ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 54

DOE Review Release 04/03/2014

Feed Materials Production Center –

Occupational External Dose

ORAUT-TKBS-0017-6 Rev. 01

Effective Date: 03/25/2014

Supersedes: Revision 00

ORAUT-OTIB-0073,

Revision 00

Subject Expert(s): Karen S. Kent and Matthew H. Smith

Document Owner 

Approval: Signature on File Approval Date: 03/20/2014

Karen S. Kent, Document Owner

Concurrence: Signature on File Concurrence Date: 03/24/2014

James P. Griffin, Deputy Project Director

Concurrence: Signature on File Concurrence Date: 03/20/2014

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 03/20/2014

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/25/2014

David S. Sundin, Deputy Director, DCAS

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-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 2 of 54

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

04/20/2004 00 New technical basis document for Fernald Environmental 

Management Project (FEMP) – Occupational External Dose. 

Incorporates formal internal and NIOSH review comments. First 

approved issue. Initiated by Leo G. Faust.

03/25/2014 01 Revision initiated to resolve NIOSH comments from 2006 and to 

incorporate the coworker data from ORAUT-OTIB-0073 and cancel 

that document. Added Purpose, Scope, and Special Exposure 

Cohort sections. Included additional neutron information. Updated 

and added references including but not limited to ORAUT-OTIB0017, ORAUT-OTIB-0012, ORAUT-OTIB-0020, ORAUT-OTIB-0055,

and DCAS-TIB-0013. Incorporated changes in response to formal 

internal and NIOSH review comments and Sanford Cohen & 

Associates matrix issues. Constitutes a total rewrite of the 

document. Training required: As determined by the Objective 

Manager. Initiated by Matthew H. Smith and Karen S. Kent.
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 3 of 54

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 5

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

6.1.1 Purpose........................................................................................................................ 8

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

6.1.3 Special Exposure Cohort.............................................................................................. 8

6.2 Overview of Site Operations..................................................................................................... 9

6.3 Basis of Comparison.............................................................................................................. 11

6.4 Historical Dosimetry Practices................................................................................................ 12

6.4.1 Administrative Practices ............................................................................................. 12

6.4.2 Incidents..................................................................................................................... 17

6.4.3 Dosimetry Technology................................................................................................ 18

6.4.3.1 Multielement Film Dosimeters ...................................................................... 19

6.4.3.2 Thermoluminescent Dosimeters................................................................... 19

6.4.4 Calibration .................................................................................................................. 20

6.4.4.1 Beta/Photon Dosimeters .............................................................................. 20

6.4.4.2 FMPC Beta/Photon Dosimeter ..................................................................... 21

6.4.5 Workplace Beta/Photon Radiation Fields.................................................................... 21

6.4.5.1 FMPC Beta/Photon Dosimeter Response Testing........................................ 21

6.4.5.2 FMPC Workplace Beta/Gamma Dosimeter Response ................................. 22

6.4.6 Neutron Dosimetry...................................................................................................... 24

6.4.6.1 Development of the Neutron-to-Photon Ratio............................................... 26

6.4.6.2 Neutron Activation Analysis.......................................................................... 30

6.4.6.3 Workplace Neutron Radiation Fields ............................................................ 30

6.5 Recorded Doses .................................................................................................................... 31

6.6 Adjustments to Recorded Dose.............................................................................................. 31

6.6.1 Adjustment for Neutron Quality Factor........................................................................ 32

6.6.2 Dose Conversion Factors ........................................................................................... 33

6.7 Missed Dose .......................................................................................................................... 33

6.7.1 Missed Beta/Photon Dose .......................................................................................... 33

6.7.2 Unmonitored Neutron Dose ........................................................................................ 34

6.8 Unmonitored Dose ................................................................................................................. 34

6.8.1 Coworker Assigned Photon Dose ............................................................................... 34

6.9 Organ Dose............................................................................................................................ 34

6.10 Uncertainty............................................................................................................................. 35

6.11 Geometric Correction Factor.................................................................................................. 35

6.12 Attributions and Annotations .................................................................................................. 35

References ........................................................................................................................................ 36
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 4 of 54

Glossary ............................................................................................................................................ 43

ATTACHMENT A, FERNALD COWORKER DOSE ASSIGNMENT................................................... 48

LIST OF TABLES

TABLE TITLE PAGE

6-1 Uranium beta and gamma emissions of interest..................................................................... 11

6-2 Recycled materials receipts ................................................................................................... 13

6-3 Dosimeter characteristics....................................................................................................... 13

6-4 Dosimetry calibration practices............................................................................................... 17

6-5 Lower limits of detection for DOELAP categories ................................................................... 23

6-6 Maximum viable category dose ranges .................................................................................. 24

6-7 Angular response for DOELAP energies of importance.......................................................... 25

6-8 Beta and photon energies and percentages........................................................................... 26

6-9 Measured neutron dose rates ................................................................................................ 27

6-10 Statistical parameters of neutron-to-photon ratio distributions ................................................ 28

6-11 NAA source neutron-to-photon ratios ..................................................................................... 30

6-12 Default neutron dose fractions ............................................................................................... 31

6-13 Historical recorded dose practices ......................................................................................... 31

6-14 Missed beta/photon dose ....................................................................................................... 33

6-15 Uncertainty.............................................................................................................................35

A-1 Missed external doses ...........................................................................................................51

A-2 Annual external coworker doses modified to account for missed dose................................... 53

A-3 Annual external coworker doses modified in accordance with ORAUT-OTIB-0052 ................ 54

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Uranium-235 decay series ..................................................................................................... 14

6-2 Uranium-238 decay series ..................................................................................................... 15

6-3 Thorium-232 decay series...................................................................................................... 16

6-4 Beta dose vs. skin depth ........................................................................................................ 22

6-5 Comparison of neutron-to-photon ratio distributions for depleted and low-enriched UF4......... 28

6-6 Relative percentage of uranium receipts by material category ............................................... 29

6-7 Plant 5 gamma spectrum ....................................................................................................... 32

6-8 ORNL film dosimeter response1 ............................................................................................ 32
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 5 of 54

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANSI ANSI International (formerly American National Standards Institute)

CFR Code of Federal Regulations

cm centimeter

d day

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DOELAP DOE Laboratory Accreditation Program

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

FEMP Fernald Environmental Management Project

FMPC Feed Materials Production Center

FR Federal Register

ft foot

g gram

hr hour

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiological Units

IIR incident investigation report

in. inch

INEL Idaho National Engineering Laboratory

IREP Interactive RadioEpidemiological Program

keV kilovolt-electron, 1,000 electron volts

LOD limit of detection

MDL minimum detection limit

MeV megavolt-electron, 1 million electron volts

mg milligram

mm millimeter

mR milliroentgen

mrad millirad

mrem millirem

mrep millirep

MTU metric tons of uranium

NAA neutron activation analysis

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NLO National Lead Company of Ohio

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 6 of 54

POC probability of causation

R roentgen

REF Radiation effectiveness factor

rpm revolutions per minute

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

wk week

§ section or sections
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 7 of 54

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

] 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, 42 C.F.R. Pt. 82) restrict the “performance of duty” 

referred to in 42 U.S.C. § 7384n(b) to nuclear weapons work (NIOSH 2010a).

The statute also 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 excludes Naval Nuclear Propulsion Facilities from being covered under the Act, 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 occupational radiation exposures are considered valid for inclusion in a 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 (NIOSH 2010a):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 8 of 54

6.1.1 Purpose

The purpose of this document is to describe the external dosimetry systems and practices at the Feed 

Materials Production Center (FMPC) and to assist in the evaluation of occupational external 

exposures from processes that occurred at FMPC. FMPC was also known as the Fernald

Environmental Management Project (FEMP), the Fernald Closure Project (FCP), and is now the 

Fernald Preserve. This document discusses the dosimeters FMPC used along with their exchange 

periods and other technical parameters of their use from the early 1950s to the 2000s. It provides 

supporting technical data to evaluate, with assumptions favorable to the claimant, occupational 

external doses that can reasonably be associated with radiation exposures to both monitored and 

unmonitored workers.

6.1.2 Scope

External radiation dosimetry refers to the measurement of radiation external to (i.e., outside) the body 

such as occurs with medical X-rays, cosmic rays, or radiation from naturally occurring radioactivity in 

the earth. Primary types of radiation typically significant to exposure to workers are beta, photon (i.e., 

X-ray and gamma) and neutron radiation, respectively, each with characteristic properties of origin 

and interaction with matter. Facilities that contain natural or manmade radionuclides have the 

potential for external radiation exposure of workers. External radiation dosimetry can be contrasted 

with internal radiation dosimetry, which is concerned with quantification of radiation exposure from 

radionuclides internal (i.e., inside) to the body.

This document provides a technical basis to evaluate external radiation exposure to workers that can 

reasonably be associated with Fernald operations under EEOICPA legislation. Consistent with 

NIOSH guidelines, this document identifies options to adjust historical recorded occupational external 

dose to account for current scientific methods and protection factors. The methods and concepts of 

measuring occupational external doses to workers have evolved since the beginning of Fernald 

operations. In particular, this document presents the methods to prepare worker dose information for 

input to the NIOSH Interactive RadioEpidemiological Program (IREP).

6.1.3 Special Exposure Cohort

The Secretary of the U.S. Department of Health and Human Services has designated three classes of 

employees at Fernald as additions to the Special Exposure Cohort (SEC):

All employees of DOE, its predecessor agencies, and their contractors, or 

subcontractors who worked at the Feed Materials Production Center in Fernald, Ohio, 

from January 1, 1968 through December 31, 1978, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment, or 

in combination with work days within the parameters established for one or more other 

classes of employees included in the SEC (77 FR 150; August 3, 2012). 

It was determined that NIOSH lacked the sufficient information to allow it to estimate with 

sufficient accuracy the potential internal doses from exposure to thorium, which employees at 

this facility may have been subjected (Sebelius 2012).

All employees of the DOE, its predecessor agencies, and their contractors and 

subcontractors who worked at the Feed Materials Production Center in Fernald, Ohio, 

from January 1, 1954, through December 31, 1967, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment, or 

in combination with work days within the parameters established for one or more other 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 9 of 54

classes of employees included in the Special Exposure Cohort (78 FR 229; November 

27, 2013).

It was determined that NIOSH lacked the sufficient information to allow it to estimate with 

sufficient accuracy the potential internal doses from exposure to thorium, to which employees 

working at this facility may have been subjected (Sebelius 2013a).

All employees of the Feed Materials Production Center in Fernald, Ohio, who were not 

employed by National Lead of Ohio, NLO, or the Department of Energy or its 

predecessor agencies, who worked at FMPC from January 1, 1951, through December 

31, 1983, for a number of work days aggregating at least 250 work days, occurring 

either solely under this employment, or in combination with work days within the 

parameters established for one or more other classes of employees included in the 

Special Exposure Cohort (78 FR 229; November 27, 2013).

It was determined that NIOSH lacked the sufficient information to allow an estimate with sufficient 

accuracy the potential internal doses from exposure to uranium, to which employees of FMPC 

working at this facility may have been subjected 

This dose reconstruction infeasibility for the period 1951 through 1983 applies only to subcontractors. 

NIOSH has access to an electronic data set that contains the results of the uranium urinalysis 

bioassay program for all the years of FMPC operations, and the overwhelming majority of employees 

of the prime contractor (National Lead of Ohio, later named NLO Inc.) have results in the bioassay 

data set. However, the data set does not contain bioassay results for employees of companies other 

than the prime contractor (i.e., non-prime contractor employees). NIOSH has obtained a limited 

number of bioassay samples from non-prime contractor employees through data captures, but cannot 

be certain that all non-prime contractor employees’ bioassay data were retained by the site or 

captured by data capture efforts. Additionally, there are some reasons to conclude that the prime 

contractor did not consistently evaluate whether non-prime contractor employees should be monitored 

for radiation exposure because of the transitory nature of their work (Sebelius, 2013 b).

