000 dpm/100 cm2 

(Samson – 60 cpm)

Shoes Contamination 

or radiation 

level

Personal 500 cpm (hand 

and foot monitor)

Issued 5,000 cpm

0.15 mR/hr

1.5 mR/hr

Issued 400,000 

dpm/100 cm2

Clothing, tables, 

equipment, tools, 

etc.

Radiation level 0.15 mR/hr Coveralls, gloves200,000 dpm/100 

cm2

Personal clothing –

9,000 dpm/100 cm2

Airborne activity Contamination

Radiation level

20 cpm/ft3 (filter paper 

wrapped around GM 

tube)

2,000 dpm/ft3

Hand tools and 

other equipment

Fixed

Removable

25,000 dpm/100 cm2

3,000 dpm/100 cm2

Process shop and 

purge equipment

Fixed

Removable

200,000 dpm/100 cm2

3,000 dpm/100 cm2

a. PORTS (1963, 1975); GAT (1990).

maintenance. For example, with 99Tc trap maintenance, a typical magnesium trap for 99Tc used for 

liquid UF6 is a 10-in.-diameter by 13.25-in.-tall cylinder with about 5,650 g of magnesium trap material. 

About 1.7 μCi of 99Tc/g of absorbent material is trapped, which leads to about 10 mCi of 99Tc in the 

trap (gaseous UF6

99Tc traps accumulate much less activity). Using the VARSKIN code, the beta 

dose rate result for 10 mCi in a trap at 30 cm (Swinth 2004) is 0.16 mrem/hr at the face. Because 

there are three traps for every change-out (PORTS) and 50 hr/yr, the resultant dose is 24 mrem/yr 

(0.16 mrem/hr × 3 traps/change-out × 50 hr/yr). This method assumes that the personnel were 

potentially exposed to these beta radiation fields even though most personnel did not perform 99Tc 

trap maintenance. In addition, this is a conservative estimate because it assumes there was no 

shielding of the magnesium trap material, the accumulation of activity at the top of the cylinder 

occurred, and the personnel contact was very close.

In addition, there are alumina traps throughout the cascade buildings. They usually accumulate 

uranium progeny and some transuranic elements (237Np), but 99Tc is possible. PORTS indicated that 

external beta fields as high as 2 R/hr have been measured around these alumina traps and that 

uranium progeny buildup is probably responsible for these fields. However, personal dosimetry would 

have monitored uranium progeny adequately.

A second method can be based on the LODR for an end geometric mean of 0.08 mrem/hr 

(Table 6-32) for the whole body. With the assumption of a maximum exposure of 2,000 hr/yr, the
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 43 of 59

exposure calculation would be 0.08 mrem/hr × 2,000 hr/yr, which is 160 mrem/yr. Another 

approximation can be made using the conservative assumption that the technetium:uranium progeny 

ratio of about 0.4 (Basset 1986a), along with the maximum exposure from ambient levels of about 0.2 

mrem/hr (Basset 1986a), would yield a result of 160 mrem (0.2 mrem/hr × 0.4 × 2,000 hr/yr = 160 

mrem/yr.)

This method might be more plausible because it represents a chronic exposure situation that was 

more common among all PORTS workers. 

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 6-33 summarizes these benchmark values for SDE rates as 

determined from VARSKIN for skin contamination and for external exposure with intervening air. 

Table 6-33. SDE rates for 99Tc.

Condition

Dose rate

(mrem/hr per 

dpm/cm2) Dose rate (mrem/hr)

Skin contamination 1.6 × 10-3 N/A

External, 10 cm air 1 × 10-4 0.08 mrem/hr lower limit of dose 

limit whole body

External, 30 cm air 1 × 10-6 0.20 mrem/hr ambient

(× 0.4 99Tc/U daughter ratio)

For method 1, the total SDE would be 264 mrem/yr (240 mrem skin dose + 24 mrem external beta 

dose). For method 2, the total SDE would be 400 mrem/yr (240 mrem skin dose + 160 mrem external 

beta dose).

6.5.2

Photon dose could have been missed if a zero dose was recorded for a dosimeter reading less than 

the LOD. Estimates of missed dose can be computed from dose before or after the missed dose 

period or from coworker data. NIOSH (2002) recommends calculating missed dose by multiplying 

LOD/2 by the number of zero dose results. The missed photon dose for dosimeter results less than 

the LOD is particularly important for earlier years because LODs and dosimeter exchange frequencies

were higher. The last column in Table 6-34 lists the resultant estimates of this annual missed dose for 

different years at PORTS. The LOD/2 method seems to be more conservative for most periods. The 

methods to be used are also discussed in ORAUT-OTIB-0040, External Coworker Dosimetry Data for 

the Portsmouth Gaseous Diffusion Plant, (ORAUT 2005a).

