ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Fernald Environmental 

Management Project – Occupational External Dose

Document Number:

ORAUT-TKBS-0017-6 

Effective Date: 04/20/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 36

Subject Expert: Leo G. Faust

Document Owner

Approval: Signature on File Date: 04/14/2004

Terry A. Kuykendall, TBD Team Leader

Approval: Signature on File Date: 04/14/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 04/19/2004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 04/20/2004 James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ................................................................................................................... 3

Page

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

6.0 Occupational External Dose........................................................................................................5

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

6.2 Basis of Comparison...................................................................................................................6

6.3 Site Historic Dosimetry Practices ................................................................................................7

6.3.1 Site Historic Administrative Practices................................................................................... 7

6.3.2 Site Dosimetry Technology ................................................................................................ 10

6.3.2.1 Multielement Film Dosimeters ................................................................................... 11 

6.3.2.2 Thermoluminescent Dosimeters................................................................................ 11

6.3.3 Calibration .........................................................................................................................12

6.3.3.1 Beta/Photon Dosimeters ........................................................................................... 12

6.3.3.2 FEMP Beta/Photon Dosimeter .................................................................................. 13

6.3.4 Workplace Beta/Photon Radiation Fields........................................................................... 13

6.3.4.1 FEMP Beta/Photon Dosimeter Response Testing..................................................... 13

6.3.4.2 FEMP Workplace Beta/Gamma Dosimeter Response .............................................. 14

 6.3.5 FEMP Neutron Dosimetry....................................................................................18 

 6.3.5.1 Neutron Doses in the Work Place...................................................................18 

6.3.5.2 Workplace Neutron Radiation Fields...............................................................18 

6.4 Recorded Doses.......................................................................................................19 
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6.5 Adjustments to Recorded Dose................................................................................................. 19

6.6 Missed Dose .............................................................................................................................19

 6.6.1 Missed Beta/Photon Dose ................................................................................. 19 

 6.6.2 Missed Neutron Dose....................................................................................... 20 

6.6.6.6.3 Missed Dose for Unmonitored Employees...............................................................20 

6.7 Organ Dose ..............................................................................................................................20

6.8 Bias and Uncertainty.................................................................................................21 

References ........................................................................................................................................30

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

LIST OF TABLES

Table

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

Page

6

6-2 FEMP recycled materials receipts (DOE 2003) ........................................................................ 8

6-3 FEMP MED/AEC/DOE dosimeter characteristics ..................................................................... 9

6-4 FEMP dosimetry calibration practices ......................................................................................9

6-5 Lower limits of detection for DOELAP categories ................................................................... 15

6-6 Maximum viable FEMP category dose ranges ....................................................................... 15

6-7 Angular response for DOELAP energies of importance to FEMP........................................... 16

6-8 Beta and photon energies and percentages for FEMP........................................................... 17

6-9 Measured Neutron Dose Rates at Fernald..................................................................18

6-10 Statistical Parameters of Neutron to Photon Ratio distributions.......................................20

6-11 FEMP Facility Default Neutron Dose Fractions...........................................................22

6-12 Historical recorded dose practices............................................................................ 22

6-13 FEMP missed beta/photon dose ............................................................................................ 23

6-14 Missed dose for unmonitored employees....................................................................24 

6-15 Organ dose default geometries .............................................................................................. 24

6-16 Bias and uncertainty...............................................................................................25 

LIST OF FIGURES

Figure

6-1 Neutron - Photon Ratio...........................................................................................19

Page

6-2 Relative Uranium Receipts by Material Category..........................................................21

6-3 Uranium-235 decay series ..................................................................................................... 26

6-4 Uranium-238 decay series ..................................................................................................... 27

6-5 Thorium-232 decay series...................................................................................................... 28

6-6 Beta dose vs. skin depth ........................................................................................................ 29
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 3 of 36

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

11/12/2004 11/12/2003 00-A New technical basis document for Fernald 

Environmental Management Project (FEMP) – 

Occupational External Dose. Initiated by Leo G. 

Faust. 

12/30/2003 12/30/2003 00-B Incorporates OCAS comments. Initiated by Leo G. 

Faust. 

02/29/2004 02/29/2004 00-C Incorporates additional OCAS comments. Initiated 

by Leo G. Faust. 

04/13/2004 04/13/2004 00-D Incorporates additional OCAS comments. Initiated 

by Leo G. Faust.

04/20/2004 04/20/2004 00 First approved issue. Initiated by Leo G. Faust.

 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 4 of 36

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANSI American National Standards Institute 

Do dose observed

DOE U.S. Department of Energy

DOELAP U.S. Department of Energy Laboratory Accreditation Program

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

FEMP Fernald Environmental Management Project

FMPC Feed Material Production Center

HF hydrogen fluoride

ICRP International Commission on Radiological Protection

INEL [DOE] Idaho National Engineering Laboratory

keV kilovolt-electron, 1,000 electron volts)

MDL minimum detection limit or level

MED Manhattan Engineering District

MeV megavolt-electron, 1 million electron volts

mg milligram

mm millimeter

mR milliroentgen

mrad millirad

mrem millirem

mrep millirep

mSv millisievert

MTU metric tons uranium

NIOSH National Institute for Occupational Safety and Health 

NLO National Lead of Ohio, Inc.

ORNL [DOE] Oak Ridge National Laboratory

PNL [Battelle] Pacific Northwest Laboratories

R roentgen

Rd dose read

rem roentgen equivalent man

rep roentgen equivalent physical

rpm revolutions per minute

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 5 of 36

6.0

This section provides the technical basis for estimating worker radiological dose based on exposure to 

occupational external radiation sources at the U.S. Department of Energy (DOE) Fernald 

Environmental Management Project (FEMP).