Dose reconstruction guidance in this document for the period before January 1, 1984, is presented to 

provide a technical basis for partial dose reconstructions for claims not compensated under the SEC 

(i. e., nonpresumptive cancers and SEC employment <250 days). Although it is not possible to 

completely reconstruct internal radiation doses for all workers for the period January 1, 1951, through 

December 31, 1983, NIOSH has determined, and HHS has concurred, that it is feasible to reconstruct 

external radiation doses for all FMPC workers for the period from January 1, 1951, through December 

31, 1983 (Sebelius 2013b).

6.2 OVERVIEW OF SITE OPERATIONS

Having begun operations in 1951, FMPC was a rather late addition to the United States nuclear 

weapons complex, but it was a very important one. The site included chemical processing, foundries, 

machine shops, storage yards, a uranium recycler, and a variety of other capabilities. In addition, 

FMPC was assigned a mission that to date had never been performed before, at least on such a large 

scale. Processing such large quantities of radioactive material, in this case uranium up to 10,000 

MTU annually along with small amounts of thorium (ASI ca. 1986, pp. 4–5) – with a staff of up to 

almost 2,900 – was a new type of endeavor.

When operational, FMPC was a large integrated facility that produced uranium metal feed materials in 

DOE defense program facilities throughout the United States (WMCO 1988, p. 3). FMPC used a 

number of processes that involved a variety of forms of uranium, such as uranium ore concentrate, 

uranium hexafluoride (UF6), and recycled uranium scrap from throughout the DOE complex. The 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 10 of 54

products were “variously sized, highly purified uranium metal forms of assorted standard isotopic 

assays” (WMCO 1988, p. 6) ranging from depleted to slightly enriched uranium metal products. The 

primary facilities, referred to as plants, are described in detail in the latest revision of ORAUT-TKBS0017-2, Technical Basis Document for the Fernald Environmental Management Project (FEMP) – Site 

Description (ORAUT 2004), and addressed in this document by reference. The radiological hazards 

associated with these processes and products resulted from the radioactivity of uranium, thorium,

their progeny, and in some instances impurities in the recycled material.

The occupational doses FMPC workers received were a function of the physical location of the 

workers on the site, the process, and the type and quantities of materials. The dose also varied with 

the “age” (with respect to radiological decay) of the material being processed. The concerns with 

aging and processing involve the disequilibria of the radioactive material progeny that often are 

affected by the process to which the parent is subjected. For example, the temperature of the 

process can volatilize a progeny in the parent radioactive decay chain (such as thorium) which in turn 

can, until equilibrium is reestablished, affect the resulting dose rates. These and other factors are 

often encountered in the processing of radiological materials, all of which can affect the amount and 

magnitude of external dose.

Many cycles of activities took place throughout the operational lifetime of FMPC. The throughput of 

material varied considerably as did the sources of feed materials, one of which was ore from the 

Belgian Congo. This ore, pitchblende, contained large quantities of radium that required shielding 

(EPA 2000). Wastes from this process were stored on the site in the K-65 Silos. These silos also 

received waste from a site near Niagara Falls, New York. The silos became a large contributor to site 

background dose rates.

The introduction of recycled uranium at FMPC, mostly from the DOE Hanford Site, started in 1958 

(DOE 2003, p. 75) and reached a peak in 1970. DOE (2000, p. 26) indicates that Fernald received, 

shipped, and processed uranium products from 1961 to 1989 that used recycled uranium feed stocks 

that contained constituents of concern. This material contained some “carryover” fission products and 

minute quantities of transuranic elements. One of the carryover products of interest for occupational 

external exposure is 99Tc, which has 0.292 MeV beta energy at a yield of 100%. It is important 

because its specific activity is approximately 4 orders of magnitude greater than that of uranium, and 

FMPC received an estimated 135 kilograms of 99Tc. Technetium-99 is a major contributor to shallow 

dose because it contributes most of its energies at the tissue depths that are used in defining shallow 

dose. The same is true of the similar beta energy in 234Th, the first progeny of the 238U decay series.

A more penetrating beta energy occurs with the decay of protactinium, the product of 234Th decay (the 

first progeny radionuclide of 238U decay), which undergoes a metastable state as 234mPa, which in turn 

decays (0.13% of time) to 234mPa with a 2.29 MeV beta at a yield of 98% and two lesser energy betas 

of 1 MeV each. Complete decay chain diagrams are provided in Table of Isotopes (Lederer and 

Shirley 1978) or the decay diagrams of the Radiological Health Handbook (PHS 1970) in figures at 

the end of this section.

There are several beta particles in the 235U decay chain, most of which are associated with 231Th, the 

first progeny of 235U. The maximum beta energy is 0.3 MeV ±0.005, and there are several lesser 

energies. In addition, several gamma and X-rays are emitted during the decay process. Table 6-1 

lists the radiations of major concern.

Workers at FMPC who might have been exposed to the sources of radiation that are discussed in this 

technical basis document (TBD) were employed during the period starting in late 1951. The peak 

production years occurred from the late 1950s to mid-1970s, peaking in 1960 at an annual rate of 

approximately 10,000 metric tons of uranium (MTU) (Voillequé et al. 1995, p. 306). Individual worker 

monitoring methods were implemented by work locations, length of time, and facility monitoring. No 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 11 of 54

Table 6-1. Uranium beta and gamma emissions of interest (MeV).

Radionuclide Max. beta energy (max.) Gamma energy

U-238 None None

Th-234 0.103 (21%) 0.063 (3.5%)

0.193 (79%) 0.093 (4%)

Pa-234m 2.29 (98%) 0.765 (0.3%)

1.00 (0.6%)

U-235 None 0.144 (11%)

0.186 (54%)

0.205 (5%)

Th-231 0.140 (45%) 0.026 (2%)

0.220 (15%) 0.084 (10%)

0.305 (40%)

U-234 None 0.053 (0.2%)

early radiological policy documentation at FMPC was found during the TBD investigations. However, 

individual doses from personal dosimeters worn by the workers are available, and this TBD pertains to 

the analysis of these records. OCAS-IG-001, External Dosimetry Implementation Guideline (NIOSH 

2007) has identified these records to represent the highest quality records for retrospective dose 

assessments.

6.3 BASIS OF COMPARISON

Occupational whole-body doses at the time of FMPC startup in 1951 were controlled to 0.3 R/wk and 

an extremity dose of 1.5 R/wk. The annual limit for maximum whole-body dose for any 1-year period 

was limited to 12 rem, and the annual extremity limit was 75 rem; both values had associated 

administrative limits that were fractions of the annual limits per calendar quarter. In 1955, the wholebody dose limits were reduced to 3 rem per 13 weeks, not to exceed 5 rem per year. The terms 

roentgen, rad, rem, and rep (roentgen equivalent physical) often are used interchangeably 

(Heatherton 1960a, p. 143), and in this document the favorable to claimant assumption was made that 

they are considered equal.

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

dose since the start of FMPC in 1951. A basis of comparison for dose reconstruction is the concept of 

personal dose equivalent Hp(d), where d identifies the depth in millimeters and represents the point of 

reference for dose in tissue. For penetrating radiation of significance to whole-body dose (e.g. highenergy photons), d = 10 mm and is noted as Hp(10). For weakly penetrating radiation of significance 

to skin dose, d = 0.07 mm, and is noted as Hp(0.07). These are the radiation quantities 

recommended by the International Commission on Radiological Units (ICRU) in Report 51 (ICRU

1993), and the radiation quantities used in the DOE Laboratory Accreditation Program (DOELAP; 

DOE 1986) to accredit personal dosimetry programs at DOE sites. FMPC was the first DOE site to 

become DOELAP accredited, in 1987. While this accreditation is of significant value in validating data 

from 1987 and later, there is no analogous validation for data from before 1987 and especially back to 

the 1950s. The accuracies of the dosimetry system(s), their recorded doses, and their comparability 

to current systems depend on:

• Administrative practices based on technical, statutory, and administrative requirements;

• Workplace radiation fields, materials, quantities, etc.;

• Dosimetry technologies and calibrations;

• Process technologies; and

• Training programs and practices.
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6.4 HISTORICAL DOSIMETRY PRACTICES

This section provides a summary of the external dosimetry practices FMPC used throughout its 

history of operations and activities.

6.4.1 Administrative Practices

FMPC started operation in October 1951 (DOE 2000, p. 737) using a variety of chemical and 

metallurgical processes to perform its mission of supplying uranium metal products to the U.S. Atomic 

Energy Commission (AEC) and its successor agencies. This included fuel cores for production 

reactors at the Hanford and Savannah River Sites and uranium metal products for Y-12 at Oak Ridge 

National Laboratory (ORNL) and Rocky Flats Plant operations. Raw uranium-bearing ores, ore 

concentrates, and later recycled uranium compounds were received primarily from the Hanford and 

Savannah River Sites.

Some of the raw uranium ore (especially ores from the Belgium Congo) contained considerable 

amounts of radium that later became a waste problem, while recycled material had trace quantities of 

transuranic elements and some fission products. Fernald receipts data (DOE 2003, p. 75) presented 

in Table 6-2 lists the quantities of recycled materials along with other radionuclides and their sources 

from key shipping sites.

Figures 6-1 and 6-2 present decay schemes for 235U and 238U, respectively. Of prime importance is 

Figure 6-2 (238U) because its decay scheme includes 234mPa, a major contributor to FMPC worker 

dose because FMPC processed only low-enriched uranium (i.e., <2% 235U). For the most part, only 

uranium that had been subjected to processing (primarily concentration) that resulted in the disruption 

of its decay chain was the main feed material. Typically, only the first two progeny of 238U are of 

importance, and in particular 234mPa with its 2.29 MeV beta energy 98% of the time. The FMPC

processed only low-enriched uranium (i.e. typically <2% 235U) but it varied between <0.7 and 5% (in 

limited quantities).

FMPC also became the storage site for thorium in the United States and processed some thorium into 

reactor fuel for the weapons complex. The radiological properties of thorium are different from those 

of uranium because it has higher energy gamma rays and a shorter time to reestablish equilibrium 

with its progeny after processing.

The dosimetry requirements for uranium and thorium are similar; that is, the first two progeny of 232Th 

decay by emission of beta particles along with a few photons. Because the progeny are volatile, 

when thorium is processed the equilibrium is disrupted and considerable time is required for the 

processed material to again reach equilibrium. Figure 6-3 provides the decay scheme for 232Th. Later 

in plant operations, 99Tc became a contributor to external dose when the FMPC started processing 

recycled materials that included small quantities of transuranic elements and fission products (e.g., 

237Np, plutonium, and 99Tc) (DOE 2003, p. 21). The transuranic elements shown in Table 6-2 

contributed to the internal dose, but only on a limited basis. Technetium contributed primarily to 

external exposure (skin or extremity) dose due to the quantities present, and probably was contacted 

through contamination of apparel (in particular gloves).

Dosimeters for production workers were always used at FMPC. After the security credential and the 

dosimeter were combined, all employees wore them (WMCO 1987). However, exposures have not 

always been determined for all employees. During certain periods, female employees were not 

routinely monitored (Dugan 1974). Periods when male and female employees were monitored were:

• 1951 to 1960, male employees only;

• 1961 to 1968, male and female employees;
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Table 6-2. Recycled materials receipts (DOE 2003, p. 75). 

Year

Key shipping sites (MTU) Contaminants

Hanford

Savannah 

River

West 

Valley Total

Plutonium 

(g)

Neptunium 

(g)

236U

(kg)

Technetium 

(kg)

1953 0 0 0.0 0 0.0

1954 0 0 0.0 0 0.0

1955 0 0 0.0 0 0.0

1956 0 0 0.0 0 0.0

1957 0 0 0.0 0 0.0

1958 5 5 0.0 2 0.0

1959 19 19 0.1 7 0.2

1960 21 21 0.1 7 0.2

1961 50 1 51 0.2 18 0.4

1962 170 170 0.8 60 1.4

1963 1,002 1,002 4.5 351 8.0

1964 1,097 1 1,097 4.9 384 8.8

1965

1966 2,025 202 48 2,274 9.7 752 17.1

1967 1,458 859 101 2,418 8.7 679 15.0

1968 1,692 412 168 2,273 9.1 690 15.5

1969 1,870 706 124 2,700 10.3 799 17.8

1970 2,237 22 78 2,336 10.4 798 18.2

1971 0 60 69 129 0.4 21 0.4

1972 0 31 31 0.1 4 0.1

1973 3 3 0.0 0 0.0

1974 15 15 0.0 3 17 1.2

1975

1976 12 12 0.0 2 14 1.0

1977 0 14 15 0.0 3 0.4

1978 28 28 0.1 5 23 1.0

1979 66 66 0.1 12 0.2

1980 36 36 0.1 7 17 2.5

1981 0 0 0.0 0 0.0

1982

1983 23 23 0.0 4 0.1

1984 706 27 733 3.2 252 5.7

1985 918 918 4.1 321 7.3

1986 1,151 1,151 5.2 403 9.2

1987 314 314 1.4 110 2.5

1988 123 123 0.6 43 1.0

1989 0 0 0.0 0 0.0

Total 14,859 2,486 621 17,966 74.3 5,735 71 135

Table 6-3. Dosimeter characteristics.