Missed and Unmonitored Photon Dose

Individuals with no dose recorded and if it is definitely established that the individual was not a 

radiation worker, then the assigned dose is the environmental dose discussed in the occupational 

environmental dose portion of this PORTS site profile and ORAUT-OTIB-0040 (ORAUT 2005a). 

For calculating unmonitored dose, the dose reconstructor should consult ORAUT-OTIB-0040 (ORAUT 

2005a) for instructions. Table 6-35, “Reported Gamma, Photon or DDE dose by year in rem,” Table

6-36, “Departmental DDE dose ratios” and Figure 6-6, PORTS reported DDE average and maximum 

reported dose by year” are for informational purposes only.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 44 of 59

Table 6-34. Missed photon dose adjustments to recorded deep dose according to PORTS facility.

Facility 

type Dosimeter

LOD 

(rem) Period of use Exchange frequency

Max. annual 

missed dose 

(rem)

All facilities PORTS two-element film 0.03 9/22/54–7/16/57 Weekly (n=50) 0.75

7/17/57–9/30/58 Biweekly (n=25) 0.38

10/01/58–4/8/59 Weekly (n=50) 0.75

10/01/58–4/8/59 Monthly (n=12) 0.18

4/9/59–7/31/60 Every 4 wk (n=13) 0.20

8/1/60–7/5/64 Monthly (n=12) 0.18

8/1/60–7/5/64 Quarterly (n=4) 0.06

7/6/64–12/28/69 Quarterly (n=4) 0.06

12/29/6912/30/73 Quarterly (n=4) 0.06

12/31/73–6/29/75 Quarterly (n=4) 0.06

12/31/73–6/29/75 Semiannual (n=2) 0.03

6/30/75–12/31/80 Quarterly (n=4) 0.06

6/30/75–12/31/80 Monthly (n=12) 0.18

PORTS Harshaw w/o 

window 4-element TLD

0.015 1/1/81–12/31/82 Monthly (n=12) 0.09

1/1/81–12/31/82 Quarterly (n=4) 0.03

All facilities PORTS Harshaw with 

window 4-element TLD

0.010 1/1/83–12/31/98 Quarterly (n=4) 0.02

1/1/93–12/31/96 (BJC ) Quarterly (n=4) 0.02

ICN TLD 760 0.010 1/1/99–present (USEC) Quarterly (n=4) 0.02

ORNL TLD Panasonic 0.010 1/1/99–present (BJC ) Quarterly (n=4) 0.02

Figure 6-6. PORTS reported DDE average and maximum dose by 

year. Data provided by PORTS HP department. (Demopoulos

2004)

Table 6-35 is based on the average accumulated deep dose of PGDP employees from 1953 to 1988. 

The ratios were calculated from dividing the average cumulative dose column by the Steam Plant 
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 45 of 59

group cumulative total. Because PORTS and PGDP had similar operations, the relative dose in each 

departmental area should be similar. 

Table 6-35. Reported gamma, photon, or DDE dose by year (rem).a

Year

Number 

monitored Total DDE

Maximum 

DDE

Average 

DDE

(Geo Mean)