OCCUPATIONAL EXTERNAL DOSE

6.1 INTRODUCTION

FEMP (formerly known as the Feed Materials Production Center, or FMPC) was a rather late addition 

to the Manhattan Engineering District/U.S. Atomic Energy Commission (MED/AEC) weapons 

complex, although 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, 

FEMP was assigned a mission that to date had never been preformed 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 and small amount of thorium with a staff of up to almost 2,900 (ASI date unknown), was 

a new type of endeavor.

When operational, FEMP was a large integrated facility that produced uranium metal products used 

as feed materials in DOE defense program facilities throughout the United States (RAC 1995). FEMP 

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

concentrate, uranium hexafluoride (UF6), and recycled uranium scrap generated throughout the DOE 

complex. The products were “variously sized, highly purified uranium metal forms of assorted 

standard isotopic assays” (Unknown 1988) ranging from depleted to slightly enriched uranium metal 

products. The primary facilities, referred to as plants, are described in detail in Section 2, Site 

Description, of the FEMP Technical Basis Document (TBD) and are addressed in this section of the 

FEMP TBD by reference. The radiological hazards associated with these processes and products 

resulted from the radioactivity of uranium, thorium and their daughter products, and in some instances 

impurities in the recycled material.

The occupational dose received by workers at FEMP was 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 daughter products that often 

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

process can volatilize a daughter product 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 FEMP. 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 and probably to resultant doses to individuals.

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

(DOE 2003) and reached a peak in 1970. This material contained some “carryover” fission products 

and minute quantities of transuranics. 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 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 6 of 36

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

dose because it contributes most of its energies at tissue depths used in defining shallow dose. The 

same is true of the similar beta energy found in 234Th, the first daughter product of the 238U decay 

series.

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

first daughter 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 et. al

1978) or the decay diagrams of the Radiological Health Handbook (PHS 1970) that are provided in 

tables 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 daughter product of 235U. The maximum beta energy is 0.3 MeV ±0.005 and several lesser 

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

lists the radiations of major concern.

Table 6–1. Uranium beta and gamma emissions of interest. 

Radionuclide Beta energy, MeV (Max) Gamma energy, MeV

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%)

U-234 None 0.053 (0.2%)

Workers at FEMP who might have been exposed to the sources of radiation discussed in this TBD 

were employed during the period starting in late 1951. Of most concern are those employed during 

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

approximately 10,000 metric tons of uranium (MTU) (RAC 1995). Individual worker monitoring 

methods were implemented by work locations, length of time, and facility monitoring. No early 

radiological policy documentation at FEMP 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.

6.2 BASIS OF COMPARISON

Occupational whole-body doses at the time of FEMP 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 one year 

period was limited to12 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 whole 

body 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 and in this 

document the claimant favorable 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 FEMP in 1951. A basis of comparison for dose reconstruction is the concept of 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 7 of 36

Personal Dose Equivalent, Hp(d), where d identifies the depth (in mm) 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 Hs(0.07). These are the radiation quantities recommended 

by the ICRU Report 51 (ICRU1993), and the radiation quantities used in DOELAP (DOE Laboratory 

Accreditation Program; DOE 1986) to accredit personal dosimetry programs at DOE sites. FEMP 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 obtained 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. 

6.3 SITE HISTORIC DOSIMETRY PRACTICES

This section provides a summary of external dosimetry practices utilized throughout the history of 

FEMP operations and activities. 

6.3.1

FEMP started operation in October 1951 (DOE 2000) using a variety of chemical and metallurgical 

processes to perform its mission of supplying uranium metal products to the 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.

Site Historic Administrative Practices

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 

transuranics and some fission products. Fernald Receipts Data (DOE 2003) 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 (at the end of this TBD section) 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 FEMP worker dose since FEMP processed only low-enrichment 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 daughter products of 238U are of importance, and in particular 234mPa with its 2.29 

MeV beta energy 98% of the time. The FEMP processed only low-enriched uranium (i.e. typically 

<2% 235U) but it varied between <0.7 and 5% (in limited quantities).

The FEMP also became the storage site for thorium in the U.S. and processed some thorium into 

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

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

its daughters after processing. 
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Table 6–2. FEMP recycled materials receipts (DOE 2003) 

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

The dosimetry requirements for uranium and thorium are similar; i.e., the first two daughter products 

of 232Th decay by emission of beta particles along with a few photons. Since the daughter products 

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 (at the end of this TBD section) 

provides the decay scheme for 232Th. Later in plant operations 99Tc became a contributor to external 

dose when the FEMP started processing recycled materials that included small quantities of 

transuranics and fission products, (e.g., 237Np, plutonium, and 99Tc). The transuranics shown in Table 

6-2 contributed to the internal dose, but only on a limited basis (DOE 2003). Technetium contributed 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 9 of 36

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

contacted through contamination of apparel (in particular gloves). 

The use of a dosimeter for production workers has always been employed at FEMP. After the 

security credential and the dosimeter were combined, all employees wore them. However, exposures 

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

not routinely monitored. Periods when male and female employees were monitored were:

1951 – 1960: male employees only

1961 – 1968: male and female employees

1969 – 1978: male employees only

1979 - 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” (Dugan 1981). Additional information and data are 

provided in Section 6.6.2 on missed dose for unmonitored employees and Table 6-14, “Missed dose 

for unmonitored employees.”

FEMP 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.

Table 6–3. FEMP MED/AEC/DOE dosimeter characteristics. 

Years Dosimeter Filters

MDL 

(mRad)

Routine 

exchange

1951-54 Two-element film Open, Cd 1 mm 40 Weekly

1954-58 ORNL dosimeter Cu, Cd, plastic, Pb, open 30 Bi-weekly

1959-85 ORNL dosimeter Cu, Cd, plastic, Pb, open 30 Monthly

1985-92 Commercial Panasonic TLD Multiple 30 Monthly

1993-present Commercial Panasonic TLD Multiple 20 Quarterly

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

correction use for FEMP dosimetry.

Table 6–4. FEMP dosimetry calibration practices. 