Years Dosimeter Filters MDL (mrad)

Routine 

exchange

1951–1953 Two-element film Open, Cd 1 mm 40 Weekly

1954–1958 ORNL dosimeter Cu, Cd, plastic, Pb, open 30 Biweekly

1959–1985 ORNL dosimeter Cu, Cd, plastic, Pb, open 30 Monthly

1985–1992 Commercial Panasonic TLD Multiple 5 Monthly

1993–present Commercial Panasonic TLD Multiple 5 Quarterly
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 14 of 54

Figure 6-1. Uranium-235 decay series (PHS 1970, p. 124).
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 15 of 54

Figure 6-2. Uranium-238 decay series (PHS 1970, p. 123).
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 16 of 54

Figure 6-3. Thorium-232 decay series (PHS 1970, p. 121).

• 1969 to 1978, male employees only; and

• 1979 to present, male and female employees.

Female employees were not monitored during certain periods because “the potential did not exist for 

them to exceed 10% of the quarterly standards” (NLO 1981, p. 4; Noyes 1968a, p. 2). In the later 

years, when operations were oriented towards cleanup and restoration, and to meet compliance with 

10 CFR Part 835, external radiation monitoring was performed for all occupational workers who 

entered areas at the FMPC and who had the potential to receive a 100-mrem whole-body dose in 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 17 of 54

1 year. For visitors, monitoring was performed for those who entered areas in which they had the 

potential to receive a 50-mrem whole-body dose in 1 year (Robinson 2002, p. 5).

Additional information is provided in Section 6.8 for reconstructing dose to unmonitored workers.

FMPC used several types of personnel dosimeters throughout its operational period. There were also 

several changes in occupational and administrative exposure limits during that period, including 

dosimeter exchange periods. Table 6-3 lists details of these changes (ORAUT 2003, pp. 8, 25).

Table 6-4 summarizes the calibration practices, calibration sources, quantity, and background 

correction use for FMPC dosimetry.

Table 6-4. Dosimetry calibration practices.

Years

Calibration 

source

Dosimetric 

quantity

Air or 

phantom

Natural background

correction?

1952–? RaU slab Exposure & rep Air Yes

?–1985 RaU Exposure & rad Air Yes

1985–present Commercial Rem Phantom Yes

During these periods dose or exposure limits changed with the lowering of limits as more knowledge 

was gained about radiation protection practices. Exposure limits to the whole body were established 

at 0.1 R/d in the early 1940s, and were reduced in 1948 to 0.05 R/d to the whole body and 0.3 R/wk to 

blood-forming organs. Skin and extremities limits were established at 0.6 and 1.5 R/wk, respectively. 

By 1955, whole-body exposure was reduced to 3 rem per 13-week period and 5 rem/yr from all 

sources of radiation. In 1959, the International Commission on Radiological Protection (ICRP)

Publication 2 recommended a limit for employees over 18 years old as 5 × (N – 18) rem total lifetime, 

where N is the age in years (ICRP 1959, p. 17). This limit included an annual restriction of 10 rem. 

The extremity limit remained constant at 75 rem/yr but was later reduced to 50 rem/yr. Fernald 

established site-specific lesser values as a guide to ensure that workers would not exceed limits. The 

guides were often one-fourth to one-third of the annual limits per dosimeter wearing period or per 

calendar quarter. In addition, exposures sometimes were controlled using other types of dosimeters 

such as pocket-type ion chamber “pencil dosimeters.” Time limits were based on the product of 

measured dose rates and time (Noyes 1968b, pp. 2–5).

While current minimum detection limits (MDLs) as listed in Table 6-3 are well defined (Cooper 1998), 

earlier limits were not. Because it is difficult to estimate MDLs for the early dosimetry systems, the 

values in this TBD are those for the analogous ORNL system. Important dose reconstruction 

parameters for FMPC workers are based on the following administrative practices:

• Dosimeter exchange policies,

• Dosimeter assignment policies,

• Lost dosimeter and dose assignment policies,

• Incident handling policies and information on how results were recorded, and

• Past exposure records for new employees.

6.4.2 Incidents

External radiation dose from worker involvement in incidents is included in the dose of record for 

Fernald workers. In addition, “Radiation Exposure Investigation” reports are documented in claimants 

records when site personnel investigated badge readings (NLO 1971a); these were tracked by 

dosimetry personnel on logsheets in the 1990s (FEMP 1990). A review of 1995 incidents indicated 

that incident investigation reports (IIRs) were initiated by either employees or supervisors. When 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 18 of 54

involved in or in proximity to an incident, an employee reported to the Medical Department at the end 

of the shift and filled out an IIR (Bogar 1986, p. 190).

6.4.3 Dosimetry Technology

The dosimetry technology at FMPC was approximately equivalent to that in use throughout the 

nuclear industry at the time. FMPC followed the ORNL program for dosimeter design and calibration. 

The exception was the lack of a requirement for neutron dosimetry at FMPC. Table 6-3 above lists 

data about dosimeter types, periods of exchange, and MDLs, while Table 6-4 lists calibration data 

over the same periods.

Extremity dosimetry at Fernald involved the use of wrist dosimeters (rather than finger dosimeters) 

together with application of an appropriate correction factor. Extremity exposures are believed to 

have been monitored using wrist film badges from approximately 1969 until 1977 (NLO 1981, p. 2). 

The use of thermoluminescent dosimeters (TLDs) for extremity dose measurements began on 

February 16, 1977 (Boback 1978, p. 1). There is some evidence that a correction factor of 3 (Abee 

1951 p. 3) was used as observed in the review of early worker exposure records. Documentation of 

the correction factor was not established until a study by Jones determined that a factor of 2.06 times 

the wrist dosimeter value should be used to estimate the dose to the extremity (Jones 1988, p. 179). 

The wrist dosimeter in use at the time of the Jones study was a Teflon disk embedded with 

CaSO4:Dy; however, records indicate that film previously had been used at FMPC.

The method of using wrist-to-finger ratios to estimate extremity doses is not a particularly accurate 

practice. This approach was used at many DOE sites, with each site determining its own correction 

factor. Extremity Hp(0.07) doses could be overestimated by this method by as much as 20% due to 

shielding by protective clothing on the extremities (e.g. gloves) because the wrist dosimeter is worn 

outside of any clothing. Therefore, the recorded extremity doses should be favorable to claimants and 

should provide the best estimate of Hp(0.07) for individual monitored employees.

This reasoning is also true for whole-body doses. After reviewing the tables in this TBD it is possible 

to determine that the preponderance of the radiation consists of beta particles, and that while this form 

of radiation can deliver substantial doses to bare skin in proximity it does not penetrate deeply into the 

body. The dose rate from the photon component from the radioactive decay of uranium is “minor 

compared to the beta dose rate” (Alvarez et al. 1984, p. 9). In addition, protective measures such as 

distance, shielding, clothing, gloves, etc., reduce beta dose rates appreciably without excessive bulk 

by approximately 20%.

Individual exposure records indicate an “open window” design for personnel monitoring devices that 

allowed both beta and photon radiations to reach the measuring element (film or TLD). Some DOE 

sites, including FMPC, incorporated a security credential in the dosimeter holder that in some 

instances covered the open window of the dosimeter. However, FMPC did not cover the open 

window with their security credential (NLO 1981, p. 2), which provided more accurate results than if 

the window had been covered.

An additional radiological concern at several locations at FMPC occurred when workers were 

subjected to high levels of radioactive material-bearing dust. This widespread source of 

contamination was a concern for personal dosimeters, so at times the dosimeters were enclosed in 

plastic bags for protection against dust contamination. The manner in which these contaminated 

dosimeters were handled was not identified, but this should not be an issue in dose reconstruction 

because the dosimeters were calibrated in plastic bags and no adjustments were made to the 

dosimeter results for either Hp(0.07) or Hp(10).
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 19 of 54

An evaluation of original recorded doses for FMPC workers based on these parameters should yield a 

good (best available) estimate of Hp(10). Where necessary, Hp(0.07) for those individual workers 

who came in direct contact with radiological source materials can also be obtained because open 

window results were recorded with no adjustments to those readings.

6.4.3.1 Multielement Film Dosimeters

At startup in late 1951, FMPC used the two-element beta/photon dosimeter that was developed at the 

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

dosimeter was basically the same as the Clinton Laboratory dosimeter in use at Oak Ridge (Wilson et 

al. 1990). It used DuPont 552 film, an open window, and a cadmium filter. The open window was 

recorded as skin dose, and beta plus gamma with the filtered response was recorded as whole-body 

gamma dose. The Pardue dosimeter was processed off the site from startup in 1951 to June 1952, 

when onsite processing began. In mid-1953, FMPC converted to an advanced ORNL dosimeter, 

which consisted of an open window along with cadmium, lead, copper, and plastic filters. However, 

there are discrepancies in the documentation about the use of this dosimeter at FMPC. A health 

protection review of National Lead of Ohio Company (NLO, the Fernald operating contractor at the 

time) on May 14 to 16, 1963, indicated that the ORNL-type badge and DuPont 545 and 558 film 

packets were used. The 545 packet contained the 555 film dosimeter film and the 558 contained the 

508 and the high-range 1290 dosimeter films. All NLO employees, AEC area office personnel, and 

visitors were monitored in this manner (Johnson and Heacker 1963, p. 5). A letter dated September 

11, 1981, in response to a Dosimetry Assessment Fact Sheet states that in “January 1961, the ORNL 

badge meter, Model II was put into service at this site” (NLO 1981, p. 2). From 1951 through 1960, 

the film packet was sealed in plastic and placed in a metal case that was attached to the worker’s 

security badge. A portion of the film was not covered by the metal case and served as a means of 

monitoring skin dose. Regardless, all designs were calibrated for each batch of film, with the optical 

density of the film was determined after irradiation. The unknown exposure was compared optically to 

the known calibration and recorded accordingly. Therefore, any multielement dosimeter would be 

adequate because only the open window and the element that provided approximately 1,000 mg/cm2 

shielding was used.

6.4.3.2 Thermoluminescent Dosimeters

TLDs were introduced in or around 1978 or 1979, but only on an experimental basis. An exception 

was the extremity dosimeter program, which introduced TLDs in 1977; however, no data were found 

in the literature that described these TLDs other than “they were the Teledyne Teflon impregnated 

with calcium sulfate type” (NLO 1981, p. 2). Therefore, this TBD has assumed that the TLD was used 

in a wrist dosimeter configuration, and that dose calculations involved use of a modifying factor to 

provide some estimate of actual extremity exposure. The reviewed sample of records showed that 

while extremity doses were often near limits, the imposed whole-body restrictions limited worker 

extremity exposure to less than the extremity limit. It was concluded that the dose of record is the 

best to use for reconstruction of the extremity dose. At FMPC, extremity doses were calculated by 

correcting the wrist dosimeter results using a modifying factor of 3 for film and 2.06 in accordance with 

the Jones (1988) study. Jones (pp. 177, 202) also concluded that these measured workplace values 

should be reduced by 14%. However, FMPC did not retrofit the calculated extremity dose based on 

the new modifying reduction factor; therefore, the extremity ratios that were applied at Fernald are 

favorable to the claimant.

