ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 73

DOE Review Release 10/21/2014

Argonne National Laboratory – East – 

External Dosimetry

ORAUT-TKBS-0036-6 Rev. 01

Effective Date: 10/16/2014

Supersedes: Revision 00 

Subject Expert(s): Eugene W. Potter

Document Owner 

Approval: Signature on File Approval Date: 10/02/2014

Vincent A. King, Document Owner

Concurrence: Signature on File Concurrence Date: 10/14/2014

Christopher M. Tornes, Objective 1 Representative

Concurrence: Signature on File Concurrence Date: 10/02/2014

Scott R. Siebert, Deputy Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 10/01/2014

James P. Griffin, Deputy Project Director

Approval: Signature on File Approval Date: 10/16/2014

Stuart L. Hinnefeld, 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-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 2 of 73

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

02/09/2006 00 New technical basis document for Argonne National Laboratory –

East – External Dosimetry. First approved issue. Training required: 

As determined by the Task Manager. Initiated by Norman D. Rohrig.

10/16/2014 01 Revision initiated to clarify and update document. Modified Section 

6.5.1 to clarify wording on blank and zero recorded values and badge 

exchange frequencies. Deleted former Section 6.5.3 on Missing 

Entry; requirements now incorporated in section 6.5.1. Modified 

Section 6.5.3 on Exposure Geometry (former Section 6.5.4) to add 

rotational and isotropic geometries for specified cancers. Modified 

Section 6.8.2 regarding application of a 1.6 correction factor for open 

window reading to obtain beta doses in millirem. Modified Section 

6.9 regarding the assignment of environmental dose. Incorporates 

formal internal and NIOSH review comments. Training required: As 

determined by the Objective Manager. Initiated by Vincent A. King.
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 3 of 73

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 6

6.1 Introduction .............................................................................................................................. 8

6.1.1 Purpose........................................................................................................................ 9

6.1.2 Scope........................................................................................................................... 9

6.2 External Dosimetry Overview ................................................................................................... 9

6.3 Interpreting the External Dosimetry Record............................................................................ 11

6.3.1 Dosimetry Records Systems....................................................................................... 14

6.3.2 Observed Data Discrepancies .................................................................................... 19

6.4 Historical Administrative Practices ......................................................................................... 19

6.4.1 Badged Population ..................................................................................................... 19

6.4.2 Badge Exchange Frequency ...................................................................................... 20

6.4.3 Field-Specific Calibration Factors ............................................................................... 20

6.4.4 Minimum Reported Dose............................................................................................ 21

6.5 Common Issues ..................................................................................................................... 21

6.5.1 Interpretation of Blanks and Gaps in Monitoring Records ........................................... 21

6.5.2 Discrepancies............................................................................................................. 23

6.5.3 Exposure Geometry.................................................................................................... 23

6.6 Photon Dose .......................................................................................................................... 23

6.6.1 Energy Groups ........................................................................................................... 23

6.6.2 Calibration Factor and Units for Dose Conversion Factors ......................................... 24

6.6.3 Missed Dose............................................................................................................... 24

6.6.3.1 Limit of Detection ......................................................................................... 24

6.6.3.2 Determination of Missed Dose ..................................................................... 25

6.6.3.3 Unmonitored Energy Range......................................................................... 25

6.6.4 Angular Dependence.................................................................................................. 26

6.6.5 Uncertainty ................................................................................................................. 26

6.6.5.1 Film.............................................................................................................. 26

6.6.5.2 Thermoluminescent Dosimeters................................................................... 27

6.7 Neutron Dose......................................................................................................................... 27

6.7.1 Energy Groups ........................................................................................................... 28

6.7.1.1 Neutron Energy Spectra............................................................................... 28

6.7.1.2 Reported Dose to Energy Groups ................................................................ 31

6.7.2 Calibration Factor and Units for Dose Conversion Factors ......................................... 32

6.7.3 Missed Dose............................................................................................................... 32

6.7.3.1 Limit of Detection ......................................................................................... 32

6.7.3.2 Unmonitored Energy Range......................................................................... 33

6.7.3.3 Neutron Dose Reconstruction Project .......................................................... 34

6.7.4 Angular Dependence.................................................................................................. 34

6.7.5 Uncertainty ................................................................................................................. 34

6.7.5.1 Film.............................................................................................................. 34

6.7.5.2 Thermoluminescent Dosimeter..................................................................... 35

6.7.6 Application of the Neutron Correction Factors ............................................................ 35
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6.8 Electron Dose ........................................................................................................................ 36

6.8.1 Energy Groups ........................................................................................................... 36

6.8.2 Calibration Factor and Units for Dose Conversion Factors ......................................... 37

6.8.3 Missed Dose and Limit of Detection............................................................................ 37

6.8.4 Angular Dependence and Exposure Geometry........................................................... 38

6.8.5 Uncertainty ................................................................................................................. 38

6.8.5.1 Film.............................................................................................................. 38

6.8.5.2 Thermoluminescent Dosimeter..................................................................... 39

6.8.6 Skin Contamination .................................................................................................... 39

6.9 Exposures Outside Radiologically Controlled Areas............................................................... 40

6.10 Extremity Dosimetry............................................................................................................... 40

6.11 Attributions and Annotations .................................................................................................. 41

References ........................................................................................................................................ 42

Glossary ............................................................................................................................................ 47

ATTACHMENT A, EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS.......................... 52

ATTACHMENT B, FACILITIES BY BUILDING NUMBER AND SECTION CODE.............................. 70

LIST OF TABLES

TABLE TITLE PAGE

6-1 Summary of beta-gamma film-processing history................................................................... 10

6-2 ANL-E external dosimetry history........................................................................................... 12

6-3 Default dosimeter exchange frequencies ............................................................................... 20

6-4 Historical dose limits ..............................................................................................................22

6-5 Dose limits based on exchange frequency ............................................................................. 22