Years

Calibration 

source

Dosimetric 

quantity

Air or 

phantom

Natural background 

correction?

1952-? Ra--U slab Exposure & rep Air Yes

?-1985 Ra--U 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 with respect to radiation protection practices. Exposure limits to the whole body were 

established at 0.1 R per day in the early 1940s, and were reduced in 1948 to 0.05 R per day to the 

whole body and 0.3 R per week to blood-forming organs. Skin and extremities limits were established 

at 0.6 and 1.5 R per week respectively. By 1955, whole-body exposure was reduced to 3 rem per 

13-week period and 5 rem per year from all sources of radiation. In 1959, the International 

Commission on Radiological Protection issued ICRP-2 (ICRP 1959), which recommended a limit for 
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employees over 18 years old as 5(N-18) rem total lifetime, where N is the age in years. This limit 

included an annual restriction of 10 rem. The extremity limit remained constant at 75 rem per year, but 

was later reduced to 50 rem per year. 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. Additionally, 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 1968).

While current minimum detection limits (MDLs) as listed in Table 6-3 are well defined (Cooper 1988), 

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

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

reconstruction parameters for FEMP 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

• Past exposure records for new employees

6.3.2

The dosimetry technology at FEMP was approximately equivalent to that used throughout the nuclear 

industry at the time. FEMP followed the ORNL program for dosimeter design and calibration. The 

exception was the lack of a requirement for neutron dosimetry at FEMP. Table 6-3 lists data used at 

FEMP, including dosimeter type, period of exchange, and MDLs, while Table 6-4 lists calibration data 

over the same periods.

Site Dosimetry Technology

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

together with application of an appropriate correction factor. There is some evidence that a factor of 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 (Jones 1988) determined that a factor of 

2.06 times the wrist dosimeter value should be used to estimate the dose to the extremity. 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 FEMP.

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 Hs(0.07) doses could be over estimated by this method by as much as 20% due to 

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

side of any clothing. Therefore, the recorded extremity doses should be claimant favorable and

should provide the best estimate of Hs(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 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 associated with the radioactive decay of uranium is 

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

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

bulk by approximately 20%.
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Individual exposure records indicate an “open window” design for personnel monitoring devices that 

allows both beta and photon radiations to reach the measuring element [film or thermoluminescent 

dosimeter (TLD)]. Some DOE sites, including FEMP, incorporated a security credential in the 

dosimeter holder that in some instances covered the open window of the dosimeter. However, FEMP 

did not cover the open window with their security credential (Dugan 1981) which provided more 

accurate results than if the window had been covered. 

An additional radiological concern at several locations at FEMP 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; however, 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 Hs(0.07) or Hp(10). 

Because Hp(10) may be of predominant interest in dose reconstruction, uranium beta radiation can be 

considered insignificant. An evaluation of original recorded doses for FEMP workers based on these 

parameters should yield a good (best available) estimate of Hp(10). Where necessary, Hs(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. 

6.3.2.1 Multielement Film Dosimeters

At startup in late 1951, FEMP used the two-element beta/photon dosimeter 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 used at Oak Ridge. It utilized 

DuPont 552 film, an open window, and a cadmium filter. The open window was recorded as skin 

dose, beta plus gamma and the filtered response as whole body gamma dose. The Pardue dosimeter 

was processed offsite from start up in 1951 to June 1952, when on-site processing began. In mid1953, FEMP converted to an advanced ORNL dosimeter, which consisted of 552 film, an open 

window, with cadmium, lead, copper, and plastic filters. However, there are discrepancies in the 

documentation with regard to the use of this dosimeter at FEMP. A letter dated September 11, 1981, 

in response to a Dosimetry Assessment Fact Sheet (Dugan 1981) states that in “January 1971 the 

ORNL badge meter, Model II was put into service at this site.” Regardless, all designs were 

calibrated for each batch of film, with the optical density of the film determined after irradiation. The 

unknown exposure was compared optically to the known calibration and recorded accordingly. 

Therefore, any multielement dosimeter would be adequate since only the “open window” and the 

element that provided approximately 1000 mg/cm2 shielding was used. 

6.3.2.2 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” (Dugan 1981). 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 sampling of records reviewed 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 FEMP, extremity doses were calculated by 
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correcting the wrist dosimeter results utilizing a modifying factor of 3 for film and 2.06 as per the Jones 

study (ibid). FEMP did not retrofit the calculated extremity dose based on the new modifying factor.

Whole-body TLDs using the commercial Panasonic system were placed in service in January 1985. 

The whole system was subjected to an extensive study starting in the fall of 1981 (Plato and Miklos 

1982). The study included screening of 1,800 dosimeters, calibration of a manual reader, 

development of correction factors for each of the four TLD chips (E1–E4) in the dosimeter, and the 

development of an algorithm and the calibration of FEMP 137Cs source. The TLD used was the 

Panasonic UD-802 model, some of which were modified by National Lead of Ohio, Inc. (NLO, the

Fernald operating contractor at the time) 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 developed by this study proved to be less than adequate, although the system did 

satisfy American National Standards Institute (ANSI) Standard N13.11-1983 (ANSI 1983). The 

system lacked sufficient precision in estimating beta energies. Alvarez (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 used 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 had no 

effect on deep dose since neither the filter nor the evaluation of deep dose was changed. 

6.3.3

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

radiations to be measured. For FEMP, this would involve uranium and its daughter products, in some 

cases 99Tc and (perhaps early in FEMP 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 

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.

Calibration

6.3.3.1 Beta/Photon Dosimeters

FEMP 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, 642, and 3,840 mrep (Film Calibrations 1971, 1974, 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; however, 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.” 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 13 of 36

This TBD assumes that when FEMP 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 

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

exposed; this practice was implemented at FEMP.