Whole-body TLDs using the commercial Panasonic system were placed in service in 1983 for testing

(Bogar 1986, pp. 177, 179; Adams 1984, p. 153). The system was subjected to an extensive study 

starting in the fall of 1981 (Plato and Miklos 1982, p. 74). The study included screening of 1,800 

dosimeters, calibration of a manual reader, development of correction factors for each of the four TLD 

chips in the dosimeter (E1 to E4), and the development of an algorithm and the calibration of FMPC
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 20 of 54

137Cs source. The TLD was the Panasonic UD-802 model, some of which were modified by NLO by 

removal of the plastic filter over element E2 on the front of the dosimeter. This modification was to 

improve the difference in the E1:E2 ratio to provide a better differential of the beta energies.

The algorithm that was developed by this study proved to be less than adequate, although the system 

did satisfy ANSI International Standard N13.11-1983 (ANSI 1983). The system lacked sufficient 

precision in estimating beta energies. Alvarez et al. (1984) found that, on average, skin dose was in 

error by about 1%. At a given location, this error could vary between –36% and +45%, and the review 

of the algorithm for calculating skin dose indicated that individual TLD readings of approximately 4% 

caused variation in skin doses of 15% to 25%. This led to the development of a new algorithm and 

the previously mentioned change in the plastic filter in the dosimeter. These changes had an impact 

on skin dose and in some cases eye dose, but they had no effect on deep dose because neither the 

filter nor the evaluation of deep dose was changed.

The appropriate TLD correction factors (based on the Gesell algorithm) were retroactively applied to 

dosimetry records from the beginning of TLD implementation (i.e., the period from 1983 to 1985) as 

described in Bogar (1986) and Adams (1984). Both references describe the deficiencies in the

original TLD algorithm that was developed by Plato, and the studies, comparisons, and fieldwork that

were undertaken to develop a more precise algorithm (the Gesell algorithm). The deficiencies of the 

Plato algorithm affected the precision of electron dose measurements and resulted in overestimates 

of electron dose.

6.4.4 Calibration

It is always prudent and technically defensible to calibrate to the same types and energies of the 

radiations to be measured. For FMPC, this would involve uranium and its progeny, in some cases 

99Tc, and (perhaps early in FMPC operations) 226Ra. Potential errors in recorded doses depend not 

only on the response of the specific dosimeter to the radiation to which it is exposed and calibrated, 

but also on the dosimeter geometry, how it is worn, and the simple variables in shielding that are 

afforded by clothing and other materials. The shielding effect is especially significant when the 

radiations are primarily beta particles or low-energy photons, both of which are predominant with 

uranium.

6.4.4.1 Beta/Photon Dosimeters

FMPC dosimeters were originally calibrated using a slab of natural uranium for beta energy and lowenergy photons and 226Ra for gamma energy. The uranium calibration determined the nonpenetrating 

or skin dose by measuring the film density behind the open window of the dosimeter. The penetrating 

dose was determined by measuring the film density behind the metal filter. All calibrations were made 

“in air” (i.e., no phantom) and for each batch of film. From a review of the available documentation, 

exposures were made for 226Ra gamma radiation at 50, 150, 450, 1,356, and 4,077 mR. Uranium 

slab exposures were made at 40, 80, 160, 320, 640, and 3,840 mrep (NLO 1971b, 1974, 1975, 1977). 

By January 1977, calibration values had changed to a range of 25 to 4,077 mR in 10 increments for 

226Ra and to a range of 20 to 3,840 mrep for beta emissions (for a uranium slab). All dosimeters were 

calibrated with and without “bags.” It was not determined when bags were first incorporated in the 

calibration process, but calibration data from September 1974 indicated that five gamma exposures 

ranging from 50 to 4,077 mR 226Ra and 10 uranium slab exposures ranging from 40 to 8,000 mrep 

were conducted “with bags.”

This TBD assumes that when FMPC began the practice of using bags for contamination control, 

attendant calibration procedures were revised to accommodate the new practice. Small changes in 

film density can mean large changes in recorded exposure. Therefore, it is desirable to calibrate 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 21 of 54

dosimeters using the same radiological sources and energies to which worker dosimeters will be 

exposed; this practice was implemented at FMPC.

6.4.4.2 FMPC Beta/Photon Dosimeter

As previously stated, FMPC beta/photon dosimeters were calibrated using 226Ra for gamma energy 

and uranium for beta and low-energy photons. All calibrations were in air (i.e., no phantom was 

used). This practice continued until the change to TLDs in 1983. When participation in DOELAP 

performance testing was required in the mid- to late 1980s (DOE 1986, p. 12), FMPC began using 

137Cs as the gamma calibration source while continuing to use the uranium slab. In addition, the use 

of a phantom was required at this time. Whether a change was determined to be necessary in the 

recorded penetrating radiation dose as a consequence of this change has not been determined. It is 

probable that a decrease in Hp(10) would result from these actions because most other sites 

experienced similar results. No change in the recorded dose is proposed to account for what could be 

a 10% decrease in dose.

The use of the millirep unit is somewhat unique to FMPC because it declined in use elsewhere after 

the 1950s. There are few references, including the Radiological Health Handbook (PHS 1970), so 

this TBD assumes that a rep is approximately 93 ergs/g of tissue. Because FMPC used this term 

interchangeably with rad (100 ergs/g of any receptor), there is a further inherent conservativeness of 

approximately 7%.

6.4.5 Workplace Beta/Photon Radiation Fields

Alvarez et al. (1984, pp. 9, 13) characterized typical FMPC radiation fields in preparation for 

introduction of the TLD system. The radiation fields consisted of a complex mixture of beta, X-rays, 

and gamma energies. These were supplemented by higher energy gamma radiation associated with 

226Ra transitions that account for the dose rates associated with the K-65 Silos.

6.4.5.1 FMPC Beta/Photon Dosimeter Response Testing

No data or evidence has been identified of early response testing of FMPC dosimeters, but the site 

used both the Pardue and ORNL dosimeters since the beginning of operations. Tests of these or 

similar dosimeters indicated an overestimate for Hp(10) with energies greater than 100 keV for the 

two-element dosimeter and an underestimate of dose for energies less than 100 keV (unless an 

adjustment is made using a fraction of the response of the shielded portion of the dosimeter). It is 

reasonable to assume that the early versions of the FMPC dosimeters reacted similarly to the ORNL 

dosimeter system, given that FMPC used the ORNL system until changing to a commercial system in 

1985.

In December 1983, there was an intersite comparison of the FMPC system between FMPC, Pacific 

Northwest Laboratories, and the Idaho National Engineering Laboratory (INEL) (Hayes 1983, p. 1;

Author unknown ca. 1983). The dosimeters were exposed in air, mounted on a 0.75-in. plywood disk 

and rotated at 2 rpm during exposure to several different uranium or uranium-bearing materials. 

Conclusions reached as a result of these tests were:

• NLO skin dose results were usually high but satisfactory.

• Penetrating dose compared to within ± a few percent (<10%), and NLO results were usually 

high.

• All three systems were satisfactory for adequate determination of both types of exposures.
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After these tests, NLO conducted several projects to improve its ability to determine incident beta 

energies. NLO contracted with INEL to characterize FMPC radiation fields (Alvarez et al. 1984, p. 9). 

These actions, along with the conclusion and attendant suggestions, resulted in changing the 

algorithm that was developed by Plato and Miklos (Plato 1983, p. 3; Cooper 1988, p. 7). 

Documentation indicates that official use of TLDs began on January 1, 1985, but the tests took place 

in late 1983 and early 1984 (NLO, 1984, p. 1-2). The documentation review for preparation of this 

TBD determined that FMPC used the total response of the unshielded or open window for skin dose 

and the response behind the shield for penetrating dose before the introduction of the TLD system. 

This approach is consistent with accepted practices of the time throughout the weapons complex, 

especially in non-plutonium facilities. This results in a conservative dose estimate that is favorable to 

claimants.

6.4.5.2 FMPC Workplace Beta/Gamma Dosimeter Response

Results of tests of FMPC dosimeters that were used during the 1960s included the conclusion that the 

half-value thickness of absorption of UX-2 (234Pa) beta energy was approximately 110 mg/cm2

(Heatherton 1960b, p. 19). It was determined that “the combined dose rate from the surface of 

uranium metal in equilibrium with its two daughters, UX-1 (234Th) and UX-2 (234Pa), is about 

240 mrad/hr.” It was also determined that approximately 95% of the surface dose rate, or 

approximately 228 mrad/hr, originated from the UX-2 in the metal. The processing of the metal 

resulted in separation of uranium progeny, which produced much higher dose rates in portions of the 

product, process equipment, and byproducts. The reason for this increase in dose rates is the loss of 

self-shielding that is afforded by the mass of the in-process uranium. Further studies involved the 

absorption of such materials as film wrapping paper, polyethylene, cardboard, and Lucite. These 

materials were used because they were assumed to be nearly tissue equivalent. It was determined 

that the half-value thickness for tissue was approximately 110 mg/cm2 and, therefore, that the dose to 

the eyes or gonads was approximately 15% of the skin dose. It was also determined that coveralls 

that were worn by workers (about 30 mg/cm2

) reduced uranium beta exposure to the skin by 

approximately 20% (Hinnefeld 1983, p. 43). Figure 6-4 summarizes these data.

Figure 6-4. Beta dose vs. skin depth (Heatherton 1960b, p. 47).
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 23 of 54

While it was not explicitly stated in the documentation (Heatherton 1960b), it is assumed that the 

dosimeter in use at FMPC was the ORNL version and that the film was a DuPont type (Johnson and 

Heacker 1963, p. 5). There are some general data of results between film and TLD dosimeters from

November and December 1982. Review of these data indicate that the two types of dosimeters did 

not agree and that the ratio of film to TLD varied with the location of the exposure. In some instances 

the ratio was greater than 1 and in others it was less than 1. Most often, film results provided higher 

dose estimates than TLD results, which supports the conclusion that early film dosimeter results are 

favorable to claimants. However, TLD results have been documented to be more accurate than film 

dosimeters and provide a more representative measure of the true exposures (this is particularly true 

when measuring beta and low-energy photons). ICRU Report 43 concludes that TLDs, when placed 

under an appropriate absorber “...will constitute a dosimeter having a spectral and angular response 

close to the ideal” while, about film, it states, “Dosimeters based on sensing elements that are not 

tissue equivalent (e.g. photographic film) can also be used, though in general it is more difficult to 

ensure that the variation of response with energy and angle of incidence is correct. Such problems 

are often enhanced for beta and low-energy x rays” (ICRU 1988, p. 9).

The forms of radiation at FMPC varied from plant to plant with Plants 5 and 9 exhibiting the highest 

potential workplace dose rates. These plants were involved with metal reduction, casting, and rolling, 

and these processes generated the separation and migration of progeny 234Th and 234Pa (UX-1 and 

UX-2). As stated above, 234Pa contributes approximately 95% of the total beta dose rate. Therefore, 

any location in the process where this material accumulated resulted in the potential for higher 

exposure rates. Other areas of potential high radiation exposure included areas where progeny

contaminated other materials (i.e., crucibles, saws, and rolling mills) or where large quantities of the 

parent material were present.

Studies in May 1984 (Boback 1984, p. 160) of various activities in Plant 5 indicated that the wholebody dose rate Hp(10) ranged from 0.1 to 1 mrem/hr. Studies in November and December 1982 

using the same criteria (i.e., 80% of the time workers were performing their jobs) indicated that dose 

rates ranged from 0.08 to 16.5 mrem/hr. These dose rates were established for Plants 5, 6, and 9, 

with Plant 9 (Green 1981, p. 4) exhibiting the highest rate and Plant 5 the lowest. The dose rate 

values were derived using dosimeter data and applying the 80% work rate rule. These values were 

not intended to be precise measurements of actual dose rates and resultant doses to workers, but 

rather as representations of the dose rate ranges that were present. These and other studies resulted 

in the use of shielding and establishment of limits on materials that could be present at any time in 

certain locations. There is discussion in the documentation of the use of improved housekeeping to 

assist in reducing worker exposure (Boback 1984, pp. 145–146). There is also discussion of process 

changes, (i.e., Lucite face shields, rubber matting, ingots surface rinsing) that were used during “ingot 

pickling” to remove “beta emitting daughters.” The FEMP External Dosimetry Technical Basis Manual

contains information on dosimeter response data after 1984 (Robinson 2002).