6-6 Selection of photon energies for IREP.................................................................................... 24

6-7 Photon dose units for use with organ dose conversion factors ............................................... 24

6-8 Photon dosimeter LODs......................................................................................................... 25

6-9 Uncertainty for photon dose ................................................................................................... 27

6-10 Neutron dose assessment methods ....................................................................................... 28

6-11 Percent of NCRP Report 38 dose equivalent in the IREP energy groups at ANL-E 

facilities ..................................................................................................................................30

6-12 ICRP Publication 60 dose multipliers ..................................................................................... 31

6-13 Multipliers recommended to assign ICRP Publication 60 equivalent dose to the IREP 

neutron energy intervals......................................................................................................... 32

6-14 Neutron dose units for use with organ dose conversion factors.............................................. 32

6-15 Neutron dosimeter LODs........................................................................................................ 33

6-16 NTA film unmonitored dose correction factors........................................................................ 34

6-17 Uncertainty for NTA film and TLD neutron dose ..................................................................... 35

6-18 Beta dosimeter LODs............................................................................................................. 37

6-19 Beta dosimeter thickness and associated under-reporting ..................................................... 38

6-20 Uncertainty for beta dosimeters.............................................................................................. 39

B-1 Facilities by building number and section code ...................................................................... 70
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LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Neutron spectrum per unit lethargy and the integral spectrum of flux and dose equivalent 

for the LAMPF ER-1 spectrum ............................................................................................... 29

A-1 External Radiation Exposure Record ..................................................................................... 53

A-2 Radiation Exposure Record for Doses Assigned in CY 1989 or Later .................................... 55

A-3 Individual Meter Record ......................................................................................................... 56

A-4 Cumulative Exposure Record................................................................................................. 57

A-5 Accumulated Exposure Report............................................................................................... 58

A-6 Semi-Annual Accumulative External Exposure Report ........................................................... 59

A-7 Whole Body Radiation Exposure Summary............................................................................ 60

A-8 Master Radiation Exposure File ............................................................................................. 61

A-9 Master External Radiation Exposure Report........................................................................... 62

A-10 Neutron Exposure Report....................................................................................................... 63

A-11 Special Exposure Records ..................................................................................................... 64

A-12 Special Film Request .............................................................................................................65

A-13 Special Meter Assignment and Radiation Exposure Report ................................................... 66

A-14 Special TLD Finger Exposure Report..................................................................................... 67

A-15 Personnel Dose Equivalent Evaluation – Level 1 ................................................................... 68

A-16 Termination Occupational Exposure Report........................................................................... 69
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ACRONYMS AND ABBREVIATIONS 

a.u. arbitrary units

AGHCF Alpha Gamma Hot Cell Facility

ANL-E Argonne National Laboratory-East

ANL-W Argonne National Laboratory-West

APS Advanced Photon Source

ATLAS Argonne Tandem Linear Accelerator System

ATSR Argonne Thermal Source Reactor

amu atomic mass unit

AWA Argonne Wakefield Accelerator

CFR Code of Federal Regulations

Ci curie

cm centimeter

CP-1 Chicago Pile 1

CP-2 Chicago Pile 2

CP-3 Chicago Pile 3

CP-3ʹ Chicago Pile 3 Prime

CP-5 Chicago Pile 5

d day

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

EBR-II Experimental Breeder Reactor No. 2

EBWR Experimental Boiling Water Reactor

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

eV electron-volt

g gram

GeV gigaelectron-volt, 1 billion electron-volts

hr hour

HVEM High-Voltage Electron Microscope

in. inch

INEEL Idaho National Engineering and Environmental Laboratory

IPNS Intense Pulsed Neutron Source

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

LAMPF Los Alamos Meson Physics Facility

LINAC linear accelerator

LOD limit of detection

mA milliampere

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mm millimeter

mo month

mR milliroentgen
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mrem millirem

mrep millirep

MWSF Mixed Waste Storage Facility

n neutron

nC nanocoulomb

NCRP National Council on Radiological Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NRC National Research Council

NRTS National Reactor Testing Station

NTA nuclear track emulsion, type A

ORAU Oak Ridge Associated Universities

PIC pocket ionization chamber

PNNL Pacific Northwest National Laboratory

POC probability of causation

QF quality factor

qtr quarter

R roentgen

SRDB Ref ID Site Research Database Reference Identification (number)

SSTR solid-state track recorder

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

wk week

wR radiation weighting factor

yr year

ZGS Zero Gradient Synchrotron

ZPR-I Zero Power Reactor No. 1

ZPR-II Zero Power Reactor No. 2

ZPR-IV Zero Power Reactor No. 4

ZPR-IVʹ Zero Power Reactor No. 4 prime

ZPR-IX Zero Power Reactor No. 9

ZPR-V Zero Power Reactor No. 5

ZPR-VI Zero Power Reactor No. 6

ZPR-VII Zero Power Reactor No. 7

α alpha particle

γ gamma ray, particle, or photon

§ section or sections
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6.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA. 

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] 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 2010). 

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 2010):

• 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.
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6.1.1 Purpose

Argonne National Laboratory-East (ANL-E) was established on July 1, 1946; this technical basis 

document (TBD) covers the period from that date to the present. The work at the site has been a 

continuation of that done by the Metallurgical Laboratory of the University of Chicago beginning in 

1941, which was an Atomic Weapons Employer under EEOICPA. The job locations did not change 

until land and buildings were acquired for the laboratory.

The purpose of this TBD is to describe the external dosimetry systems and practices at ANL-E. This 

document discusses historical and current practices in relation to the evaluation of external exposure 

data for monitored and unmonitored workers. 