6.3.3.2 FEMP Beta/Photon Dosimeter

As previously stated, FEMP 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 January 1985. When participation in 

DOELAP performance testing was required in the mid to late 1980s (DOE 1986), FEMP 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 mrep unit is somewhat unique to FEMP because it declined in use 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 FEMP used this term 

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

approximately 7%. 

6.3.4

Alvarez et al. (1984) characterized typical FEMP 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.

Workplace Beta/Photon Radiation Fields

6.3.4.1 FEMP Beta/Photon Dosimeter Response Testing

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

used both the Pardue and ORNL dosimeters since the beginning of operations. Tests performed on 

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 FEMP dosimeters reacted similarly to the ORNL 

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

1985. 

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

Northwest Laboratories (PNL), and the Idaho National Engineering Laboratory (INEL) (Hayes 1983). 

The dosimeters were exposed in air, mounted on a 3/8-inch 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.
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 14 of 36

• 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.

After these tests, NLO conducted several projects to improve its ability to determine incident beta 

energies. NLO contracted with INEL to characterize FEMP radiation fields (Alvarez et al. 1984). 

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

algorithm developed by Plato and Miklos (1982). Documentation indicates that official use of TLDs 

began on January 1, 1985; however, the tests took place in the late 1983–early 1984 period. The 

documentation review for preparation of this TBD determined that FEMP used the total response of 

the unshielded or open window for skin dose and the response behind the shield for penetrating dose 

prior to the introduction of the TLD system. This approach is consistent with accepted practices of the 

time throughout the AEC complex, especially in nonplutonium facilities. This results in a conservative 

dose estimate that is claimant-favorable.

6.3.4.2 FEMP Workplace Beta/Gamma Dosimeter Response

Results of tests of FEMP dosimeters used during the 1960s (Heatherton 1960) included the 

conclusion that the half-value thickness of absorption of UX-2 (234Pa) beta energy was approximately 

110 mg/cm2

. 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 daughter 

products, 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 afforded by the 

mass of the in-process uranium. Further studies involved the absorption afforded by 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, the dose to the eyes or gonads was 

approximately 15% of the skin dose. It was also determined that coveralls worn by workers (about 30 

mg/cm2

) reduced uranium beta exposure to the skin by approximately 20%. Figure 6-6, at the end of 

this TBD, summarizes these data.

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

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

1963). There is some general data of results between film and TLD dosimeters dated November and 

December 1982. After review of these data, the conclusion is 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 provide higher dose 

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

claimant-favorable. 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 (ICRU 1988) concludes (pg. 9) that 

TLDs, when placed under an appropriate absorber “...will constitute a dosimeter having a spectral 

and angular response close to the ideal”, while in regard to 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”.
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 15 of 36

The forms of radiation encountered at FEMP 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 daughter products 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 daughter products contaminated other materials (i.e., crucibles, saws, and rolling mills), or 

where large quantities of the parent material were present. 

Studies conducted in May 1984 (Boback 1984) of various activities in Plant 5 indicated that the wholebody dose rate Hp(10) ranged from 0.1 to 1 mrem/hr. Studies conducted 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 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 material limits 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). There is also discussion of process changes (i.e., “ingot 

pickling” to remove “beta emitting daughters from the use of Lucite face shields, rubber matting, ingots 

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

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

maximum viable dose ranges for DOELAP categories.

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

Category Shallow Hs (mrem) Deep Hd (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

 

Table 6–6. Maximum viable FEMP 

category dose ranges. 

Category Hs (rem) Hd (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
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Category Hs (rem) Hd (rem)

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 FEMP dosimeter satisfied the DOELAP Angular Dependence 

Requirements (Hinnefeld 1989). 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). FEMP 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 FEMP. 

Table 6–7. Angular response for DOELAP energies of 

importance to FEMP. 

Energy* Angle Plane Hp(10)

Rd/Do Avg.**

Hs(0.07)

Rd/Do Avg.**

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
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Energy* Angle Plane Hp(10)

Rd/Do Avg.**

Hs(0.07)

Rd/Do Avg.**

M150 ±20 H 1.061 1.068

M150 ±40 H 1.102 1.102

* The entries for M30, S60, and M150 represent calibration energy categories 

utilized by DOELAP.

** “Rd: is “dose read” or the value read from the dosimeter; “Do” is dose 

observed, or the value given by DOELAP.

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

expected). Clothing, etc., may offer some reduction in Hs(0.07); however no adjustments were made 

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

recorded doses are claimant-favorable. The selection of beta and photon energies for the major 

FEMP facilities is summarized in Table 6-8.

 Table 6–8. Beta and photon energies and percentages for FEMP. 

Building Description Radiation Energy, keV Percentage

Plants 2/3 Production of UO3 Beta

Photon

>15

>30 and <250 

>250

100

40

60

Plant 7* 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 5a Metal production beta

photon

>15

<30

>30 and <250

>250

100

13

33

54

Plant 6a Metal fabrication beta

photon

>15

<30

>30 and <250

>250

100

13

33

54

Plant 9a,b Special products beta

photon

>15

<30

>30 and <250

>250

100

13

33

54
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 18 of 36

Building Description Radiation Energy, keV Percentage

Plant 4b UF4 beta

photon

>15

>30 and<250

>250

100

40b 

60b

a. Alvarez et al. 1984. b. For Plants 4, 8, 9 and the pilot plant the years processing thorium are: Plant 4, (1954); 

Plant 8, (1967-71); Plant 9, (1954-55); and Pilot Plant (1964-80) and for those workers employed there, the 

energy range is 25% >30 and <250 KeV and 75% >250 KeV. *Operated only two years June 1954-May 1956. 

Alvarez (Alvarez et al. 1984) also estimated dose rates typical for the spectra 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); for photons >30 keV and <300 keV a dose rate of 0.86 mR/hr was assigned. 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 percent 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.