Table 6-5 lists the lower limits of detection for DOELAP categories for FMPC, and Table 6-6 lists the 

maximum viable dose ranges for DOELAP categories.

Table 6-5. Lower limits of detection for DOELAP categories.

Category Hp(0.07) (mrem) Hp(10) (mrem)

Cesium <5 <5

M30 (X-ray L) <16 <16

S60 (X-ray M) <16 <16

M150 (X-ray H1) <16 <16

M150 (X-ray H2) <16 NA

Sr/Y-90 (Beta) <16 NA
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Table 6-6. Maximum viable category 

dose ranges.

Category

Hp(0.07)

(rem)

Hp(10)

(rem)

Acc gamma (Cs) 5,150 5,150

Acc X-ray H1 7,050 7,350

X-ray L 5,400 2,250

X-ray M 5,750 5,350

X-ray H1 7,050 7,350

X-ray H2 7,050 7,050

Gamma (Cs) 10.3 10.3

B\B+G 1,257 10.3

X-ray L+B 10.2 5.9

X-ray M+B 5,188 10.2

X-ray H1+B 5,625 11.2

X-ray H2+B 5,513 11

X-ray L+G 10.4 6

X-ray M+G 10.6 10.4

X-ray H1+G 11.3 11.4

X-ray H2+G 11.3 11.3

The angular dependence of the FMPC dosimeter satisfied the DOELAP Angular Dependence 

Requirements (Hinnefeld 1989, p. 163). The system was DOELAP-accredited in 1987 by meeting all 

requirements at that time. It has maintained accreditation as indicated in the FEMP External 

Dosimetry Technical Basis Manual (Robinson 2002, p. 6). FMPC has further improved its dosimetry 

by the addition of an improved computer for the system that enables it to generate element correction 

factors for each TLD chip in each dosimeter. New correction factors are generated on a biannual 

basis.

Table 6-7 lists average annular responses for the DOELAP energies of importance at FMPC.

The Hp(10) response is accurate for up to ±60 degrees, while Hp(0.07) varies drastically (as would be 

expected). Clothing and other barriers can offer some reduction in Hp(0.07), but no adjustments were 

made to dosimeter results for any exposure [e.g., Hp(10) or Hp(0.07)]. The logical conclusion is that 

the recorded doses are favorable to claimants. The selection of beta and photon energies for the 

major FMPC facilities is summarized in Table 6-8. In the absence of a claimant specific work location, 

or if the claimant had multiple work locations in a given year, dose reconstructors should use the 

favorable to claimant default assumptions of 40% 30- to 250-keV and 60% >250-keV photon energy 

ranges along with 100% >15-keV energy for electrons and 100% 0.1- to 2-MeV energy range for 

neutrons, as applicable.

Alvarez et al. (1984) also estimated dose rates typical for the spectra that were measured at several 

locations on the plant. Those values are 0.35 mR/hr for low-energy photons <20 keV (typical for 

shallow skin dose) and 0.86 mR/hr for photons >30 keV and <300 keV. For energies >300 keV a 

deep dose rate of 1.4 mR/hr was calculated. These values are provided here for example only and if 

ratioed yield 13%, 33%, and 54%, respectively, for the energy categories used in dose reconstruction. 

The average photon radiation dose would be much lower and is “almost negligible compared with the 

beta skin dose,” as previously stated.

6.4.6 Neutron Dosimetry

The documentation for FMPC did not include any reference to neutron dosimetry with the exception of 

high-range gamma-sensitive film. This film was packaged with the sensitive 508 film and exchanged 

on an annual basis. There was some concern expressed in AEC audit letters (Johnson and Heacker 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 25 of 54

Table 6-7. Angular response for DOELAP energies of 

importance.

Energya Angle Planeb Hp(10)Rd/Do avg.c Hp(0.07)Rd/Do avg.c

M30 0 H 1.02 1.012

M30 ±20 H 1.012 0.997

M30 ±40 H 0.967 0.955

M30 ±60 H 0.769 0.764

M30 ±60 V 0.928 0.918

S60 0 H 1.019 1.021

S60 ±20 H 1.028 1.031

S60 ±40 H 1.021 1.017

S60 ±60 H 0.964 0.967

S60 ±60 V 0.893 0.896

M150 0 H 1.026 1.026

M150 ±20 H 1.061 1.068

M150 ±40 H 1.102 1.102

M150 ±60 H 1.02 1.02

M150 ±60 V 1.058 1.058

Cs137 0 H 0.964 0.964

Cs137 ±20 H 0.973 0.972

Cs137 ±40 H 0.966 0.966

Cs137 ±60 H 0.934 0.934

Cs137 ±60 V 0.890 0.891

Sr90/Y90 0 H 0.995 N/A

Sr90/Y90 ±20 H 0.982 N/A

Sr90/Y90 ±40 H 0.720 N/A

Sr90/Y90 ±60 H 0.367 N/A

Sr90/Y90 ±60 V 0.340 N/A

M30 ±60 V 0.928 0.918

S60 0 H 1.019 1.021

S60 ±20 H 1.028 1.031

S60 ±40 H 1.021 1.017

S60 ±60 H 0.964 0.967

S60 ±60 V 0.893 0.896

M150 0 H 1.026 1.026

M150 ±20 H 1.061 1.068

M150 ±40 H 1.102 1.102

a. The entries for M30, S60, and M150 represent calibration energy 

categories utilized by DOELAP.

b. H = horizontal; V = vertical.

c. Rd = “dose read” meaning the value read from the dosimeter; Do = 

dose observed meaning the value given by DOELAP. N/A = not 

applicable.

1963, p. 5) that mention “the badge also contains components to evaluate personnel exposure from 

criticality accidents” (an event that never occurred at Fernald). There was no established need for 

neutron dosimetry at FMPC even though there were large quantities of UF4 and UF6. Enrichments 

were low enough (typically <2% 235U) that alpha-neutron reactions were limited. The limitations of 

nuclear track emulsion, type A, were well documented including an MDL of approximately 40 mrem 

for fast neutrons. Based on studies and calculations it was concluded that fast neutron exposures at 

FMPC would be less than the MDL (Cooper 1998, p. 14). The purpose of the following section is to 

discuss and develop a neutron-to-photon ratio for estimating neutron doses at Fernald. A possible 

source of low-level neutron exposure at Fernald is the alpha-neutron reaction from the uranium alpha 

particle interactions with fluorine atoms. This reaction primarily occurs with the production and 

storage of UF4 (green salt). The areas at Fernald that produced and/or stored green salt include the 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 26 of 54

Table 6-8. Beta and photon energies and percentages.

Building Description Radiation Energy (keV) Percentage

Plants 2/3 Production of UO3 Beta

Photon

>15

>30 and <250

>250

100

40

60

Plant 7a Reduction of UF6 to UF4 Beta

Photon

>15

>30 and <250

>250

100

40

60

Plant 1 Sampling plant Beta

Photon

>15

>30 and <250

>250

100

40

60

Refinery Beta

Photon

>15

>30 and<250

>250

100

40

60

Pilot Plantb Scrap recovery Beta

Photon

>15

>30 and <250

>250

100

40

60

UF6 to UF4 reduction Beta

Photon

>15

>30 and <250

>250

100

40

60

Plant 8b Scrap recovery Beta

Photon

>15

>30 and <250

>250

100

40

60

Plant 5b Metal production Beta

Photon

>15

<30

>30 and <250

>250

100

13

33

54

Plant 6b Metal fabrication Beta

Photon

>15

<30

>30 and <250

>250

100

13

33

54

Plant 9b,c Special products Beta

Photon

>15

<30

>30 and <250

>250

100

13

33

54

Plant 4c UF4 Beta

Photon

>15

>30 and<250

>250

100

40c

60c

a. Operated only 2 years from June 1954 to May 1956.

b. Alvarez et al. (1984, p. 17).

c. For Plants 4, 8, 9, and the Pilot Plant, the years processing thorium were: Plant 4, 1954; 

Plant 8, 1967–1971; Plant 9, 1954–1955; and Pilot Plant, 1964–80. For workers employed 

there, the energy range is 25% >30 and <250 keV and 75% >250 keV.

Pilot Plant, Plant 4, Warehouse 4B, and any other warehouse at Fernald for which the stored material 

is unknown. The neutron-to-photon ratio described below should only be applied to workers who 

were exposed to uranium fluoride materials (i.e. UF4, UF6) at Fernald.

6.4.6.1 Development of the Neutron-to-Photon Ratio

Neutron exposures at Fernald were evaluated in 1998 in Warehouse 4B where the storage of 

enriched UF4 (green salt) in arrays of drums had the potential for generating measurable neutron 

fluxes. To determine if neutron doses in Warehouse 4B were of significant magnitude to warrant 

special neutron monitoring for workers who were involved in the inspection and packaging project, 

long-term neutron dose rates were measured at various locations using Landauer Neutrak-ER area 

dosimeters. The Landauer dosimeter is a dual-element device that uses a lithium albedo chip and a 

CR-39 foil so that an extended range of neutron energies can be monitored. A set of six Neutrak-ER 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 27 of 54

dosimeters were placed in fixed locations in the warehouse. To provide confidence that the long-term 

dosimeter measurements were representative of the dose rates in the building, and that there were no 

elevated hot spots of neutron dose rates, a special area survey was performed. This survey was 

performed using a Nuclear Research Corporation Model NP-2 "snoopy" portable neutron meter

(Robinson 2001).

Table 6-9 provides the neutron dose rate survey results. Seven measurements were taken using the 

10-minute integration period at locations near and among the various drums of material. As indicated 

in the table, none of the dose rate measurements exceeded the minimum detectable dose rate of 

0.02 mrem/hr. The gamma dose rates were subsequently measured at the above locations. With the 

exception of the control point, the dose rates were found to vary from 3 to 6 mrem/hr on contact and 

from 2 to 4 mrem/hr at 1 ft.

Table 6-9. Measured neutron dose rates (Robinson 2001).

Location

Measured neutron 

dose (mrem/hr) Location

Measured neutron 

dosea (mrem/hr)

Control point 0.001 Enriched Area 0.009

Red drum array 0.002 Enriched Area 0.018

Red drum array 0.003 Drum Conveyor 0.005

Red drum array 0.014

a. All results are below the calculated minimum detectable neutron dose rate of 0.02 

mrem/hr. These values are included here for completeness.

It was concluded that the external gamma dose rates within the established radiation area were 

consistently greater than 10 times that of the highest detected neutron dose rate.

To develop a neutron-to-photon ratio for UF4 (green salt), photon survey data was also required. 

Photon surveys were conducted in 2001 on 56 drums of UF4 (Ward and Hopson 2001, pp. 2–6). The 

drum midline dose rate varied from below detection limits to 1.5 mrem/hr, with a mean and standard 

deviation of 0.79 ±0.26 mrem/hr on contact. The photon dose rate at the bottom of the drum ranged 

from below detection limits to 3.5 mrem/hr, with a mean and standard deviation of 1.71 ±0.82 

mrem/hr. These two datasets were combined to estimate an overall uncertainty. The photon dose 

rate variability is primarily the result of differences in the fill level of the individual drums. The top of 

the drums were discarded from this analysis because the measured general photon dose rate was 

below detection limits. After combining the datasets, a distribution was fit to the data. The combined 

datasets resulted in a lognormal distribution with a geometric mean of 1.1 mrem/hr and a geometric 

standard deviation of 1.7.

The neutron-to-photon ratio (Rn/γ) is calculated by dividing the neutron dose rate in Table 6-9 by the 

photon dose rate:

Photon Dose Rate

Neutron Dose Rate Rn / γ = (6-1)

To propagate the uncertainty, a Monte Carlo simulation was conducted using distributions in place of 

each constant dose rate value. Figure 6-5 depicts the frequency distribution of the simulated neutronto-photon ratios for depleted and low-enriched uranium.