6.1.2 Scope

ANL-E, operated by the University of Chicago, has played an important role in the development of the 

U.S. nuclear program. ANL-E operations have involved research and development in nuclear 

reactors, high-energy physics, and nuclear materials. This TBD is part of the overall ANL-E Site 

Profile, which describes plant facilities and processes, historical information, and environmental data 

in relation to dose reconstruction for ANL-E workers. This document contains supporting 

documentation to assist in the evaluation of occupational external doses from these processes in 

accordance with OCAS-IG-001, External Dose Reconstruction Implementation Guideline

(NIOSH 2007). 

The methods and concepts of measuring occupational external doses to workers have evolved since 

the beginning of ANL-E operations. An objective of this document is to provide supporting technical 

data to evaluate the external ANL-E occupational doses that can reasonably be associated with 

worker radiation exposures under EEOICPA. These doses include occupational external exposures 

in ANL-E facilities and onsite exposures to ANL-E environmental releases. This document addresses 

the evaluation of unmonitored and monitored worker exposure and missed dose. Consistent with 

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

dose to account for current scientific methods and protection factors. 

In addition, this document presents the technical basis of methods for preparing ANL-E worker dose 

information for input to the NIOSH Interactive RadioEpidemiological Program (IREP). Information on 

measurement uncertainties is an integral component of the approach. This document describes the 

evaluation of uncertainty for ANL-E exposure and dose records. 

6.2 EXTERNAL DOSIMETRY OVERVIEW

Over the years, ANL-E used dosimetry badge designs that were typical for other government 

contractors of the respective periods, but the laboratory continued to use film dosimeters somewhat 

longer than at other sites. The initial design was a two-element badge incorporating a steel case, an 

open window, and a 1-mm cadmium filter. The same basic design was probably modified for wear on 

the wrist, head, or leg for extremity monitoring. No record of the type of beta-gamma film ANL-E 

initially used in these badges is available, but DuPont films were in use by the late 1950s. The 

badges eventually included Kodak nuclear track emulsion, type A (NTA) films for locations where 

neutrons were of concern. The beginning of neutron monitoring is uncertain, but it appears to have 

been in use at other sites as early as 1951. Indirect references to it at ANL-E begin around 1953. By 

the mid-1950s, some film badge rings (provided by a contractor) were in use on at least a trial basis to 

supplement wrist monitoring. 

Before March 1954, ANL-E processed films for itself as well as several universities. ANL-E then 

contracted the service to three commercial laboratories before resuming in-house processing in 1965 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 10 of 73

(Strom 1982). Table 6-1 summarizes beta-gamma film-processing history. Even during the period of 

contracted film service, there is some indication that an in-house capability was maintained, probably 

for emergencies, special studies, etc. The Atomic Film Badge Corporation was the only reported 

contractor that processed and read Kodak NTA film. Due to the labor necessary to read NTA film, it 

appears that films were developed but not routinely read before 1960. Films were apparently not 

evaluated unless there had been a measured gamma dose for the same period (Dolecek 1981). 

Table 6-1. Summary of beta-gamma film-processing history. 

Film processor Dates

Onsite 1942a–1954

Tracerlab Mar 1954–Jun 1958

Landauer Jul 1958–Jun 1960

Atomic Film Badge Corporation Jul 1960–Dec 1964

Tracerlab Jan 1965–Sep 1965

Onsite Oct 1965–1988

a. Film processing was in existence before the establishment of ANL-E 

in 1946. 

Around 1960, ANL-E transitioned to a four-filter film badge. It is uncertain whether the design was 

from its contractor, Atomic Film Badge Corporation, or was made to ANL-E specifications. The badge 

offered by the Atomic Film Badge Corporation had an open window backed by 0.060 in. of plastic, 

0.008 in. of copper (front and back), 0.024 in. of copper (front and back), and 0.032 in. of lead (front 

and back). The whole-body and wrist versions were identical except for the means of securing the 

badge to the body. A ring badge was offered with a pair of 0.032-in. lead filters, suitable for X-ray, 

and gamma radiation only. 

A new whole-body film badge design was put in place in 1962. The design was identical to that at the 

National Reactor Testing Station (NRTS) in Idaho and at the Hanford Site; it was referred to as the 

ANL or NRTS type. This badge featured a holder constructed of Tenite II and a slide insert. The 

holder contained three filtered areas and an open window. The filters were 0.0191 in. of aluminum, 

0.005 in. of silver, and 0.0395 in. (1 mm) of cadmium. The filters and an open window were on both 

sides of the film. Dose from photon energies greater than 150 keV, dose from photon energies less 

than 150 keV, and dose from beta radiation were interpreted by reference to various calibration 

curves (Strom 1982). The design was probably modified to be worn as an extremity monitor and 

incorporated Kodak NTA film when deemed necessary. Around 1970, Kodak films replaced DuPont 

films for beta-gamma monitoring. This badge design was used for whole-body dosimetry until being 

replaced by thermoluminescent dosimeters (TLDs) in 1988 to 1989. 

Work on TLDs for monitoring dose to the fingers began around the mid-1960s, and these dosimeters 

were in use by 1967 in at least some facilities. This ring design differed from the current design. As 

at most sites, the limitations of NTA film were recognized and work was started on methods to replace 

or supplement the film. A solid-state track recorder (SSTR), incorporating 235U foils and mica (to 

record track damage) was tested at the Zero Power Reactor (ZPR) facility. This method was used for 

the dose of record for a brief time, and was followed by the Hankins and Hosger albedo neutron 

designs using TLD-600 and TLD-700 chips. At least initially, NTA film was still added to the betagamma badges. The record is unclear about exactly when NTA film was discontinued. 

In 1984, ANL-E decided to change from film to an automated TLD system for beta-gamma dosimetry. 