Geometric 

Standard 

Deviation

1954 172 51.945 1.500 0.302 1.99

1955 903 112.711 0.867 0.125 2.30

1956 610 48.002 0.590 0.079 2.37

1957 482 25.805 0.355 0.054 2.24

1958 873 130.699 1.090 0.150 2.34

1959 866 296.067 5.000 0.342 3.16

1960 488 35.840 0.900 0.073 2.94

1961 549 38.691 1.740 0.070 3.97

1962 693 65.015 0.765 0.094 2.46

1963 640 69.772 0.960 0.109 2.54

1964 621 45.244 1.388 0.073 3.00

1965 356 27.710 1.150 0.078 3.17

1966 236 21.884 0.790 0.093 2.50

1967 141 16.460 0.885 0.117 2.38

1968 611 86.630 1.040 0.142 2.35

1969 901 94.836 1.890 0.105 3.46

1970 414 64.633 1.145 0.156 2.35

1971 167 22.595 1.420 0.135 2.75

1972 234 26.931 1.440 0.115 2.96

1973 207 32.415 2.140 0.157 3.07

1974 189 24.395 1.720 0.129 3.04

1975 285 46.225 1.420 0.162 2.54

1976 363 55.424 1.280 0.153 2.49

1977 523 62.290 1.410 0.119 2.89

1978 618 75.170 1.210 0.122 2.68

1979 393 22.835 0.715 0.058 2.94

1980 1,004 117.394 1.500 0.117 2.99

1981 587 26.203 0.535 0.045 2.89

1982 784 18.467 0.864 0.024 4.66

1983 1,092 20.014 0.543 0.018 4.32

1984 1,002 17.855 0.529 0.018 4.27

1985 1,428 16.765 0.452 0.012 4.75

1986 1,308 20.277 1.973 0.016 7.90

1987 1,147 27.207 0.826 0.024 4.57

1988 1,189 21.957 0.395 0.018 3.76

1989 1,491 29.255 0.348 0.020 3.41

1990 1,862 39.745 0.430 0.021 3.65

1991 1,509 24.650 0.306 0.016 3.55

1992 889 19.173 0.446 0.022 3.64
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 46 of 59

Table 6-35. Continued. Reported gamma, photon, or DDE dose by year (rem).a

Year

Number 

monitored Total DDE

Maximum 

DDE

Average 

DDE

(Geometric

Mean)

Geometric 

Standard 

Deviation

1993 667 18.992 0.428 0.028 3.22

1994 583 8.852 0.259 0.015 3.40

1995 2,112 24.181 0.821 0.011 6.37

1996 3,461 35.346 0.317 0.010 4.41

1997 3,394 30.374 0.292 0.009 4.45

1998 355 10.182 0.285 0.029 2.67

1999 189 6.406 0.183 0.034 2.06

2000 703 15.404 0.173 0.022 2.42

2001 182 4.259 0.176 0.023 2.40

2002 224 7.168 0.316 0.032 2.67

2003 385 17.590 0.496 0.046 2.77

a. Data provided by PORTS HP department. This dataset contains only positive dose workers. (Litton 2004; Demopoulos 2004)

Table 6-36. Departmental DDE dose ratios.

Description

Average cumulative 

dose in mrema

Ratio (using steam 

plant employees 

as the basis)

Plant feed operators 3,814 34.67

Decontamination 2,788 25.35

Feed Plant Mechanics 2,587 23.52

Cascade Operators 627 5.70

Chemical Operators 595 5.41

Instrument 538 4.89

Transportation Pool 371 3.37

Process Maintenance 364 3.31

Environ Control 338 3.07

Electricians 298 2.71

Mat. Term Mgr 295 2.68

PEMU Decontamination 253 2.30

Mechanical Inspection 170 1.55

Plant services 147 1.34

Converter Test 145 1.32

Feed Plant Mechanics 143 1.30

Nitrogen Plant 142 1.29

Metals Building 132 1.20

Fabrication Shops 127 1.15

Steam Plant 110 1.00

a. Adapted from PACE and University of Utah (2000, Table 7-2).

Facility/Location

Table 6-34 lists the missed photon dose at PORTS facilities. The same information is listed in 

Table 6-3, which contains the table references. The maximum annual missed dose is the LOD/2 
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 47 of 59

multiplied by the exchange frequency. The LOD is based on laboratory irradiations or from citations

from dosimetry vendors. 

6.5.3

Table 6-37 summarizes missed neutron dose for PORTS (see Table 6-3 for references). For PORTS,

neutron-to-photon dose ratio is 0.2. Table 6-23 lists these values for different PORTS facilities. The 

photon dose should be adjusted for missed dose before estimation of the neutron dose. Because 

routine monitoring for neutron exposure began in 1997, the neutron-to-photon dose ratio method 

should be used before 1997. The LOD method can apply after 1997 for neutron exposures. Dose 

reconstructors should multiply this DE by the neutron dose correction factors listed in Table 6-28 for 

different processes or buildings. 

Missed and Unmonitored Neutron Dose

Table 6-37. PORTS dosimeter type, period of use, exchange frequency, LOD, and potential annual 

dose missed.