6.3.5

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

high-range, gamma-sensitive 1290 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 1963) 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 FEMP 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 NTA film 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 FEMP 

would be less than the MDL (Robinson 2001). 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,n 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 which produced and/or stored green salt include the Pilot Plant, Plant 4, 

Warehouse 4B and any other warehouse at Fernald for which the material stored is unknown. The 

neutron to photon ratio described below should only be applied to workers who were exposed to 

Uranium – Fluorine materials (i.e. UF4, UF6) at Fernald. 

FEMP Neutron Dosimetry

6.3.5.1 Development of the Neutron to Photon Ratio

Neutron exposures at Fernald were evaluated by Baker (1995). This survey was conducted in 

Warehouse 4B using bubble dosimeters to measure the neutron dose rate on UF4 (green salt) 

canisters of different enrichments. Table 6-9 presents the average measured neutron dose rate for 

depleted and low enriched uranium (1.25% – 2%). 

 Table 6-9. Measured neutron dose rates at Fernald (Baker 1995). 

Location Depleted Uranium

Dose Rate (mrem/hr)

Location Low Enriched Uranium

Dose Rate (mrem/hr)

1 0.085 3 0.1069

2 0.063 4 0.1069

5 0.080

Average 0.076 ± 0.012 Average 0.1069

 
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Measurements also were conducted for 10% enriched uranium; however, based on enriched material 

receipt information contained in the DOE Recycled Uranium Reports (DOE 2003), it was determined 

that the vast majority of the enriched material at Fernald was of approximately 1% enrichment. 

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

Unfortunately, Baker (1995) only measured the neutron dose rate of individual canisters. Photon 

surveys were conducted in 2001 on 56 drums of UF4 (Fernald 2001). 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 in order 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 since the measured general photon dose rate was below detection limits. 

After combining of the datasets, a distribution was fit to the data. The combined datasets resulted in a 

log-normal 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 found in Table 6-9 by 

the photon dose rate (Equation 6-1). 

Photon Dose Rate

Neutron Dose Rate Rn 

 / γ = (Equation 6-1) 

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

each constant dose rate value. Figure 1 depicts the frequency distribution of the simulated neutron to 

photon ratios for depleted and low enriched uranium. 

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 6-1. Comparison of neutron-photon ratio distributions for depleted and low 

enriched UF4. 
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Statistical parameters of the distributions depicted in Figure 6-1 are provided in Table 6-10. The 

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

standard deviation of 1.71 and an upper 95th% confidence of 0.23. 

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

Enrichment Geometric Mean GSD Upper 95th%

Depleted Uranium 0.07 1.74 0.17

Low Enriched Uranium 0.10 1.71 0.23

There are multiple approaches that 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 with in a 

process or storage area. The recorded photon dose is also a function of the quantity of uranium 

processed 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 claimant favorable simplification 

since 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 resulting 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 6-2 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 24.4 x 103 metric tons of uranium.

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-n reaction is directly proportional to the alpha 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 compared to depleted uranium and 

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

As can be observed from the Figure 6-2, most of the uranium work prior to about 1965 was natural 

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

received 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 

periods (1965-1975), work with enriched uranium initially increased such that by 1968 enriched 

uranium work constituted about 50-60% of the uranium work at Fernald. Following this peak, work 

with enriched uranium decreased to about 10-20% of the total uranium work at Fernald.
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 21 of 36

Year

1950 1960 1970 1980 1990

Relative Percentage of Uranium Recieved (%)

0

20

40

60

80

100

Total Uranium as a Percentage of Maximum Reciept year (1967) 

Percentage of Natural Uranium Recieved

Percentage of Enriched Uranium Recieved

Percentage of Depleted Uranium Recieved

Figure 6-2. Relative Percentage of Uranium Receipts by Material Category (Natural, 

Enriched, and Depleted). 

While the data in Figure 6-2 could be used to estimate annual neutron to photon ratios based on the 

percentage of enriched, natural, and depleted uranium work conducted, 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. Since this uncertainty cannot be properly tracked without detailed job history and material 

tracking information, and in order 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 claimant favorable assumption, 

given the limited data available.

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

green salt (especially since 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 warehouse for which the stored 

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

uncertainty in the origin of the photon dose. 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-6 Page 22 of 36

6.3.5.2 Workplace Neutron Radiation Fields

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

DOE Health Physics Manual of Good Practices for Uranium Facilities (Rich et. al 1988), 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 probability of causation calculations in IREP uses a Radiation Effectiveness Factor (REF) to 

estimate the probability of causation. The most claimant favorable neutron energy group (i.e., the 

neutron energy group with the largest REF) is the 0.1 – 2.0 MeV energy group. Since the neutrons at 

Fernald are assumed to be generated (born) in this energy group and although some fraction will 

scatter to a lower (less claimant favorable) energy groups, all of the neutron dose should be assumed 

to result from the 0.1 – 2 MeV energy group. In the absence of data, this is a reasonable and 

claimant favorable assumption. The default neutron dose fractions are given in Table 6-11.

Table 6–11. FEMP facility default neutron dose fractions. 

Facilities 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%

6.4 RECORDED DOSES

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

window and a single filter of approximately 1000 mg/cm2

. The FEMP historical dose record practices 

are given in Table 6-12. 

Table 6–12. Historical recorded dose practices. 

Year Dosimeter Measured Quantities Compliance Dose Quantities

1951-54

Film

OW-Open Window-mrep

S=Cd - mR

Skin=OW+S

WB=S

1954-85

Film

OW - mR

S - mrem

Skin=OW+S

WB=S

1985-present

TLD

Non-penetrating (Npen)

Penetrating (Pen)

Skin=Npen+WB

WB=Pen

6.5 ADJUSTMENTS TO RECORDED DOSE 

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

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

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

that the corrections identified 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 suggested. 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 
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filter was cadmium in both dosimeters, which attenuated the lower energy photons (see Table 6-8), 

and should have resulted in an underestimated response behind that filter for measured dose and 

Hp(10). Because most, but not all, penetrating radiations are above 30 keV, it is suggested that 

adjustments are necessary to satisfy claimant-favorable dose reconstruction criteria of pre-1985 

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

which include the L-X-rays from both uranium and thorium. It is estimated that a correction equal to 

10% of the <250 keV values given in Table 6-8 be added to the Hp(10) dose due to the contribution of 

these low energy photons to penetrating dose but are absorbed in the thick filter. 