The Panasonic system adopted by DOE in Idaho and used at ANL-West (ANL-W) was selected for all 

but the albedo neutron portion. The TLD portion consisted of three (natural abundance) lithium borate 

elements and one calcium sulfate element designated the Panasonic UD-814AS4. The system used 

the Panasonic UD-710A automatic dosimetry reader. Two of the lithium borate elements under thin 

filters (16 mg/cm2 and 58 mg/cm2

) are used to determine shallow dose equivalent. A lithium borate 

element under a thicker filter (600 mg/cm2 of plastic and aluminum) provides a flat energy response 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 11 of 73

for deep dose equivalent. A calcium sulfate element under the same thick filter provides an indication 

of low-energy photon radiation (Dolacek 1996). This dosimeter passed DOE Laboratory Accreditation 

Program (DOELAP) testing in May 1988. Although the Panasonic UD-814AS4 dosimeter is not used 

for quantitative neutron dose determinations, the response to neutron radiation of the natural lithium 

borate versus the response of the calcium sulfate provides a flag that the wearer may have been 

exposed to neutrons. An albedo dosimeter with Harshaw TLD-600 and TLD-700 chips in a “polybox” 

is inserted into the ANL-E holder for workers with the potential for neutron exposure (Dolacek 1996). 

There is no evidence that the ANL-E albedo system has been accredited by DOELAP. 

From the early days, Shonka-type pocket ionization chambers (PICs; both remote-reading and selfreading), often referred to in site documents as “dosimeters” or “pocket meters,” of various ranges 

were in use. Even though their readings were not considered to be the official dose, the readings 

were recorded in the individual dosimetry records through 1964. Film badges and PICs were stored 

at the entrances to the different facilities which were controlled for radiological purposes (Novak 1956 

and ANL 1973-1984). The locations where the dosimeters were used are generally included in the 

dose record (exceptions are noted below). When workers entered facilities for which they did not 

have an assigned film badge, they were assigned a “rover” dosimeter (generally if not always a PIC) 

(ORAUT 2005a). The rover dosimeter’s dose was then added to the film badge’s penetrating reading 

for the same wear period. 

Table 6-2 summarizes the ANL-E external dosimetry history. The periods of use are approximate. In 

some cases, changes were probably phased in over time or were implemented only in certain facilities 

or for certain groups. For example, TLD badges were already being used by security personnel by 

the summer of 1981 (Strom 1982). 

6.3 INTERPRETING THE EXTERNAL DOSIMETRY RECORD

When DOL requests a worker's dosimetry records, the ANL-E Chemical Engineering Division 

searches for the relevant records. For external dosimetry, there are three record sources that might 

contain the needed information (ANL-E records can also include those of the predecessor facility 

employees who became employees of ANL-E upon its establishment in 1946). The Worker 

Protection System is a database of employees between 1945 and the present. A second database 

(created in dBase III plus) contains the external dosimetry records for workers between 1989 and the 

present. Last, there are hard-copy records of the external doses for workers between 1942 and 1988. 

From these sources, a two-page External Radiation Exposure Record (summary) is developed (Luck 

2002). See Figure A-1 in Attachment A for an example. For early workers, only this summary report 

is provided initially. For workers with assigned dose in 1989 or later, a separate computer-generated 

summary is provided (Figure A-2) for those years. In response to requests for more detailed 

information, ANL-E has provided records in the various formats from the hard-copy files. The records 

generally provide data for comparison to occupational exposure limits in effect at the time. 
 

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Table 6-2. ANL-E external dosimetry history. 

Year Holder

Beta/gamma Neutron Extremity

Detector

Filtration

Processor Detector Processor Detector

Processor

Deep Shallow

1945a University of

Chicago (two

filter)

Film, type

Unknown

1 mm Cd Open

(28 mg/cm2

Wrapper)

ANL-E Film ANL-E

1946

1947

1948

1949

1952 “catastrophic” film 

1953 added Kodak NTA ANL-E Film rings

Ring, wrist, 

& 

head films

R.C. Scientific

& others

Wrist, head 

same as whole 

body

1954

1955 Tracerlab

1956

1957

1958 Landauer

1959 DuPont 553

1960 Four filter DuPont 558 0.008 in. 

Cu, 0.024 

in. Cu, 

0.032 in 

Pb, front 

and back

Open 

Window 

(0.060 in. 

plastic 

backing)

Atomic Film 

Badge

Corporation

Kodak NTA Atomic Film 

Badge

Corporation

1961

1962 Hanford Type

(four filter)

DuPont 558 0.0191-in.

Al

0.005-in.

Ag

0.0395-in.

Cd

Open

1963 Window ANL-E Kodak NTA ANL-E

1964

1965 DuPont 553

1966 DuPont 558

1970 Kodak Type 3 TLD ring ANL-E

1971 Kodak Type 2

1973 Kodak NTA 

& 

SSTR

1974

1975

1976

1977 Kodak NTA 

& 

albedo TLD

1978 Current TLD

1979– Ring

1987
 

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Year Holder

Beta/gamma Neutron Extremity

Detector

Filtration

Processor Detector Processor Detector

Processor

Deep Shallow

1988 –

present

INEEL design

(beta & gamma

only)

TLD

Panasonic 

UD814AS4 

(3 lithium borate &

1 calcium sulfate)

CaSO4 &

Li2B4O7

under

600 

mg/cm2

plastic &

Al

Li2B4O7

under

16 mg/cm2

 & 

58 mg/cm2

plastic

ANL-E TLD

Harshaw

TLD 

600/700

a. Film processing was in existence before the establishment of ANL-E in 1946. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 14 of 73

The pre-1989 record summaries (Figure A-1, two pages) report external dosimetry results as:

• Deep dose: Sum in millirem of penetrating photon plus neutron dose equivalent at a depth of 

1.0 cm in a 30-cm sphere of soft tissue with density of 1 g/cm3

. 

• Shallow dose: Dose equivalent in millirem at a depth of 0.007 cm in a 30-cm sphere of soft 

tissue with density of 1 g/cm3

. 

• An “M” for years in which the dose was less than the minimum sensitivity of the measuring 

device. 