Dosimeter Period of use

Exchange 

frequency

Laboratory 

LOD (rem)

Maximum 

annual missed 

dose (rem)a

Neutron dosimeters

PORTS TLD albedo 

dosimeter {USEC and 

BJC }

1/1/1992–12/31/94

{Unmoderated Cf-252 

calibrated}

Quarterly (n=4) 0.02a 0.04

PORTS TLD albedo 

dosimeter {USEC and 

BJC }

1/1/95–12/31/96

{Moderated Cf-252 calibrated}

Quarterly (n=4) 0.02a 0.04

ICN TLD 760 {USEC} 1/1/97–present

{Moderated Cf-252 calibrated}

Quarterly (n=4) 0.01b 0.02

Y-12 Panasonic TLND 

{BJC employees}

1/1/97–12/31/98 Quarterly (n=4) 0.01c 0.02

ORNL Panasonic TLND 

8806 four-element TLD 

{BJC employees}

1/1/1999–present Quarterly (n=4) 0.01d 0.02

a. (Wagner 2003).

b. ICN (2003).

c. Souleyrette (2003).

d. McMahon (2003).

Unmonitored neutron dose may be assigned, or it may be calculated with photon coworker data as 

presented in ORAUT-OTIB-0040 (ORAUT 2005a). The same method as described in the above 

paragraph can than be applied to the unmonitored neutron dose calculation. 

6.6 ORGAN DOSE

6.6.1

NIOSH (2002) describes the methodology to calculate the organ dose distribution for the respective 

radiation types using identified exposure geometries. The missed or unmonitored dose is to be 

estimated as described in Section 6.5. A correction of 16.5% for film dosimetry (1954 to 1980), 12.5%

for TLD dosimetry (1981 to 1986) and 4% for TLD dosimetry (1987 to present) is recommended. After 

the corrections are applied, the R-to-organ DCFs should be applied for 1954 to 1986 and the Hp(10)-

to-organ DCFs should be applied for 1987 to the present. The DCFs can be found in Appendix B of 

NIOSH (2002). 

Organ Dose Conversion Factors
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 48 of 59

6.7 DOSE RECONSTRUCTION

As much as possible, dose to individuals should be based 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:

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

considered adequate. A correction of 16.5% for film dosimetry (1954 to 1980), 12.5% for TLD 

dosimetry (1981 to 1986) and 4% for TLD dosimetry (1987 to present) is recommended. Beta 

energies are greater than 15 keV and photon energies should be considered to be in the range 

of 30 keV to 250 keV.

• Workers for whom dosimetry records provide zero beta/photon values for Hp(10) and Hp(0.07)

should have missed dose assigned on the basis of LOD/2 times the number of zeros, as 

described in Sections 6.5.1 and 6.5.2 (NIOSH 2002). 

• If an individual has no recorded dose, and if it is definitely established that the individual was 

not a radiation worker, then the assigned dose is the environmental dose discussed in the 

Occupational Environmental Dose portion of this PORTS Site Profile (ORAUT 2004b). 

• The missed dose is to be estimated as described in Section 6.5. A correction of 16.5% for film 

dosimetry (1954 to 1980), 12.5% for TLD dosimetry (1981 to 1986), and 4% for TLD dosimetry 

(1987 to present) is recommended. After the corrections are applied, the R-to-organ DCFs 

should be applied for 1954 to 1986 and the HP(10)-to-organ DCFs should be applied for 1987 

to the present. The DCFs can be found in Appendix B of NIOSH (2002). 

• Reported and missed neutron DEs should be adjusted according to Section 6.4.3 to account 

for ICRP (1991).

• Cylinder yard workers, security guards, and general workers for whom no neutron dose is 

recorded should have missed neutron DE assigned based on a neutron-to-photon dose ratio 

of 0.2 (Meiners 1999). 

• Special attention should be paid to the possibility of skin contamination incidents for workers 

involved in 99Tc recovery operations (Section 6.5.1).

• Uncertainty is discussed in Tables 6-13, 6-14, 6-19 and 6-25. 
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 49 of 59

REFERENCES

A. H. J., 1987, Neutron Survey Results, Piketon, Ohio, March 26.

Bassett, A. C., 1986a, Thermoluminescent Dosimeter Response and Calibration, GAT-S-56, Piketon, 

Ohio, April 20.

Bassett, A. C., 1986b, “External Dose Rates from Five-Inch Product Cylinders,” memorandum to E. R. 

Wagner, Piketon, Ohio, February 17.

Bassett, A. C., 1986c, “Assessment of Neutron Exposure Potential for Uranium,” memorandum to E. 

R. Wagner et al., Piketon, Ohio, February 18.

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

Ohio Site Report, BJC/PORTS-139/R1, U. S. Department of Energy, Office of Environmental 

Management, Washington, D.C.

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.