Statistical parameters of the distributions in Figure 6-5 are provided in Table 6-10. The geometric 

mean of the low-enriched uranium neutron-to-photon distribution was 0.10 with a geometric standard 

deviation of 1.71 and an upper 95-th percentile confidence of 0.23. This distribution is combined with 

measured and missed dose distributions using Monte Carlo methods described in ORAUT-OTIB0012, Technical Information Bulletin: Monte Carlo Methods for Dose Uncertainty Calculations

(ORAUT 2005a).
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 28 of 54

Neutron - Photon Ratio

0.00 0.05 0.10 0.15 0.20 0.25 0.30

Frequency

0

100

200

300

400

500

600

Depleted Uranium

Enriched Uranium

Figure 6-5. Comparison of neutron-to-photon ratio distributions for 

depleted and low-enriched UF4.

Table 6-10. Statistical parameters of neutron-to-photon ratio 

distributions.

Enrichment

Geometric 

mean

Geometric standard 

deviation

Upper 95th

percentile

Depleted uranium 0.07 1.74 0.17

Low-enriched uranium 0.10 1.71 0.23

Multiple approaches can be used to estimate the neutron dose using the neutron-to-photon ratios as 

presented. The factors that affect an individual’s neutron dose include: (1) the quantity of uranium 

processed, (2) the enrichment, and (3) the time an employee worked in a process or storage area. 

The recorded photon dose is also a function of the quantity of processed uranium and the exposure 

duration. The uranium enrichment has some small effect on the recorded photon dose in that 

enriched uranium generally has a higher photon dose rate than depleted uranium. This effect is 

ignored in this analysis for simplicity and is considered a favorable to claimant simplification because

a higher photon dose rate would decrease the neutron-to-photon ratio rather than increase the ratio 

(see Equation 6-1). Using the measured and missed photon dose should correctly account for 

changes in exposure that resulted from decreases in uranium production and changes in an 

individual’s exposure duration.

Through the years, Fernald workers conducted operations with enriched uranium, natural uranium, 

and depleted uranium. Figure 6-6 depicts the percentage of total uranium that was received and 

processed by Fernald in each of these three categories (DOE 2003).

The receipt for the maximum year (1967) was 24,400 MTU.

It should be noted that although natural uranium is not presented in Table 6-10 or calculated in this 

analysis, the neutron-to-photon ratio for natural uranium would fall in between the depleted and 

low-enriched ratios. This is because the alpha-neutron reaction is directly proportional to the alpha 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 29 of 54

activity of the sample. Depleted uranium has the least amount of 235U, which has a significantly higher 

specific activity than 238U and therefore an overall lower specific activity when comparing natural and 

enriched uranium. Natural uranium has a slightly higher specific activity in comparison with depleted 

uranium, and enriched uranium has the highest specific activity of the three forms.

As can be observed from the Figure 6-6, most of the uranium work before about 1965 was with 

natural uranium with a decrease beginning around 1960, at which time the percentage of enriched 

uranium increased. By 1970, relatively little work with natural uranium was conducted at Fernald. 

Beginning in the mid-1960s, work with depleted uranium began to increase. By the mid 1970s, work 

with depleted uranium constituted the majority of the uranium work at Fernald. During the transition 

period (1965 to 1975), work with enriched uranium initially increased so that by 1968 enriched 

uranium work constituted about 50% to 60% of the uranium work at Fernald. After this peak, work 

with enriched uranium decreased to about 10% to 20% of the total uranium work at Fernald.

 Figure 6-6. Relative percentage of uranium receipts by material category

 (natural, enriched, and depleted) (Voillequé et al. 1995, p. 689 and DOE 2000.

While the data in Figure 6-6 could be used to estimate annual neutron-to-photon ratios based on the 

percentage of enriched, natural, and depleted uranium work, to do so would unreasonably complicate 

the dose reconstruction and introduce additional uncertainty. The additional uncertainty stems from 

whether an employee might have worked predominately with one type of material or another. 

Because this uncertainty cannot be properly tracked without detailed job history and material tracking 

information, and to simplify the dose reconstruction, the low-enriched uranium neutron-to-photon ratio 

should be used. This assumption will tend to slightly overestimate the actual neutron-to-photon ratio 

and is considered a reasonable but necessary favorable to claimant assumption given the limited data 

available.
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 30 of 54

While there is a low probability that all of an employee’s photon dose would result from exposure to 

green salt (especially because Fernald is known to have processed large quantities of uranium metal, 

yellow cake, black oxide, etc.), the probability is not zero and cannot be excluded on an individual 

basis without significant job history information. As a result, the neutron-to-photon ratio should be 

applied to all measured and missed photon dose for employees who worked in the Pilot Plant, Plant 4, 

any warehouse or other area known to store UF4 or any of the onsite warehouses for which the stored 

material is not known. This favorable to claimant assumption is necessary to account for the 

uncertainty in the origin of the photon dose.

6.4.6.2 Neutron Activation Analysis

From 1955 until 1978, a variety of neutron sources were used to perform analytical neutron activation 

analysis (NAA) procedures. These sources included:

• RaBe source (1955 to 1967) (Ross 1967),

• Cockcroft-Walton accelerator (1964 to 1978) (Hervin and Johnson 1964, p. 8; NLO 1980, p. 1), 

and

• 252Cf source (1972 to1975) (Deily 1983; Boback 1975, p. 4).

Exposure to these sources would have been limited to workers in the Technical Division. Evidence of 

work with these sources would be found in claimant interview files or other workplace documentation 

from DOE. In relation to assigning neutron dose for NAA activities – if applicable – the neutron-tophoton ratios in Table 6-11 can be used. The recommended ratios for work with the RaBe source and 

the Cockcroft-Walton accelerator are based on gamma and neutron survey data that were taken by 

FMPC staff (Starkey 1963, Burgett and Thomas 1965, p. 13–24; Klein and Ross 1965). The ratio for 

the 252Cf source is based on gamma and neutron measurements that were made at ORNL for similar 

work with 252Cf (ORAUT 2007).

Table 6-11. NAA source neutron-to-photon ratios

Source Ratio Reference

RaBe 3.0 Starkey 1963

Cockcroft-Walton accelerator 4.1 ± 2.3a Burgett and Thomas 1965, 

Klein and Ross 1965

Cf-252 2.1 ORAUT 2007

a. Represents an average of the neutron-to-photon ratio for measured gamma and 

neutron dose rate values at the operator’s position with moderating water tank 

filled and unfilled and a use factor of 25% based 10 hours of use per week.

6.4.6.3 Workplace Neutron Radiation Fields

To date, specific neutron energy spectra of UF4 have not been found or modeled. According to the 

DOE Health Physics Manual of Good Practices for Uranium Facilities (Rich et al. 1988, p. 39), 

neutrons of approximately 2 MeV are generated from uranium alpha particle interactions with fluorine. 

If the neutrons are generated at this energy, through scattering interactions with surrounding 

materials, some fraction of the neutrons will have energy less than 2 MeV.

The POC calculations in IREP use a radiation effectiveness factor (REF) to estimate the POC. The 

most favorable to claimant neutron energy group (i.e., the neutron energy group with the largest REF) 

is the 0.1- to 2.0-MeV energy group. Because the neutrons at Fernald are assumed to be generated 

(born) in this energy group and although some fraction will scatter to lower (less favorable to claimant) 

energy groups, all of the neutron dose should be assumed to result from the 0.1- to 2-MeV energy 

group. In the absence of data, this is a reasonable and favorable to claimant assumption.
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 31 of 54

In situations that could have involved neutron activation analysis work, a variety of sources might 

have been used. Information about sources would be found in claimant interview files or other 

workplace documentation from DOE. The default neutron dose fractions are given in Table 6-12.

Table 6-12. Default neutron dose fractions.

Facilities/activities with neutron 

radiation Source

Neutron

energy (MeV)

Default dose 

%

Pilot Plant, Plant 4, and warehouses UF4, UF6, and various enrichments 0.1 to 2.0 100%

NAA RaBea 0.1 to 2.0

2.0 to 20

10%

90%

NAA Cockcroft-Walton acceleratorb 0.1 to 2.0

2.0 to 20

5%

95%

NAA Cf-252c <0.01

0.1 to 2.0

2.0 to 20

25%

57%

18%

a. NBS (1960).

b. Burgett and Thomas (1965).

c. ORAUT (2007).

6.5 RECORDED DOSES

FMPC recorded both skin and penetrating doses by determining film densities behind the open 

window and a single filter of approximately 1,000 mg/cm2

. The FMPC historical dose record practices 

are given in Table 6-13.

Table 6-13. Historical recorded dose practices.

a

Year

Dosimeter measured

quantities

Compliance dose

quantities

1951–1953

Film

OW: mrep

S (Cd): mR

Skin = OW+S

WB = S

1954–1985

Film

OW: mR

S: mrem

Skin = OW+S

WB = S

1985–present

TLD

Nonpenetrating (Npen)

Penetrating (Pen)

Skin = Npen+WB

WB = Pen

a. OW = open window; S = shielded; WB = whole body.

6.6 ADJUSTMENTS TO RECORDED DOSE

Corrections to the FMPC reported dose are required due to uncertainties in the recorded data and 

lack of significant data, especially before 1980. To satisfy the favorable to claimant criteria for dose 

reconstruction and considering the variability in radiation due to the FMPC processes, it is suggested 

that the corrections in the following discussion be considered.

A review of available data and documentation identified no mention that suggested the need for 

corrections to recorded whole-body dose. However, due to uncertainties in the dosimeter responses 

and the extended use of the two-element (or effectively two-element) dosimeter, a correction to 

pre-1985 recorded dose data is recommended. While there was a change to a multielement 

dosimeter, the penetrating dose was evaluated by the response behind the heavy metal filter. The 

heavy metal filter was cadmium in both dosimeters, which attenuated the lower energy photons (see 

Table 6-8, Figure 6-7 (Alvarez et al. 1984), and Figure 6-8 (Thorton et al. 1961)) and resulted in an 

underestimated response behind that filter for measured dose and Hp(10). The response of the 

ORNL dosimeter, as shown in Figure 6-8, indicates a general under-response below 100 keV, 

followed by a sharp over-response at 70 keV, followed by essentially no transmission of photons 

below 60 keV. Based on the spectroscopy data in Figure 6-7, approximately 10% of the gamma 

energy range likely falls into the under-response area of Figure 6-8. Because most, but not all, 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 32 of 54

penetrating radiations are above 100 keV, an adjustment is necessary for pre-1985 recorded 

penetrating whole-body doses due to the contribution to Hp(10) from low-energy photons including the 

L X-rays from both uranium and thorium. It is estimated that a correction equal to 10% of the <250 

keV values in Table 6-8 be added to the Hp(10) dose due to the contribution of these low-energy 

photons to penetrating dose that are absorbed in the thick filter.

Figure 6-7. Plant 5 gamma spectrum (Alvarez et al. 1984).

Figure 6-8. ORNL film dosimeter response (Thorton et al. 1961).

6.6.1 Adjustment for Neutron Quality Factor

The typical practice in 1995 was to use neutron flux to dose rate conversion factors in National 

Council on Radiation Protection and Measurements (NCRP) Report 38 (NCRP 1971). Some DOE 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 33 of 54

facilities, however, used a standard quality factor of 10 to convert from absorbed dose (rad) to dose 

equivalent (rem). For dose reconstruction purposes, the Report 38 doses must be converted to ICRP 

Publication 60 values (ICRP 1991). The conversion factor from NCRP Report 38 to ICRP Publication 

60 is 1.91 [see ORAUT-OTIB-0055, Technical Basis for Conversion from NCRP Report 38 Neutron 

Quality Factors to ICRP Publication 60 Radiation Weighting Factors for Respective IREP Input 

Neutron Energy Ranges (ORAUT 2006a)]. The favorable to claimant assumption of using a factor of 

2 should be applied to the neutron-to-photon ratio to convert the estimated dose to an ICRP 

Publication 60 dose equivalent value.