• A hyphen (“-”) or “NM” for unmonitored periods; some records indicate unmonitored periods 

with a statement such as “Not monitored 1960-1962.” 

• An asterisk (“*”) after the dose when the only monitoring was by a direct reading dosimeter; 

additional codes used are explained on the report. 

The 1989 and later records summaries (Radiation Exposure Record; Figure A-2) report external 

dosimetry results as:

• Deep dose equivalent (DDE), lens of eye dose equivalent (LDE), shallow dose equivalent 

(SDE, WB), and shallow dose equivalent maximum extremity (SDE, ME). 

• “NM” meant not monitored. 

• “ND” meant not detectable. 

The reported values for the early years when the dosimeters were calibrated to exposure assumed 

that 1 mR is equal to 1 mrem. While this report (Figure A-2) is not directly useful for dose 

reconstruction, it provides a good cross check on the annual totals because it was compiled by 

someone familiar with the various records forms. 

All dosimetry results for a worker are provided as a package. Gaps in results that are not the result of 

a break in employment should be interpreted as unmonitored periods. Breaks in employment can be 

determined by a standard request for dose history in the individual’s file. For 1964 to 1975, only 

annual summary data for each individual are likely to be available (ANL 2005a). 

6.3.1 Dosimetry Records Systems

In the 1940s and 1950s, external dosimetry data were handwritten on preprinted forms. In 1961 the 

transition was made to a computer system and printouts replaced the forms. From 1964 to 1975, a 

handwritten form was again used and shows only annual totals. Computer reports are provided from 

1976 to the present. A separate neutron report and up to four reports on supplemental dosimeters 

issued could be included in some records. 

An individual worker’s records can contain one or more of the following forms and reports depending 

on the years of employment. The dates of use given here are approximate. Attachment A contains 

redacted samples; personal identifiers have been removed from each sample. 

Individual Meter Record (1946 to mid-1953) – Hand Entered (Figure A-3)

• Each card covers 1 week. 

• The units are roentgen rather milliroentgen. 
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• Daily readings and Total (weekly) from PICs, Times Worn, and a Discg column for entry of 

accidental discharge. 

• Film meter exposure values for shield and window for each day of the week. 

• Location of use (e.g., “Site A”). 

Cumulative Exposure Record (mid-1953 to 1961) – Hand Entered (Figure A-4)

• Results for each wear period. 

• The gamma units are milliroentgen (mr on the form), and the beta units are “arbitrary units” 

(a.u.; see Section 6.8.2). 

• From mid-1953 (fiscal year 1954) to 1956, the forms cover the fiscal year. From 1956 to 1961, 

they cover the calendar year. 

• Two-column format, so a year’s results can fit on one form. 

• Because dosimeters were stored at the entrances to facilities, the hard-copy records can 

contain multiple record sheets for the same period, each representing the accumulated dose 

at one facility or area. Locations are generally given as building numbers, sometimes 

preceded with a letter (e.g. “D”) meaning a building on the D site. Sometimes doses from 

additional facilities are on the same form and footnoted as shown in Figure A-4. 

• Zeroes represent a reading below the reporting threshold. Dashes mean that a badge was not 

read for that period. 

• For betas, when the change from a.u. to beta dose occurred (1961), the a.u. column heading 

was lined through and “Beta” was written next to it. 

• When a badge was not returned on time, an “N” was entered on the form. When the badge 

was read, the dose and a date (e.g., “R7/31”) were entered to the right of the “N” to indicate 

that the badge had been returned on July 31 and read late. 

• The forms were divided into 4 quarters of 13 weeks each. Each quarter was subtotaled. 

• The forms only recorded beta-gamma PIC readings, film penetrating dose (mr), and film 

nonpenetrating dose (a.u. or beta). Neutron doses from NTA film were recorded on the 

Neutron Exposure Report, see below. 

• Doses determined to be invalid or needing adjustment based on an investigation are footnoted 

and cross-referenced to a report explaining the change. In most cases, no administrative or 

notional doses were added for defective, fogged, or damaged film. However, in at least some 

cases, doses that were footnoted as invalid were added to the quarterly totals. 

Accumulated Exposure Report (1961 to 1962) – Computer Printout (Figure A-5)

• Results for each biweekly wear period shown in a single line. 

• Form includes entire calendar year. 

• Routine dosimeter (ROUT DOS), rover dosimeter (ROV DOS), and film (BETA, GAMMA, and 

NEUT) results are shown. Each of these results is totaled for the year. 
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• Biweekly total (BIWK TOTAL) and running totals for the calendar year (CAL YR) and last 

14 weeks (QTD TOTAL) are given. 

• The gamma units are milliroentgen, the beta units are millirad, and the neutron units are 

millirem. 

• Section codes are used instead of building numbers (see Attachment B for a list of section 

codes). Up to three section codes can be displayed for each period. 

• Less than the minimum reported values are shown as zeros. 

• It could not be determined how unmonitored (by film badge) periods are shown, as every 

record appears to show 26 periods per year. Zeros in the section code column might indicate 

that the individual did not have a badge for that period. 

Semi-Annual Accumulative External Exposure Report (1963 to 1964) – Computer Printout 

(Figure A-6)

• One form contains approximately 6 months of data. 

• The gamma units are milliroentgen, the beta units are millirad, and the neutron units are 

millirem. 

• The location where the badge was worn is in the LOC IDENT column. This field is a 

combination of the location code (see Attachment B for a list of codes) and the employee’s 

IHS number. 

• For 1963, results for each biweekly film dosimeter are shown in a single line. There are two 

lines below the film badge for each weekly dosimeter (i.e., PIC) result (WK DOS). For 1964, 

the entries for the weekly dosimeters were eliminated. 

• For 1963, routine dosimeter (WK DOS), rover dosimeter (R DOS), and film (BETA, GAMMA, 

and NEUT) results are shown. For 1964, there is no entry for routine dosimeters. 