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 provided 

in External Implementation Guide (NIOSH 2002) should be used. 

6.6 MISSED DOSE 

6.6.1 Missed Beta/Photon Dose

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

FEMP workers. This could have resulted when a dosimeter was lost or a worker was not monitored, 

or 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. The 

recommended procedure for missed dose is to assign with a log-normal distribution, with zero as a 

minimum value, (LOD/2 * the number of zero measurements) as the central tendency and (LOD * the 

number of zero measurements) as the upper 95% estimate. This procedure is applicable to both 

Hs(0.07) and Hp(10) since 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 6-13. The other values can be obtained from this data and the individual exposure 

record.

 

 Table 6-13. FEMP missed beta/photon dose. 

Years

MDL

(mrem/period)

Exchange

frequency

Product

(mrem/yr)

1951-53 40 Weekly 2,080

1954-58 30 Biweekly 780

1959-84 30 Monthly 360

1985-present 20 Quarterly 80

6.6.2 Missed Neutron Dose

It is not known for certain how the neutron dosimeters used by Baker (1995) were calibrated. The 

typical practice in 1995 was to use neutron flux to dose rate conversion factors in NCRP 38 (1971). 

Some DOE facilities, however, used a standard quality factor of 10 to convert from absorbed dose 

(rad) to dose equivalent (rem). For dose reconstruction purposes, the NCRP 38 doses must be 

converted to ICRP 60 values. If the dosimeters were calibrated based on NCRP 38, then the 

conversion factor from NCRP 38 (1971) to ICRP 60 (1990) is 1.91. If the dosimeters were calibrated 

to absorbed dose and a quality factor of 10 was applied then the conversion factor is 2. Since there is 

relatively little difference between these two values and additional research would be required to 

 
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determine which method was used for the calibration, the claimant favorable assumption using a 

factor of 2 should be applied to the neutron to photon ratio in order to convert the estimated dose to 

an ICRP 60 dose equivalent value. 

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 employees total photon dose would be 449 ± 82 mrem. Applying the low enriched 

neutron to photon ratio and the conversion to ICRP 60 (1990) radiation weighting factors would result 

in an estimated neutron dose of 89 mrem. The neutron dose distribution is a lognormal distribution 

with a geometric standard deviation of 1.77.

Summary of reasonable but claimant favorable assumptions: 

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

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

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

UF4 material

4. Assumption that all of the neutron energies are between 0.1 – 2 MeV

5. Assumption that neutron dosimeter used a quality factor of 10 to convert rad to rem. 

6.6.3

As noted previously (Dugan et. al 1981) in a response to a dosimetry assessment fact sheet, Dugan 

indicated that female employees at FEMP at times were not routinely monitored. This situation 

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

to postulate that there may have been other circumstances where women and men may have not 

been monitored. For unmonitored workers, 500 mrem per year will be assigned as an upper bound 

limit. This is several times above the mean doses observed for monitored workers. Since this dose is 

considered an overestimate, this upper bound will only be used in cases that will likely result in a 

Probability of Causation (PC) less than 50%. The applicable years this dose may be assigned are

shown in Table 6-14. 

Missed Dose for Unmonitored Employees

Table 6 –14. Missed dose for unmonitored employees. 

Years Dose to be Assigned for Unmonitored 

Employees (mrem/yr)

1951 – 1960

1969 – 1978

500

 

6.7 ORGAN DOSE

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

NIOSH (2002) to calculate the organ dose distribution. Reference NIOSH 2002 describes the 

methodology used to calculate organ dose using identified exposure geometries. Worker orientation 

is a primary consideration for this process; however, no definitive method is available to evaluate this 

factor. Therefore, Table 6-15 lists default values that are claimant-favorable.

Table 6–15. Organ dose default geometries. 

Claim Status Job Categories Exposure Geometry Percentage

 All Workers All AP 100
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6.8 BIAS AND UNCERTAINTY

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

FEMP workers during the years that film dosimeters were used. The FEMP film dosimeters were of 

the Pardue and Oak Ridge designs. A study performed by H. M. Parker in 1945 (Parker 1945) where 

he performed a comparison of the Hanford, Chicago and Oak Ridge dosimeters, he concluded that 

the agreement between the three programs “. . is considered to be satisfactory, especially with 

reference to the total exposure”. While this study is useful, it was all based on plutonium energies 

which other than for L-X-ray energies, the results are not directly comparable to Fernald radiations. 

For those radiation energies >250 keV, the two element dosimeter does a credible job and no 

corrections are required. Because there are energies <250 keV, corrections of 10% are 

recommended as described in Section 6-5. The uncertainty information for the FEMP Multi-Element 

Thermoluminescent dosimeter are presented in Table 6-16 is based on DOELAP accreditation data 

provided in an e-mail letter (Robinson 2003). 

Table 6 –16. Bias and uncertainty. 

FEMP Dosimetry System

Bias Magnitude and 

Range Uncertainty Factors

Overall 

Biasa

Range in 

Biasb Systematicc Randomd

Multi-element 

Thermoluminescent (1985-

Present)

1.01 0.85-1.18 1.06 1.2

a. Based on the distribution of energy levels and geometry judged most likely. Divide recorded dose by the table's bias 

value to calculate deep dose.

b. Range of overall bias factors based on alternative distributions of energy levels and geometry.

c. Systematic uncertainty resulting from lack of knowledge regarding actual distributions of energy levels and geometry.

d. Random uncertainty resulting from variation among workers in energy levels and geometry.
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Figure 6-3. Uranium-235 decay series (ref. PHS 1970). 
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Figure 6-4. Uranium-238 decay series (ref. PHS 1970). 
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Figure 6-5. Thorium-232 decay series (ref PHS 1970). 
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Figure 6-6. Beta dose vs. skin depth (Heatherton 1960).
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REFERENCES

ASI (Advanced Sciences Inc.), date unknown, document prepared for Westinghouse Materials 

Company of Ohio, Fernald, Ohio.