6.6.2 Dose Conversion Factors

Exposures were measured for the years that film dosimeters were used. Therefore, to obtain organ 

doses (as required for dose reconstruction), the exposure-to-organ dose conversion factors in OCASIG-001, External Dose Reconstruction Implementation Guide (NIOSH 2007), should be used for 

organs that are affected only by penetrating radiation (e.g., organs other than the skin, breast, and 

testes). Doses to the skin, breast, and testes (and any other cancer location potentially that is 

affected by nonpenetrating radiation) are determined based on both gamma and nonpenetrating 

columns; gamma doses are assigned as photons with an energy range consistent with information in 

this document, and nonpenetrating doses are assigned as electrons >15 keV with corrections to 

account for clothing attenuation or other considerations. Further guidance is provided in ORAUTOTIB-0017, Technical Information Bulletin: Interpretation of Dosimetry Data for Assignment of Shallow 

Dose (ORAUT 2005b).

6.7 MISSED DOSE

6.7.1 Missed Beta/Photon Dose

It can be assumed with some certainty that there have been missed doses in the recorded doses for 

FMPC workers. This could have resulted when a dosimeter was lost, if a worker was not monitored 

for short intervals of employment, or if a zero was entered because the dosimeter result was less than 

the MDL. Various methods were used to estimate lost dosimeter results such as using coworker 

results, products of time and dose rate, or previously recorded results for similar work. All of these 

required considerable review and examination.

Missed dose from MDLs is especially important when there were short exchange periods, generally 

through the 1950s and 1960s. That period also had higher MDLs. NIOSH (2007) describes options 

to calculate the missed dose. The recommended option is to estimate a potential missed dose where 

one-half of the limit of detection (LOD/2) is multiplied by the number of zero or less than LOD/2 dose 

results. This procedure is applicable to both Hp(0.07) and Hp(10) because the same dosimeter was 

used for both until the introduction of a TLD finger dosimeter. The MDL per period and exchange 

frequencies along with the product (LOD × exchanges) are listed in Table 6-14. The other values can 

be obtained from these data and the individual exposure record.

Table 6-14. Missed beta/photon dose.

Years

MDL (LOD)

(mrem)

LOD/2 

(mrem)

Exchange

frequency

Product

(mrem/yr)

1951–1953 40 20 Weekly 2,080

1954–1958 30 15 Biweekly 780

1959–1984 30 15 Monthly 360

1985–present 5 2.5 Quarterly 20
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 34 of 54

6.7.2 Unmonitored Neutron Dose

Determination of neutron dose, both measured and missed, should be done using the guidance 

regarding neutron to photon ratios given in Section 6.4.6.1.

As an example, suppose an employee worked in Plant 4 at Fernald in 1965. The employee’s 

measured photon dose was 400 ±80 mrem, and there were three zero monthly dosimeter readings. 

The energy employee’s total photon dose would be 445 ±82 mrem. [Since the LOD/2 = 15 mrem, 3 

zero readings X 15 mrem = 45 mrem (missed photon dose) + 400 mrem (measured photon dose) is 

445 mrem]). Applying the low-enriched neutron-to-photon ratio and the conversion to ICRP 

Publication 60 radiation weighting factors (ICRP 1991) would result in an estimated neutron dose of 

89 mrem, (445 mrem total photon dose x 0.1 x 2). The neutron dose distribution is a lognormal 

distribution with a geometric standard deviation of 1.77.

Summary of reasonable but favorable to claimant assumptions:

• Neutron-to-photon ratio developed using the average midline and bottom photon dose rates.

• Neutron-to-photon ratio developed using the low-enriched uranium neutron dose rate.

• Assumption that all recorded and missed photon dose was the result of exposure to enriched 

UF4 material.

• Assumption that all of the neutron energies are between 0.1 and 2 MeV.

• Assumption of a quality factor of 10 to convert rad to rem.

6.8 UNMONITORED DOSE

As noted previously, in a response to a dosimetry assessment fact sheet NLO indicated that female 

employees at FMPC at times were not routinely monitored (NLO 1981). This situation existed 

although the female workers wore a combined security and dosimeter badge. It is reasonable to 

postulate that there could have been other circumstances where workers (both women and men) 

might not have been monitored. 

6.8.1 Coworker Assigned Photon Dose

An estimated photon dose to unmonitored Fernald workers can be determined from monitored 

coworkers. ORAUT-OTIB-0020 (ORAUT 2011a) provides general instructions to evaluate the 

measured and missed doses for monitored Fernald workers to arrive at a favorable to claimant dose 

to be assigned to unmonitored workers from 1952 to 2005. If it is determined necessary to assign 

unmonitored dose after 2005, then dose reconstructors should apply 2005 coworker doses for 2006

through the present. Attachment A contains the details of the evaluation of Fernald coworker dose to 

be assigned to unmonitored workers. These measured doses include an analysis of the missed dose,

which is particularly significant for the earlier years with higher LODs and frequent dosimeter 

exchanges (weekly or biweekly).

6.9 ORGAN DOSE

Once the Hp(10) doses have been calculated for each year, these values are used according to 

OCAS-IG-001 (NIOSH 2007) to calculate the organ dose distribution. OCAS-IG-001 describes the 

method for calculating organ dose using identified exposure geometries. For the period at Fernald 

from startup through 1984, the “Exposure (R) to Organ Dose (HT)” dose conversion factors should be 
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 35 of 54

applied to the recorded and missed dose. For 1985 and later years, the “Deep Dose Equivalent 

(Hp(10)) to Organ Dose (HT)” dose conversion factors should be applied.

Worker orientation is a primary consideration for this process, but no definitive method is available to 

evaluate this factor. Therefore, Dose Reconstructors should use the guidance in Section 4.4, 

Exposure Energy and Geometry of OCAS-IG-001 (NIOSH 2007) for determination of the appropriate 

selection for DCF geometry. 

6.10 UNCERTAINTY

No information was identified to address the uncertainty in the positive recorded photon dose for 

FMPC workers during the years that film dosimeters were used. Since the FMPC film dosimeters 

were of the Pardue and Oak Ridge designs, an uncertainty value of ± 30% is recommended (ORAUT 

2007). The uncertainty information for the FMPC multielement TLD in Table 6-15 is based on 

DOELAP accreditation data from a facsimile (Author unknown 2003).

Table 6-15. Uncertainty.

FMPC dosimetry system Uncertainty 

Film (1951-1984) 1.3

Multielement thermoluminescent (1985–present) 1.06

6.11 GEOMETRIC CORRECTION FACTOR

Consideration should be given to geometry when performing dose reconstruction for uranium facility 

workers who worked with uranium metals, powders, or residues, and for workers who worked on 

equipment that was contaminated with uranium. An underestimation of the measured and missed 

photon doses could occur if the energy employee wore the dosimeter on the upper chest or lapel and 

not in the central area of the chest or on the waist. The organs in the lower torso region are most 

affected. These include, but are not limited to, the stomach, liver, kidney, ureter, gall bladder, 

pancreas, small intestine, large intestine, rectum, ovaries, uterus, urinary bladder, and prostate.

The geometric correction factor that was calculated for application to the general population of 

workers is 2.1 and should be applied as a constant in IREP as determined in DCAS-TIB-0013, 

Selected Geometric Exposure Scenario Considerations for External Dose Reconstruction at Uranium 

Facilities (NIOSH 2010b). This geometric correction factor should be applied to the measured and 

missed photon doses for operators, material handlers, and trade workers including chemical 

operators, production operators, uranium material handlers, pipefitters, carpenters, welders, sheet 

metal workers, electricians, foremen, etc. The guidance in DCAS-TIB-0013 should be used to adjust 

the measured and missed photon dose to Fernald workers. Because the factor in DCAS-TIB-0013 

was determined by an analysis of photon fluence – not dose computation – the recommendations of 

that document would apply to workers exposed to thorium as well.

6.12 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0017-6 Revision No. 01 Effective Date: 03/25/2014 Page 36 of 54

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GLOSSARY

accreditation

For external dosimetry, the assessment of whether or not a personnel dosimetry system meets 

specific criteria. The assessment includes dosimeter performance and the associated quality 

assurance and calibration programs.

accuracy

The characteristics of an analysis or determination that ensures that both the bias and 

precision of the resultant quantity will remain within the specified limits.

algorithm

Set of rules or steps for solving a problem, especially for calculating a value.

alpha particle (α)

See alpha radiation.

alpha radiation

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

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

charge of +2.

beta particle (β)

See beta radiation.

beta radiation

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

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

beta particle is a positron.

deep dose equivalent [Hp(10)]

The dose equivalent at the respective depth of 1.0 cm in tissue.

dose equivalent (H)

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

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

from the absorbed dose.

DOE Laboratory Accreditation Program (DOELAP)

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

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

records, and calibration programs.

dosimeter

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

received by an individual. See film dosimeter, pocket ionization chamber, and 

thermoluminescent dosimeter.

dosimetry system

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

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

and the interpretation and documentation of the results.
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DuPont 552 film packet

Film packet containing DuPont 502 sensitive film and DuPont 510 insensitive film.

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

(1) In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.

extremity

Portion of the arm from and including the elbow through the fingertips and the portion of the 

leg from and including the knee and patella through the toes.

favorable to claimant

In relation to dose reconstruction for probability of causation analysis, having the property of 

ensuring that there is no underestimation of potential dose, which often means the assumption 

of a value that indicates a higher dose than is likely to have actually occurred in the absence of 

more accurate information. See probability of causation.

film

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

wrapping. See film dosimeter.

film density

See optical density.

film dosimeter

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

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

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

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

densitometer. Also called film badge.

filter

Material used (1) in a dosimeter to adjust radiation response to provide an improved tissue 

equivalent or dose response and (2) in an X-ray machine to selectively absorb photons from 

the beam to reduce unnecessary exposure of individuals or to improve radiographic quality.

gamma radiation

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

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

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

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

gamma ray, particle, or photon (γ)

See gamma radiation.

ionizing radiation

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

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

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 
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high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

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

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

radiation.

luminescence

Emission of light from a material as a result of some excitation. See thermoluminescence.

neutron

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

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

neutron radiation

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

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

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

radioactive by absorbing free neutrons. See neutron.

nonpenetrating dose

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

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

open window

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

light protection). See film dosimeter.

optical density

Measure of the degree of opacity of photographic or radiographic film defined as OD = 

log10 (I0/I), the base-10 logarithm of the ratio of the reference light intensity I0 (without film) to 

the transmitted light intensity (through the film). Also called film density and density reading.

pencil dosimeters

See pocket ionization chamber.

penetrating dose

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

the skin.

personal dose equivalent [Hp(d)]

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

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

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

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

noted as Hp(0.07) and Hp(10), respectively. The International Commission on Radiological 

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

protection.

photon

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

pocket ionization chamber (PIC)

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

measure ionizing radiation. A PIC may be self-reading or require the use of a outside device 
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to be able to read the dosimeter. Also called pencil, pocket pencil, pencil dosimeter, and 

pocket dosimeter.

rad

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

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

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

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

derives from radiation absorbed dose; rad is also the plural.

radiation

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

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

radioactivity

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

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

nuclei.

rem

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

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

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

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

rep

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

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

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

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

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

is also the plural.

roentgen

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

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

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

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

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

shallow dose equivalent [Hp(0.07)]

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

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

sievert

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

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

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

equals 100 rem.

skin dose

See shallow dose equivalent.
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tissue equivalent

Substance with response to radiation equivalent to tissue. A tissue-equivalent response is an 

important consideration in the design and fabrication of radiation measuring instruments and 

dosimeters.

thermoluminescence

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

thermoluminescent dosimeter (TLD)

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

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

a measurement of absorbed dose.

thermoluminescent dosimeter chip

Small block or crystal of lithium fluoride in a thermoluminescent dosimeter. A TLD-600 

dosimeter contains a chip made from more than 95% 6

Li for neutron radiation detection, and a 

TLD-700 dosimeter contains a chip made from more than 99.9% 7

Li for photon and beta 

radiation detection. Also called crystals.