Author unknown, 1998a, Special Purpose Radiological Survey completed for Cylinder Lot 745D, June 

23. [SRDB Ref ID: 8139, p. 16]

Author unknown, 1998b, Special Purpose Radiological Survey completed for Cylinder Lot 745G, July 

9. [SRDB Ref ID: 8138]

Demopoulos, P., 2004, “PORTS and PGDP spread sheet 2004 developed by Demopoulos, P. J. May 

2004,” Oak Ridge Associated Universities, Oak Ridge, Tennessee, May. [SRDB Ref ID: 

17804]

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. 

DOE (U.S. Department of Energy), 2002a, Independent Investigation of the Portsmouth Gaseous 

Diffusion Plant, Volume 1: Past Environment, Safety, and Health Practices, Office of Oversight 

Environment, Safety, and Health, Washington, D.C.

DOE (U.S. Department of Energy), 2002b, Independent Investigation of the Portsmouth Gaseous 

Diffusion Plant, Volume 2: Current Environment, Safety and Health Practices, Office of 

Oversight Environment, Safety, and Health, Washington, D.C.

Durham, J. S., 1992, VARSKIN Mod 2 and SADDE Mod 2: Computer Codes for Assessing Skin Dose 

from Skin Contamination, NUREG/CR-5873, U.S. Nuclear Regulatory Commission, 

Washington, D.C.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 50 of 59

Fix, J. J., R. H. Wilson, and W. V. Baumgartner, 1997, Retrospective Assessment of Personnel 

Neutron Dosimetry for Workers at the Hanford Site, PNNL-11196, Pacific Northwest National 

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 Epidemiological Study,” Radiation Protection 

Dosimetry, volume 74, pp. 39–53.

GAT (Goodyear Atomic Corporation), 1963, Film Badge Procedure, Piketon, Ohio, May 3.

GAT (Goodyear Atomic Corporation), 1964, GAT Film Badge Program, Piketon, Ohio, February 26.

GAT (Goodyear Atomic Corporation), 1971, Film Badge Program, Piketon, Ohio.

ICN (International Chemical and Nuclear Corporation), 2003, Internet website, http://www.ICN.com, 

Dosimetry section, accessed October 2003. 

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

Energy and Intensity of Emissions,” Publication 38, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

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

in Radiation Protection Dosimetry, Report 51, Bethesda, Maryland.

Landauer Incorporated, 2003, Internet site, http://www.landauerinc.com/neutron.htm, accessed 

October 2.

Litton, R., 2004, “PORTS Historical Film/TLD Summary – DDE>0, SDE > 0 or SDEWB > 0”, personal 

communication, February.

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

Gaseous Diffusion Plant, Paducah, Kentucky.

McMahon, 2003, “ORNL Panasonic 8805/8806 TLD and TLND LOD,” personal communication, 

September 17.

NCRP (National Council on Radiation Protection and Measurements), 1971, Protection Against 

Neutron Radiation, Report 38, Bethesda, Maryland.

NCRP (National Council on Radiation Protection and Measurements), 1991, Calibration of Survey 

Instruments Used in Radiation Protection for the Assessment of Ionizing Radiation Fields and 

Radioactive Surface Contamination, Report 112, Bethesda, Maryland.

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

Implementation Guideline, Rev. 1, OCAS-IG-001, Office of Compensation Analysis and 

Support, Cincinnati, Ohio.

ORAUT (Oak Ridge Associated Universities Team), 2003a, Technical Basis Document for the 

Savannah River Site, ORAUT-TKBS-0003, Oak Ridge, Tennessee.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 51 of 59

ORAUT (Oak Ridge Associated Universities Team), 2003b, Technical Basis Document for the 

Hanford Site, Draft, ORAUT-TKBS-0006, Oak Ridge, Tennessee.

ORAUT (Oak Ridge Associated Universities Team), 2004a, Technical Basis Document for the 

Portsmouth Gaseous Diffusion Plant – Site Description, Draft, ORAUT-TKBS-0015-2, Oak 

Ridge, Tennessee.

ORAUT (Oak Ridge Associated Universities Team), 2004b, Technical Basis Document for the 

Portsmouth Gaseous Diffusion Plant – Occupational Environmental Dose, ORAUT-TKBS0015-2, Rev. 00, Oak Ridge, Tennessee, March 17.

ORAUT (Oak Ridge Associated Universities Team), 2005b, Technical Information BulletinInterpretation of Dosimetry Data for Assignment of Shallow Dose, ORAUT-OTIB-0017, Rev.

01, Oak Ridge, Tennessee.

ORAUT (Oak Ridge Associated Universities Team), 2005a, Technical Information Bulletin- External 

Coworker Dosimetry Data for the Portsmouth Gaseous Diffusion Plant, ORAUT-OTIB-0040, 

Oak Ridge, Tennessee.