Alvarez, J. L., S. H. Daniel, L. O. Johnson, D. E. Martz, and B. L. Rich, 1984, Radiation 

Characterization and Radiological Equipment Evaluation at the Feed Materials Production 

Center, correspondence, HS-84-033, EG&G Idaho Inc. for National Lead of Ohio, Inc, Fernald, 

Ohio.

ANSI (American National Standards Institute), 1983, American National Standards for Dosimetry – 

Personnel Dosimetry Performance – Criteria for Testing, ANSI N13.11-1983, New York, New 

York.

Baker, A., (1995) Fernald Interoffice Memorandum entitled: Area Neutron Monitoring in the 4B 

Warehouse. April 7, 1995.

Boback, M. W., 1984, Radiation Exposure Problems (attachment to a letter to W. J. Adams titled 

“Plant 5 Radiation Exposure and Control Measures”), National Lead of Ohio, Inc., Fernald, 

Ohio.

Cooper, S. D., 1988, FEMP External Dosimetry Program Development History, ID-0001, Fernald 

Environmental Management Project, Fernald, Ohio.

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

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

DOE (U.S. Department of Energy), 2000, Recycled Uranium Project Report, Final Report, Ohio Field 

Office, Fernald Environmental Management Project, Fernald, Ohio.

DOE (U.S. Department of Energy), 2003, Recycled Uranium, United States Production, Enrichment 

and Utilization, DOE/SO-0003, Washington, D.C.

Dugan,T.A. to W.G.Tankersley, 1981, Response to Dosimetry Assessment Fact Sheet, National Lead 

of Ohio, Inc., Fernald, OH.

EPA (U.S. Environmental Protection Agency ), July,13, 2000, EPA Superfund Record of Decision 

Amendment, EPA/AMD/R05-00/161-2000, FEMP(USDOE),EPA ID: OH6890008976,OU 04, 

Fernald Environmental Management Project, Fernald, OH. 

Film Calibrations (documents in the ORAU FEMP database files), 1971, 1974, 1977.

Fernald (2001), Radiation Surveys of Bldg 38A (Chemical Warehouse). Survey # 01-10-07-0186

Green, L. E., 1981, “Leblonde Rapid Bore,” correspondence to M. W. Boback, Fernald Environmental 

Management Project, Fernald, Ohio.

Hayes, W. H., 1983, “Personnel External Dosimetry System Intercomparison,” DOE Contact Report to 

B. J. Davis, National Lead of Ohio, Inc., November 14.
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Heatherton, R. C., 1960, Estimation of Depth Dose from UX2 Beta Rays. 

Hinnefeld, S. L., 1989, “DOE Angular Dependence Study,” correspondence to D. Carlson, 

WMCO:QS(D):89-676, Westinghouse Management Company of Ohio, Fernald Environmental 

Management Project, Fernald, Ohio, October 06.

ICRP (International Commission on Radiological Protection), 1959, Report of Committee II on 

Permissible Dose for Internal Radiation, ICRP Report 2. 

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

ICRP Publication 60, Annals of the ICRP Vol. 21, Pergamon Press, Oxford, England.

ICRU (International Commission on Radiation Units and Measurements), 1988, Determination of 

Dose Equivalents from External Radiation Sources--Part II, ICRU Report 43, Bethesda, MD. 

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

in Radiation Protection Dosimetry, ICRU Report 51, Bethesda, MD. 

Johnson, W. A., and H. V. Heacker, 1963, Health Protection Review of National Lead Company of 

Ohio Letter Report, Atomic Energy Commission, Bethesda, Maryland.

Jones, K. L, 1988, “Extremity Dosimeter Field Calibration Studies at Westinghouse Materials Co. of

Ohio, Part 1,” correspondence, Westinghouse Management Company of Ohio, Fernald 

Environmental Management Project, Fernald, Ohio.

Lederer, Hollander, and Perlman, 1978, Table of Isotopes, 7th Edition.

NCRP, (1971) National Council on Radiation Protection and Measurements. Protection Against 

Neutron Radiation, NCRP Report 38, National Council on Radiation Protection and Measurements, 

Bethesda, Maryland.

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

Implementation Guidelines, OCAS-IG-001, Revision 1, Cincinnati, Ohio.

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

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

Parker, H. M., December 7, 1945, Comparison of Badge Film Readings at the Metallurgical 

Laboratories, Clinton Laboratories, and the Hanford Engineer Works, Hanford Record Center 

File No.7-3090, Letter to File, Richland, Washington.

Plato, P., and J. Miklos, 1982, Final Report on Development of a Personnel Dosimetry System, 

School of Public Health, University of Michigan.

PHS (U.S. Public Health Service), 1970, Radiological Health Handbook, Rockville, Maryland.

RAC ( Radiological Assessment Corp.), 1995, The Fernald Dose Reconstruction Project, Tasks 2 and 

3 Radiological Source Terms and Uncertainties, RAC Report No. CDC-5 
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Rich, B. L., Hinnefeld, S. L., Lagerquist, C. R., Mansfeld, W. G., Munson, L. H., Wagner, E. R. and 

Vallario, E. J., (1988) Health Physics Manual of Good Practices for Uranium Facilities, EGG-2530, 

UC-41, Idaho National Engineering Laboratory, EG&G Idaho, Inc., Idaho Falls, Idaho.

Robinson, S., 2002, FEMP External Dosimetry Technical Basis Manual, Revision 2, Fernald 

Environmental Management Project, Fernald, Ohio.