• Totals (penetrating) are shown for each period (TOTAL). 

• Less than the minimum reported values are shown as zeros. 

• Unmonitored (by film badge) periods are not shown on the form (i.e., a line of data was not 

created for the unmonitored period). 

Whole-Body Radiation Exposure Summary (1964 to 1975) – Hand Entered (Figure A-7)

• Only annual totals are presented for Rover Dosimeter, Film Badge (Beta, Gamma, and 

Neutron), and Penetrating Total (sum of Rover Dosimeter, Gamma, and Neutron). 

• The units are millirem. 

• PIC readings used routinely are not included. 

• Section or building assignments are not included. 

Master Radiation Exposure File (1976 to 1988) – Computer Printout (Figure A-8)

• For 1976 and 1977, only the right-hand side of the form was copied for inclusion in the hardcopy records. The worker’s name is cut off, but the Social Security Number shows. 
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• Results for each monthly period are shown in a single line. 

• Column headings are two-digit year and month (YM), location code (OHS SEC), count of 

rovers (ROV CNT), rover reading(s) (ROV DOS), film doses (BETA, GAM, NEUT), penetrating 

dose (PEN; sum of ROV DOS, GAM, and NEUT), and Remarks (RE). A common remark was 

“99,” indicating a dosimeter that was returned late. 

• The gamma units are milliroentgen, the beta units are millirad, and the neutron units are 

millirem. 

• Totals are provided for the calendar year and moving quarter. Because only the annual report 

is included in the record, the moving-quarter value is always the fourth-quarter total. 

Master External Radiation Exposure Report (1989 to present) – Computer Printout (Figure A-9)

• This report coincides with the start of TLD monitoring as the dose of record for the whole body 

and the first results from the DOELAP-accredited system. 

• Results for each monthly period are shown in a single line. 

• Column headings are location code (SECTION), BADGE TYPE, ISSUE DATE, RETURN 

DATE, SHALLOW DOSE, DEEP DOSE, NEUTRON DOSE, PENETRATING TOTAL (sum of 

DEEP DOSE and NEUTRON DOSE), and REMARKS. Remarks include codes such as “99” 

(late), “NR” (not returned), and “IN” (investigate). 

• For 1989 through 1991, the employee’s name and dosimetry identification number (which was 

also the ANL-E employee number) appear above the issue and return date. Starting in 1992, 

columns for each wear period were added for LAST NAME and MASTERID. The MASTERID 

is a combination of the wear period (YYMM), the location code, and the dosimetry 

identification number. 

• The units are millirem. 

• Shallow dose is set to the higher of the measured value or deep value. 

• Totals are provided for the calendar year. 

• A note in a heading states that a zero in the neutron column should be ignored if the badge 

type is not BGN or BGNC. In other words, a zero does not imply that the worker was 

monitored for neutrons unless a neutron dosimeter was issued. 

Neutron Exposure Report (1960 to 1961) – Hand Entered (Figure A-10)

• Multiple wear periods and results could be on one form. 

• The units are millirem. 

• At times, the neutron dose equivalent appears not to have been carried over to be added to 

penetrating dose unless the limit of detection (LOD) was exceeded. In other cases, neutron 

dose equivalents as low as 5 mrem were added. 

Special Exposure Records (1953 to 1960) – Hand Entered (Figure A-11)

• The records contain supplemental dosimeter readings (e.g., an extra badge worn on wrist or 

body). 
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• The units are milliroentgen and a.u. 

• Multiple film badges and dosimeters and wear periods could be shown on one form. 

• Information includes DATE, TYPE (film badge or dosimeter), AREA, TIME (worn), POSITION, 

READING (including mR under filter and a.u.), and comments (might explain why worn). 

Special Film Request (1958 to unknown) – Hand Entered (Figure A-12)

• This form recorded a request for special film badge and documents the results. 

• It appears to overlap in time with the Special Exposure Records (Figure A-11), but was used 

for one badge only. 

• Includes DATE, TIME WORN, AREA WORN (building), POSITIONS WORN, TYPE OF FILM 

WORN (beta-gamma or neutron), and remarks. The remarks can indicate the reason for the 

badge. 

• The units are milliroentgen and a.u. for beta-gamma, and millirem for neutron. 

Special Meter Assignment and Radiation Exposure Report (1961 to 1987) – Hand Entered 

(Figure A-13)

• The records contain supplemental dosimeter readings (e.g., a beta-gamma badge worn on the 

right wrist). 

• This form probably replaced the Special Film Request (Figure A-12). At least four variations 

were noted that corresponded to the dosimeters in use over the years, which were the twoelement beta-gamma film, four-element beta-gamma film, NTA film, and albedo TLDs. 

• The units are typically millirad for beta, milliroentgen for gamma, and millirem for neutron as 

listed on the forms. 

• One form was created for each badge worn. 

• Densities under filters and the open window as well as resultant exposures are documented. 

Special TLD Finger Exposure Report for BLDG. [number] (1978 to 1988) – Hand Entered 

(Figure A-14)

• This form appears to coincide with the start of the current finger dosimeter because it is 

designed for up to four readings. 

• Multiple wear periods are on the same form. 

• Fields shown are Type of Work, Begin Date, End Date, TLD Readings (1 to 4), and Exposure 

βλ, λ, and Neutron. The type of work was most frequently a location rather than a description. 

The gamma and neutron columns appear not to have been used. 

• Either two or four TLD readings were shown. No notation was included to indicate that more 

than one ring was being worn in the case of four TLDs. 

• The units are millirem or millirad (according to the column heading). 

Personnel Dose Equivalent Evaluation – Level 1 (1960-unknown) – Hand Entered (Figure A-15)

• Reported exposures above certain thresholds were routinely investigated (Bleiler 1964). 
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• This record documents the results of investigations for the monthly wear period. All 

evaluations for a particular area were documented on one form. 