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.

PORTS (Portsmouth Gaseous Diffusion Plant), 1963, “Acceptable Plant Limits for radiation and 

Radioactive Contamination,” Radiation Protection Manual, Portsmouth Facility, April 1.

PORTS (Portsmouth Gaseous Diffusion Plant), 1975, Section 243.5, July 1.

GAT (General Atomics Corporation), 1990, GAT/GDP-1073, volume 2, p. 4.3-35, May 10. 

Soldat, K. L., and J. E. Tanner, 1992, Neutron Dose Equivalent and Energy Spectra Measurements at 

the Portsmouth Gaseous Diffusion Plant, Pacific Northwest Laboratory, Richland, Washington.

Souleyrette, 9/12/2003 Personal communication, “Y-12 TLND LOD”.

Swinth, K. L., 200 4, E-mail correspondence dated March 5 and March 19, External Dosimetry Task 

Group. 

Thierry-Chef, I., F. Pernicka, M. Marshall, E. Cardis, and P. Andreo, 2002, “Study of a Selection of 10 

Historical Types of Dosimeter: Variation of the Response to Hp(10) with Photon Energy and 

Geometry of Exposure,” Radiation Protection Dosimetry, volume 102, number 2, pp. 101–113.

USEC (United States Enrichment Corporation), 2003, “The History of PORTS,” Internet site, 

http://www.usec.com/v2001_02/html/facilities_portsoverview.asp, accessed August 21.

Wagner, 9/2003 Personal communication, “LOD of PORTS Harshaw 2276 4-element TLD”.

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.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 52 of 59

Wooldridge, F., 1964, A Description of Co-Operative Work Assignments in Industrial Hygiene and 

Health Physics, Goodyear Atomic Corporation, Portsmouth, Ohio, June 1.

Bibliography

Bassett, A. C., T. H. Maggard, and W. S. Terry, 1985, TLD Program Processing Procedures Harshaw 

2276 Version BGXI, GAT-S-54, Goodyear Atomic Corporation, Piketon, Ohio.

Cochran, S., 2002, Radiation Calibration Facility Normal Operating Procedure, XP4-GP-RI401, p. 5, 

December 31.

GAT (Goodyear Atomic Corporation), 1963, Transportation of RAM, pp. 2-6, Piketon, Ohio, May 14.

GAT (Goodyear Atomic Corporation), 1963, Film Calibration, pp. 1-4 and appendix, Piketon, Ohio, 

June.

GAT (Goodyear Atomic Corporation), 1964, Procedure for Extending Film Badge Range, pp. 1-3, 

Piketon, Ohio, May 12.

GAT (Goodyear Atomic Corporation), 1971, General Procedure for X-raying Film Badges, Piketon, 

Ohio, April.

Hill, L. R. L., and D. J. Strom, 1993, Internal Dosimetry Technical Basis Manual for Portsmouth and 

Paducah Gaseous Diffusion Plants, PNL-8723, Battelle Memorial Institute, Pacific Northwest 

Laboratories, Richland, Washington, June.

ICRP (International Commission on Radiological Protection), 1987, “Data for Use in Protection 

Against External Radiation,” Publication 51, Annals of the ICRP, Volume 17, Pergamon Press, 

Oxford, England.

Ruggles, D. J., 1986, TLD Error Analysis, GAT-S-58, Goodyear Atomic Corporation, Piketon, Ohio, 

January 22.
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GLOSSARY

absorbed dose, D 

Amount of energy imparted by radiation to unit mass of absorbing material (100 ergs per 

gram), including tissue. The unit used prior to the use of the International System of metric 

units (SI) is the rad; the SI unit is the gray.

accreditation 

In relation to this document, recognition that a dosimeter system has passed the performance 

criteria of the DOE Laboratory Accreditation Program (DOELAP) standard (DOE 1986) in 

specified irradiation categories.

accuracy 

If a series of measurements has small systematic errors, they are said to have high accuracy. 

The accuracy is represented by the bias.

albedo dosimeter 

A TLD device that measures the thermal, intermediate, and fast neutrons that are scattered 

and moderated by the body from an incident fast neutron flux.

algorithm 

A computational procedure.