U.S. Atomic Energy Commission

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

whole-body dose

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

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

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

X-ray radiation

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

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

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

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
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TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Purpose ................................................................................................................................. 49

A.2 Background............................................................................................................................ 49

A.3 General Approach.................................................................................................................. 49

A.4 Applications and Limitations................................................................................................... 49

A.5 Coworker Data Development ................................................................................................. 50

A.6 Adjustment for Missed Dose .................................................................................................. 50

A.7 Coworker Annual Dose Summaries ....................................................................................... 51

LIST OF TABLES

TABLE TITLE PAGE

A-1 Missed external doses ........................................................................................................... 51

A-2 Annual external coworker doses modified to account for missed dose................................... 53

A-3 Annual external coworker doses modified in accordance with ORAUT-OTIB-0052 ................ 54
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A.1 PURPOSE

The purpose of this attachment is to provide information to allow dose reconstructors to assign doses 

that are based on site coworker data to FMPC workers who have no or limited monitoring data. In 

addition, the data in this attachment should be used to assign doses for gaps in dosimetry records. 

The data are to be used in conjunction with ORAUT-OTIB-0020, Use of Coworker Dosimetry Data for 

External Dose Assignment (ORAUT 2011a).

A.2 BACKGROUND

The Oak Ridge Associated Universities (ORAU) Team is conducting a series of coworker data studies 

to permit dose reconstructors to complete certain cases for which external or internal monitoring data 

are unavailable or incomplete. Cases that do not have complete monitoring data could fall into one of 

several categories:

• The worker was unmonitored and, even by today’s standards, did not need to be monitored 

(e.g., a nonradiological worker).

• The worker was unmonitored but, by today’s standards, would have been monitored.

• The worker might have been monitored, but the data are not available to the dose

reconstructor.

• Partial information is available, but it is insufficient to facilitate a dose reconstruction.

As described in ORAUT-OTIB-0020 (ORAUT 2011a), some cases without complete monitoring data 

can be processed based on assumptions and methodologies that do not involve coworker data. For 

example, many cases in the first category can be processed by the assignment of ambient external 

and internal doses based on information in the relevant site TBDs.

As described in the main text, operations at the site began in 1951. FMPC used film dosimeters 

between 1951 and 1985. Use of a multielement TLD was implemented in 1985. Exchange 

frequencies varied from quarterly to weekly dependent on the era. There does not appear to have 

been any significant administrative practice that would have jeopardized the integrity of the dose of 

record.

A.3 GENERAL APPROACH

As described in ORAUT-OTIB-0020 (ORAUT 2011a), the general approach to the development of 

coworker data for cases without external monitoring data is to assign either 50th- or 95th-percentile 

doses with the intent that the assigned doses represent, but do not underestimate, the doses that 

would have been assigned had the worker been monitored.

A.4 APPLICATIONS AND LIMITATIONS

Some FMPC workers could have worked at one or more other major sites in the DOE complex during 

their employment histories. Therefore, the data in this attachment must be used with caution to 

ensure that, for likely noncompensable cases, unmonitored external doses from multiple site 

employments have been overestimated. This typically requires the availability of the recorded doses 

or external coworker dosimetry data for all relevant sites.
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The data in this attachment address penetrating gamma radiation and nonpenetrating electron and/or 

low-energy photon radiation. Neutron data are not presented in detail; methods for determining 

neutron dose are discussed in Section A.7.

External onsite ambient dose should be applied as specified in the latest revision of ORAUT-PROC0060, Occupational Onsite Ambient Dose Reconstruction for DOE Sites (ORAUT 2006b).

A.5 COWORKER DATA DEVELOPMENT

The FMPC HIS-20 database served as the data resource for coworker data analysis. The following 

items describe assumptions that were made for the analysis:

• Different types of dosimeters were available in the HIS-20 database. Only records with a wear 

location of “chest” were used in the analysis.

• If a record had a blank Social Security Number, then the record was excluded.

• Deep and shallow dose values that were “null” were excluded and not treated as zeros. The 

fact that “null” records were not zero values was confirmed by comparison with DOE records 

from several claimants with the database.

• Dose records from 10 EEOICPA claims were compared to the database, and in all cases the 

results from HIS-20 data exactly matched the data from DOE under EEOICPA.

• The amount of time a badge was worn by an employee is determined in a calculated field 

called “WearLength” that is equal the number of days between the End_Date and the 

Begin_Date in the database.

– Only records with a WearLength more than zero and less than 400 days were considered 

in the analysis. Relatively few records were removed from the analysis by filtering the 

records in this way. Of 35,795 records, 35,251 met the filter criteria of a WearLength more 

than zero and less than 400 days. Only 544 records did not meet the filtering criteria.

– If the WearLength was more than 29 days then the dose was adjusted upward to represent 

365 days of continuous wear. Dose values were used “as is” in the analysis if the 

WearLength was less than or equal to 29 days.

• The shallow dose represents the sum of the shielded and open window portions of the 

dosimeter as discussed in Section 6.4 of this document.

A.6 ADJUSTMENT FOR MISSED DOSE

According to OCAS-IG-001, External Dose Reconstruction Implementation Guideline (NIOSH 2007), 

missed doses are assigned for reported zero readings for each monitoring cycle to account for the 

possibility that doses were received but either not recorded by the dosimeter or not reported by the 

site. In addition, reported dose values less than one-half the applicable MDLs are assigned as 

missed dose. Annual maximum potential missed doses are calculated by multiplying the number of 

zero or unrecorded badge readings by the reported dosimeter LOD and summing the results. These 

values are used as the 95th-percentile values of a lognormal distribution to calculate the POC, which 

is determined by DOL. Therefore, in IREP, Parameter 1 is equal to the calculated maximum annual 
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missed doses multiplied by 0.5 and Parameter 2 is equal to 1.52. These values represent the 

geometric mean and geometric standard deviation, respectively, for each year of analysis.

The assignment of maximum potential missed doses for monitored workers is particularly significant 

for FMPC workers from 1952 to 1953 when they could have been monitored weekly. Table A-1 lists 

the maximum annual missed dose by monitoring period based on information in the main text.

Table A-1. Missed external doses (rem).

Monitoring 

period

Penetrating and 

nonpenetrating 

LOD

Exchange 

frequency

Maximum potential 

annual missed 

penetrating and 

nonpenetrating dose

1952–1953 0.04 Weekly 2.080

1954–1958 0.03 Biweekly 0.780

1959–1984 0.03 Monthly 0.360

1985–2006 0.02 Quarterly 0.080

A.7 COWORKER ANNUAL DOSE SUMMARIES

Based on the described information and approaches, FMPC coworker annual external dosimetry 

summaries were developed for use in the evaluation of external penetrating and nonpenetrating dose 

for certain workers who were potentially exposed to workplace radiation but for whom there is no or 

limited monitoring data from DOE. These summaries were developed using the following steps:

Step 1. As described in Section A.6, the reported penetrating dose was modified for each worker to 

account for partial years of employment. This permits the dose reconstructor to assign an 

appropriate prorated dose to account for partial years of employment or potential exposure.

Step 2. One-half of the maximum potential annual missed doses in Table A-1 were added to the 

reported annual doses from Step 1 (with the exception of reported positive doses, in which 

case the maximum missed dose was reduced by the dose that corresponded to one badge 

exchange because it is not possible that all individual badge results were zero if a positive 

annual dose was reported).

Step 3. The 50th- and 95th-percentile annual coworker gamma doses were derived from the doses 

from Step 2 by ranking the data into cumulative probability curves and extracting the 

50th- and 95th-percentile doses for each year.

Step 4. Table A-2 lists the results of the coworker analysis. These percentile doses should be used 

for FMPC workers with no or limited monitoring data through the use of the methodologies in 

Section 6.0 of ORAUT-OTIB-0020 (ORAUT 2011a). In general, the 50th-percentile dose 

can be used as a best estimate of a worker’s dose when professional judgment indicates 

that the worker was probably exposed to intermittent low levels of external radiation. The 

50th-percentile dose should generally not be used for workers who were routinely exposed. 

For routinely exposed workers (i.e., workers who were expected to have been monitored and 

routinely exposed), the 95th-percentile dose should be applied. However, other options are 

available through the guidance in ORAUT-OTIB-0020. For instance, for cases in which 

routine monitoring data exist and coworker dose is used to supplement missing quarters or 

years, the percentile dose should be the one that is consistent with the recorded doses 

unless there is reason to believe that the worker’s job or location in that year differed 

significantly from the job or location during the years dose was recorded. For workers who 
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are unlikely to have been exposed, external onsite ambient dose should be used rather than 

coworker doses.

Step 5. Table A-3 lists penetrating dose values (as described in the steps above) that have been 

adjusted using the guidance in Section 8.0 of ORAUT-OTIB-0052, Parameters to Consider 

When Processing Claims for Construction Trade Workers (ORAUT 2011b). This guidance is 

applicable for construction trade workers who meet the criteria in Section 3.0 of that 

document. Because the document does not provide an adjustment factor for nonpenetrating 

dose, this dose component is not shown in this table.

Step 6. If needed, neutron dose should be calculated using the neutron-to-photon ratios in 

Table 6-10 of this document. These values would be applied to the penetrating dose values 

in Table A-2.
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FERNALD COWORKER DOSE ASSIGNMENT (continued)

Table A-2. Annual external coworker doses modified to account for 

missed dose (rem).

Year

Penetrating

95th percentile

Penetrating

50th percentile

Shallow

95th percentile

Shallow

50th percentile

1952 1,520 1,120 4,320 1,220

1953 1,315 1,040 5,332 1,221

1954 776 390 6,291 676

1955 1,177 390 5,088 676

1956 475 390 5,175 675

1957 475 390 5,099 776

1958 576 390 5,990 1,077

1959 366 180 6,482 1,268

1960 565 180 6,565 1,465

1961 867 180 5,881 867

1962 1,168 265 7,094 1,168

1963 1,769 466 9,591 1,870

1964 1,465 365 7,565 1,265

1965 1,484 265 6,798 1,168

1966 967 299 5,520 967

1967 967 265 6,927 1,168

1968 865 180 5,265 665

1969 867 180 5,179 666

1970 967 265 5,580 666

1971 466 180 3,038 366

1972 765 265 4,765 565

1973 571 265 4,176 505

1974 767 265 4,908 566

1975 666 265 3,760 494

1976 665 265 4,435 465

1977 767 179 4,321 369

1978 967 256 5,470 466

1979 867 198 5,078 366

1980 925 265 5,150 465

1981 698 180 4,906 366

1982 808 186 5,542 366

1983 566 172 7,100 274

1984 665 183 6,320 392

1985 431 35 4,843 113

1986 531 36 4,845 51

1987 485 39 3,758 42

1988 289 40 1,509 40

1989 124 40 170 40

1990 96 40 106 40

1991 111 40 115 40

1992 80 40 84 40

1993 62 40 70 40

1994 59 40 61 40

1995 57 40 64 40

1996 57 40 65 40

1997 53 40 54 40

1998 49 40 49 40

1999 43 40 43 40

2000 54 40 54 40

2001 66 40 67 40

2002 88 40 93 40

2003 82 40 83 40

2004 87 40 88 40

2005 206 40 212 40
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FERNALD COWORKER DOSE ASSIGNMENT (continued)

Table A-3. Annual external coworker doses modified 

in accordance with ORAUT-OTIB-0052 (rem).

Year

Penetrating

95th percentile

Penetrating

50th percentile

1952 1,720 1,160

1953 1,433 1,040

1954 937 390

1955 1,498 390

1956 515 390

1957 515 390

1958 656 390

1959 446 180

1960 725 180

1961 1,148 180

1962 1,569 305

1963 2,411 586

1964 1,985 445

1965 2,011 305

1966 1,288 353

1967 1,288 305

1968 1,145 180

1969 1,148 180

1970 1,288 305

1971 586 180

1972 1,005 305

1973 734 305

1974 1,007 305

1975 867 305

1976 865 305

1977 1,007 184

1978 1,288 293

1979 1,148 211

1980 1,229 305

1981 912 186

1982 1,065 194

1983 727 175

1984 865 190

1985 592 37

1986 732 39

1987 667 43

1988 392 40

1989 162 40

1990 122 40

1991 143 40

1992 100 40

1993 75 40

1994 71 40

1995 67 40

1996 68 40

1997 62 40

1998 57 40

1999 48 40

2000 64 40

2001 81 40

2002 112 40

2003 103 40

2004 110 40

2005 276 40