• The primary purpose of the form was to bring abnormal doses to the attention of the 

individuals and the area health physicists and to encourage corrective measures. 

• Contains the period worn, LEVEL in millirem, TYPE, Valid (yes or no), and ANTICIPATED (yes

or no). There were four exposure types. Type A was a dose of greater than 300 mrem/mo to 

the whole body, head and trunk, gonads, active blood-forming organs, and lens of the eye. 

Type B was 900 mrem to the skin, other organs, tissues, and organ systems. Type C was 

1,800 mrem/mo to the forearms and ankles. Type D was 4,500 mrem/mo to the hands and 

feet. 

• Unanticipated and invalid doses required additional documentation. 

• There also was a Level 2 evaluation, with four higher-level A, B, C, D types (900, 1,500, 3,000, 

7,500 mrem/mo, respectively). No examples of the Level 2 evaluation were found. 

Termination Occupational Exposure Report (1979 to unknown) – Hand Entered (Figure A-16)

• Form prepared when a worker’s employment ended during certain years. 

• Provides the beginning and ending employment dates in blocks 8 and 9, respectively. 

• The external dose in millirem is in block 11. 

Explanations of invalid data were also sent to individual files. For example, the results for all but one 

group of badges worn between June 12 and July 9, 1964, were declared invalid due to exposure 

during shipment. The individuals’ detail cards were punched with code “89,” and zero doses were

recorded. 

6.3.2 Observed Data Discrepancies

As can be seen from the descriptions in the previous section, some of the periods of use of the 

various forms overlap. Care should be taken not to count the doses for these periods twice. 

The assumption favorable to claimants is to include discrepant data in the annual total unless there is 

some explanation in the record as to why they should not be included. 

6.4 HISTORICAL ADMINISTRATIVE PRACTICES

6.4.1 Badged Population

In 1956, the site Radiation Safety Guide (Novak 1956) indicated that areas were posted and that 

individuals entering an active area (i.e., a radiologically controlled area) were required to wear 

personnel monitoring devices that were provided at the entrance. Areas that were spelled out as 

active included Building 330 (Chicago Pile 5; CP-5) and Building 211 (cyclotron and small Van de 

Graaff accelerator). Wing G of Building 203 (large Van de Graaff) was not an active area, but access 

was forbidden during operation and restricted at other times. In addition, permanent monitors for fast 

neutrons were installed. Wrist badges with neutron film were required at CP-5 (Novak 1956). 

In the 1973 to 1984 revisions of the site Health and Safety Manual (ANL 1973–1984), a radiation area 

was defined as an area where the dose (equivalent) to an individual in any calendar quarter could 

exceed 300 mrem, where radioactive materials were stored in quantity, or where equipment that 
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produced ionizing radiation was operated. Consistent with the earlier Radiation Protection Guide, 

each person who entered a radiation area was required to wear a personal monitoring device. 

Assigned film badges and self-reading dosimeters were provided at the entrance to each area and 

were to be returned to the designated storage area at the end of the work shift (ANL 1973–1984). 

6.4.2 Badge Exchange Frequency

At first there were weekly film badge exchanges. On April 16, 1951, the badge exchange frequency 

changed from weekly to biweekly for Site D and West Stands personnel. Although not stated, this 

change probably included Site A personnel. The only film badges still being developed weekly were 

for reclamation personnel (i.e., the site decontamination group) and 20 special cases at the new 

Chemistry Building at the University of Chicago (Mendelstein 1951). 

The 1961 specifications (scope of work) for film badge services indicated that the exchange frequency 

was to be biweekly (Strom 1982). During 1967 to 1974, badges were exchanged approximately 

biweekly (26 exchanges per year) (Bleiler 1968a, 1968b, 1970, 1973a, 1974). 

Due to a work force reduction in the dosimetry group, biweekly exchanges were eliminated and all film 

badges were placed on a 4-week exchange schedule as soon as possible (Bleiler 1973b). It appears 

that this change was made starting with the first exchange in 1974, which covered December 28, 

1973, to January 24, 1974 (Bleiler 1975a). Starting in 1975, exchanges were monthly according to 

Bleiler (1975b). This does not agree with “Information Concerning NIOSH Requests” (Luck 2002), 

which states that monthly exchanges started in October 1965. This discrepancy could indicate that 

some groups were on a monthly exchange at least temporarily in the mid-1960s. 

In 1981 and 1982, the exchange frequency was monthly (Strom 1982). A summer 1981 study at 

ANL-E indicated that a quarterly TLD badge exchange was anticipated (Strom 1982). There is no 

evidence that this was implemented for the general population. 

The Advanced Photon Source [APS] Accelerator Systems Safety Assessment Document states that 

personnel monitors were exchanged quarterly at that facility (ANL 1994). 

Dose reconstructors should use dosimeter exchange frequencies reflected in the record when 

available. In the absence of such records, Table 6-3 provides default dosimeter exchange 

frequencies that are favorable to claimants to be used for dose reconstructions. 

Table 6-3. Default dosimeter exchange frequencies. 

Years General population Others

1946–1959 Weekly Site D and West 

Stands (and likely 

Site A) personnel 

switched to biweekly 

in 1951

1960–1973 Biweekly

1974 4-Week (13 periods/yr)

1975–present Monthly (12 periods/yr) 1994 APS quarterly

6.4.3 Field-Specific Calibration Factors

No workplace-specific calibration factors have been found. 
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6.4.4 Minimum Reported Dose

The specifications for the multielement film badge (Hanford type) indicated that the response range of 

the film badge was to be 0.025 R to 3,000 R (Strom 1982). In 1982, measured doses of less than 

15 mR were reported as zero (Strom 1982). 

The specifications for the Atomic Film Badge Corporation multielement badge indicate that a lower 

limit of 25 mR was reported for X-ray, beta, and gamma. Amounts less than 25 mR were shown but 

not added to the cumulative total. All neutron doses above zero were reported. Neutron tracks were 

counted in 25 random fields (Strom 1982). No information on minimum reported doses was 

recovered for the other early vendors. 