Atomic Energy Commission

Original agency established for nuclear weapons and power production; a successor to the 

Manhattan Engineering District (MED) and a predecessor to the U.S. Department of Energy 

(DOE).

backscatter 

Deflection of radiation by scattering processes through angles greater than 90 degrees, with 

respect to the original direction of motion.

beta particle 

A charged particle of very small mass emitted spontaneously from the nuclei of certain 

radioactive elements. Most (if not all) of the direct fission products emit (negative) beta 

particles. Physically, the beta particle is identical with an electron moving at high velocity.

Bonner Sphere

See Multisphere Neutron Spectrometer

bremsstrahlung 

Secondary photon or X-ray radiation produced by deceleration of charged particles passing 

through matter.

buildup 

In relation to this document, increase in flux or dose due to scattering in the medium.

collective dose equivalent

The sum of the dose equivalents of all individuals in an exposed population. Collective dose is 

expressed in units of person-rem (person-sievert).
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 54 of 59

control dosimeter

A dosimeter used to establish the dosimetry system response to radiation dose. The 

dosimeter is exposed to a known amount of radiation dose.

curie

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

deep absorbed dose (Dd)

The absorbed dose at the depth of 1.0 cm in a material of specified geometry and 

composition.

deep dose equivalent (DDE)

The dose equivalent at the respective depth of 1.0 cm in tissue.

densitometer

Instrument that has a photocell to determine the degree of darkening of developed 

photographic film.

density reading

See optical density.

dose equivalent (DE)

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 gray, H is in seivert. 

(1 sievert = 100 rem.)

DOELAP

The DOE Laboratory Accreditation Program (DOELAP) accredits DOE site dosimetry 

programs based on performance testing and onsite reviews performed on a 2- year cycle.

dose equivalent index

For many years the dose equivalent index was used to calibrate neutron sources that were 

used to calibrate neutron dosimeters. A concept of summing the maximum dose equivalent 

delivered in the ICRU sphere at any depth for the respective neutron energies was used even 

though the maximum dose occurred at different depths.

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 albedo dosimeter, film 

dosimeter, neutron film dosimeter, thermoluminescent dosimeter.)

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.

error

Term used to express the difference between the estimated and "true" value. Error can also 

be used to refer to the estimated uncertainty.
Document No. ORAUT-TKBS-0015-6 Revision No. 01 Effective Date: 11/07/2006 Page 55 of 59

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

As used in the technical sense, a measure expressed in roentgens of the ionization produced 

by gamma (or X) rays in air.

extremity

That portion of the arm extending from and including the elbow through the fingertips, and that 

portion of the leg extending from and including the knee and patella through the tips of the 

toes.

fast neutron

Neutron of energy between 10 keV and 10 MeV.

field calibration

Dosimeter calibration based on radiation types, intensity and energies present in the work 

environment.

film

Generally means 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. 

film density

See optical density.

film dosimeter

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

filter

Material used to adjust radiation response of a dosimeter to provide an improved tissue 

equivalent or dose response.

gamma rays

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. 

gamma ray interactions

Interaction of gamma rays with matter occurs through three primary processes as follows:

Photoelectric absorption - Process whereby a gamma ray (or X-ray) photon, with energy 

somewhat greater than that of the binding energy of an electron in an atom, transfers all its 

energy to the electron, which is consequently removed from the atom.

Compton scattering - Attenuation process observed for X-ray or gamma radiation in which an 

incident photon interacts with an orbital electron of an atom to produce a recoil electron and a 

scattered photon of energy less that the incident photon.
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Pair production

Absorption process for X-ray and gamma radiation in which the incident photon is annihilated 

in the vicinity of the nucleus of the absorbing atom, with subsequent production of an electron 

and positron pair. This reaction only occurs for incident photon energies that exceed 1.02 

MeV.

hurst dosimeter

Film-based criticality dosimeter.

intermediate energy neutron

Neutron of energy between 0.5 electron-volt (assumed to be 0.4 electron-volt because of 

cadmium cutoff in neutron response) and 10 kiloelectron-volts.

ionizing radiation

Electromagnetic radiation (consisting of photons) or particulate radiation (consisting of 

electrons, neutrons, protons, etc.) capable of producing charged particles through interactions 

with matter.

isotopes

Forms of the same element having identical chemical properties but differing in their atomic 

masses. Isotopes of a given element all have the same number of protons in the nucleus but 

different numbers of neutrons. Some isotopes of an element may be radioactive.

kiloelectron-volt (keV)

An amount of energy equal to 1,000 electron-volts.

luminescence

The emission of light from a material as a result of some excitation.

minimum detectable level (MDL)

Often confused because the statistical parameters necessary to its calculation are not 

explicitly defined. Nonetheless, often assumed to be the level at which a dose is detected at 

the two-sigma level (i.e., 95% of the time). The MDL should not be confused with the 

minimum recorded dose. 