Robinson, S. D., Fernald DOELAP Performance, December 18, 2003, e-mail letter, Cincinnati, Ohio.

Unknown, 1988, A Closer Look at Uranium Metal Production, Feed Materials Production Center,

Fernald, Ohio, March.

Unknown, 2003, Action Plan for TLD System, Fernald Environmental Management Project, Fernald, 

Ohio.

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

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

Northwest Laboratory, Richland, Washington. 
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GLOSSARY

accreditation

Recognition that a dosimeter system has passed the performance criteria of the DOE 

Laboratory Accreditation Program (DOELAP) standard (DOE 1986) in specified irradiation 

categories.

accuracy

If a series of measurements has small systematic errors, they are said to have high accuracy. 

The accuracy is represented by the bias.

algorithm

A computational procedure.

Atomic Energy Commission

Original agency established for nuclear weapons and power production; a successor to the 

Manhattan Engineering District (MED) and a predecessor to the U.S. Department of Energy 

(DOE).

beta particle

A charged particle of very small mass emitted spontaneously from the nuclei of certain 

radioactive elements. Most (if not all) of the direct fission products emit (negative) beta 

particles. Physically, the beta particle is identical with an electron moving at high velocity.

claimant-favorable

Refers to the process of estimation based on technical considerations of the parameter 

significant to dose such that the estimated dose is not underestimated.

deep dose equivalent (Hd)

The dose equivalent at the respective depth of 1.0 cm in tissue.

dose equivalent (H)

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

The special unit is the rem. When D is expressed in gray, H is in sieverts (1 sievert = 100 

rem.)

DOELAP

The DOE Laboratory Accreditation Program (DOELAP) accredits DOE site dosimetry 

programs based on performance testing and onsite reviews performed on a 2-year cycle.

dosimeter

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

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

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

dosimeter, neutron film dosimeter, thermoluminescent dosimeter.

dosimetry system

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

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

dosimeters as well as interpretation of the results.
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DuPont 552

A film packet containing two pieces of film: a 502 sensitive film and a 510 insensitive film.

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

A measure expressed in roentgens of the ionization produced by gamma (or X) rays in air.

extremity

That portion of the arm extending from and including the elbow through the fingertips, and that 

portion of the leg extending from and including the knee and patella through the tips of the 

toes.

film

Generally means a “film packet” that contains one or more pieces of film in a light-tight 

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

using an optical densitometer. 

film density

See optical density. 

film dosimeter

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

filter

Material used to adjust radiation response of a dosimeter to provide an improved tissue 

equivalent or dose response.

gamma rays

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

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

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

not originate in the nucleus.

ionizing radiation

Electromagnetic radiation (consisting of photons) or particulate radiation (consisting of 

electrons, neutrons, protons, etc.) capable of producing charged particles through interactions 

with matter.

luminescence

The emission of light from a material as a result of some excitation.

Manhattan Engineering District (MED)

U.S. agency designated to develop nuclear weapons; a predecessor to the U.S. Department of 

Energy (DOE).

neutron

A basic particle that is electrically neutral weighing nearly the same as the hydrogen atom.
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nonpenetrating dose

Designation (i.e., NP or NPen) on film dosimeter reports that implies a radiation dose, typically 

to the skin of whole body, from beta and lower energy photon radiation.

open window

Designation on a dosimeter that implies the use of little or no shielding. It commonly is used to 

label the film response corresponding to the open window area.

optical density

The quantitative measurement of photographic blackening the density defined as 

D = log 10 (Io/I).

pencil dosimeters

A type of ionization chamber used by personnel to measure radiation dose. The results might 

be labeled “Pen” dose. Other names: pencil, pocket dosimeter, pocket ionization chamber.

penetrating dose

Designation (i.e., P, Pen, or Gamma) on Fernald film dosimeter reports that radiation.

Personal Dose Equivalent, Hp(d)

Radiation quantity recommended for use as the operational quantity to be recorded for 

radiological protection purposes by the International Commission on Radiological Units and 

Measurements. The Personal Dose Equivalent is represented by Hp(d), where d identifies the 

depth in millimeters and represents the point of reference for dose in tissue. For weakly 

penetrating radiation of significance to skin dose, d = 0.07 millimeter and is noted as Hp(0.07). 

For penetrating radiation of significance to “whole-body” dose, d = 10 millimeters and is noted 

as Hp(10).

photon

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

rad

A unit of absorbed dose equal to the absorption of 100 ergs per gram of absorbing material, 

such as body tissue.

radiation

One or more of beta, neutron, and photon radiation.

radioactivity

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

neutrons from unstable nuclei.

rem (roentgen equivalent in man)

A unit of dose equivalent equal to the product of the number of rads and the “quality factor” 

and any other modifying factors.

rep (roentgen-equivalent-physical)

Historically, the rep has been used extensively for the specification of permissible doses of 

ionizing radiations other that X-rays or gamma rays. Several definitions have appeared in the 

literature but in the sense most widely adopted, it is a unit of absorbed dose with a magnitude 

of 93 ergs per gram.
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roentgen

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

gamma (or X) rays that will produce a total charge of 2.58 x 10-4 coulomb is 1 kilogram of dry 

air. An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in 

soft tissue.

sievert 

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

skin dose

Absorbed dose at a tissue depth of 7 milligrams per square centimeter.

tissue equivalent

Term used to imply that the radiation response of the material being irradiated is equivalent to 

tissue. 

TLD chip

A small block or crystal made of lithium fluoride used in the TLD.

thermoluminescent

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

thermoluminescent dosimeter (TLD)

A holder containing solid chips of material that when heated will release the stored energy as 

light. The measurement of this light provides a measurement of absorbed energy, which can 

be related to dose through suitable calibration. The solid chips are sometimes called crystals.

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 centimeter (1,000 milligrams 

per square centimeter); also used to refer to the dose recorded.

X-ray

Ionizing electromagnetic radiation of extra-nuclear origin.