minimum recorded dose

Based on a policy decision, the minimum dose level that is routinely recorded. A closely 

related concept is the dose recording interval. PORTS has generally recorded minimum 

doses of 10 mrem and at intervals of 10 mrem (10, 20, 30, etc.).

million-electron volt (MeV)

An amount of energy equal to 1,000,000 electron-volts.

multisphere neutron spectrometer

A series of neutron moderating spheres of tissue-equivalent material with a neutron detector 

positioned at the middle of the respective spheres. Algorithms are used to unfold the data to 

calculate the neutron spectra.

neutron

A basic particle that is electrically neutral weighing nearly the same as the hydrogen atom.
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neutron, fast

Neutrons with energy equal or greater than 10 keV.

neutron, intermediate

Neutrons with energy between 0.4 electron-volt and 10 kiloelectron-volts.

neutron, thermal

Strictly, neutrons in thermal equilibrium with surroundings. In general, neutrons with energy 

less than the cadmium cutoff at about 0.4 electron-volts.

open window (OW)

Designation on TLD dosimeter reports is of little use because there was no OW throughout 

1981. Otherwise, an OW existed for the film and TLD badges used at PORTS.

optical density

The quantitative measurement of photographic blackening the density defined as D = log10

(Io/I).

penetrating dose

Designation (i.e., P or Pen) on film dosimeter reports that implies a radiation dose, typically to 

the whole body, from higher energy photon radiation.

personal dose equivalent, Hp(d)

Radiation quantity recommended for use as the operational quantity to be recorded for 

radiological protection purposes by the International Commission on Radiological Units and 

Measurements. Represented by Hp(d), where d identifies the depth (in mm) 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). 

photon

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

precision

If a series of measurements has small random errors, the measurements are said to have high 

precision. The precision is represented by the standard deviation.

quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose.

rad

A unit of absorbed dose equal to the absorption of 100 ergs per gram of absorbing material, 

such as body tissue.

radiation

One or more of beta, neutron, and photon radiation. 

radiation monitoring

Routine measurements and the estimation of the dose equivalent for determining and 

controlling the dose received by workers.
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radioactivity

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

neutrons from unstable nuclei.

radionuclide

Unstable nuclides that emit radiation, generally alpha or beta particles, gamma rays, and 

neutrons to eventually form stable nuclei.

random errors

When a given measurement is repeated, the resulting values, in general, do not agree exactly. 

The causes of the disagreement between individual values must also be causes of their 

differing from the "true" value. Errors resulting from these causes are random errors.

rem

A unit of dose equivalent, which is equal to the product of the number of rads absorbed and 

the quality factor.

roentgen

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. An 

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

scattering

The diversion of radiation from its original path as a result of interactions with atoms between 

the source of the radiation and a point at some distance away. Scattered radiations are 

typically changed in direction and of lower energy than the original radiation.

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 (SDE)

Dose equivalent at a depth of 0.07 mm in tissue.

shielding

Any material or obstruction that absorbs (or attenuates) radiation and thus tends to protect 

personnel or materials from radiation.

sievert (Sv)

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

skin dose

Absorbed dose at a tissue depth of 7 mg/cm2

.

systematic errors

When a given measurement is repeated and the resulting values all differ from the "true" value 

by the same amount, the errors are systematic.

thermal neutron

Strictly, neutrons in thermal equilibrium with surroundings. In general, neutrons of energy less 

than the cadmium cutoff of about 0.4 electron-volt.
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tissue equivalent

Used to imply that the radiation response characteristics of the material being irradiated are 

equivalent to tissue. Achieving a tissue equivalent response is typically an important 

consideration in the design and fabrication of radiation measuring instruments and dosimeters.

tissue-equivalent proportional counter (TEPC)

Device used to measure the absorbed dose from neutron radiation in near tissue equivalent 

materials and, through analysis of the counter data, determination of the effective quality factor 

and the dose equivalent.

TLD chip

A small block or crystal made of LiF used in the TLD.

TLD-600 - A TLD chip made from 6

Li (>95%) used to detect neutrons.

TLD-700 - A TLD chip made from 7

Li (>99.9%) used to detect photon and beta radiation.

thermoluminescent

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

thermoluminescent dosimeter (TLD)

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. The solid 

chips are sometimes called crystals.

whole-body dose

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

); also 

used to refer to the dose recorded.

X-ray

Ionizing electromagnetic radiation of extranuclear origin.