A 1980 survey indicated that the lower dose thresholds below which results were considered 

background were 15 mrem for nonpenetrating and penetrating radiation and 50 mrem for neutrons 

(Neal 1980). 

The algorithm for the Panasonic UD-814AS4 beta-gamma TLD (the current dosimeter) indicates that 

the minimum reported dose equivalents are 15 mrem deep (penetrating) and 30 mrem shallow 

(nonpenetrating). However, ANL (1996) also states that the reporting level is 10 mrem for deep dose 

equivalent, and this appears to be the current value. The minimum reporting level for neutrons is 

10 mrem. However, the described dosimeter is not the Panasonic UD-808AS that was tested by 

DOELAP in 1992, but rather a Hankins-type (one TLD-600 and one TLD-700 in a polybox) 

(Dolacek 1996). 

6.5 COMMON ISSUES

6.5.1 Interpretation of Blanks and Gaps in Monitoring Records

Prior to 1964

Records of individual dosimeter results before 1964 might include blanks rather than recorded zeros 

for some periods. Because all workers entering radiologically controlled areas are assumed to have 

been badged, intermittent blanks on such records should be interpreted as zeros values, and missed 

dose should be assessed using the frequencies established in Table 6-3. The exchange frequency 

should be assumed to be weekly (52 exchanges per year) before 1960 and biweekly (26 exchanges 

per year) from 1960 through 1963. This is not applicable to workers whose occupations did not 

require entry to radiologically controlled areas and for whom dose records do not exist (see Section 

6.9).

Gaps in records (such as a missing year of records for an otherwise monitored worker) may also exist 

in some cases. Dose should be assigned for such periods using dosimetry data before and after that 

period in consideration of the approach of Watson et al. (1994). This method applies to otherwise 

monitored individuals whose occupations were not likely to have changed during the period.

1964 through 1975

The available dosimetry records do not provide individual dosimeter results from 1964 to 1975, but 

instead provide only annual dose summaries (see Figure A-7; as mentioned above, there is some 

overlap in the dose records for 1964; therefore, some detail for 1964 might be available for some 

workers). For these cases, it is necessary to estimate the dosimeter exchange frequency from 1964 

to 1975 for years in which positive or zero doses are recorded. Blank values for a year on this 

summary form indicate no monitoring was performed for that year. It appears that ANL-E did not 

generally use different exchange frequencies in relation to job categories, so assumption of a single 
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frequency by period is reasonable. Table 6-3 lists the default dosimeter exchange frequencies. For 

1964 to 1973 the exchange frequency should be assumed to be biweekly (26 exchanges per year). 

For 1974 the exchange frequency should be assumed to be every 4 weeks (13 exchanges per year). 

For 1975 the exchange frequency should be assumed to be monthly (12 exchanges per year). 

If the number of zero measurements cannot be determined, the missed dose evaluation becomes 

more complex. When only the annual dose is known, the number of zero doses should be estimated 

based on the dose level and the monthly, quarterly, or annual limits for that year and the number of 

possible zero monitoring intervals. This would be the situation, for example, if an individual received a 

cumulative dose of 2,000 mrem in a given year at a facility that had a monthly monitoring frequency 

and a maximum permissible exposure limit of 5,000 mrem/yr. The minimum number of months in 

which this dose could have been received is 5. Therefore, the maximum number of missed dose 

months would be 7, and the minimum would be 0 because the dose could have been received evenly 

throughout the year. The central estimated number of months should be the median (4), but the 

upper bound would be 7 (NIOSH 2007). Table 6-4 lists historical dose limits. 

Table 6-4. Historical dose limits.

Period Daily or weekly limit Quarterly or annual limits:

1946–1955 Beta & gamma, 100 mrem/d

(Nickson 1946; Gilbreath ca. 1946)

Not applicable

1956–1972 Gamma 300 mrem/wk (AEC 1958)

Beta & gamma (sum), 500 mrep/wk 

(Novak 1956)

3 rem/qtr, 15 rem

5(N-18) rem accumulated (AEC 1958) 

1973–1988 Not applicable 3 rem/qtr (ANL 1973–1984)

5 rem/yr 5(N-18) (DOE 1981)

1989–1992 Not applicable 5 rem/yr annual effective dose equivalent 

(DOE 1988)

1993–present Not applicable 5 rem/yr total effective dose equivalent 

(DOE 1992)

Table 6-5 divides these dose limits into exchange frequencies. Using the methodology of NIOSH 

(2007), it is possible to develop an estimate of the number of zeros for calculation of the missed dose

that is favorable to claimants. 

Table 6-5. Dose limits (rem) based on exchange frequency. 

Years Limit (rem) Period (yr)

52 26 13 12 4

Weekly Biweekly 4-weekly Monthly Quarterly

1946–1955 0.1 ~260 0.5 (a) (a) (a) (a)

1956–1972 0.3 52 0.3 0.6 (a) (a) (a)

1973–1988 3 4 (a) 0.462 0.923 1.0 (a)

1989–present 5 1 (a) (a) (a) 0.417 1.25

a. This exchange frequency was not routinely used during the given period.

After 1975

Determining the number of zero readings should be straightforward beginning in 1976 since records 

are of better quality and individual cycle monitoring results are included in the record.

As with the period before 1964, if there are gaps in records (which is less likely to occur for the more 

recent period), dose should be assigned using dosimetry data before and after that period in 

consideration of the approach of Watson et al. (1994). This method applies to otherwise monitored 

individuals whose occupations were not likely to have changed during the period.
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6.5.2 Discrepancies

If the employee’s record contains discrepancies, it is favorable to claimants to use the higher dose in 

the dose reconstruction. Care must be taken to interpret dose numbers properly if units were not 

specified. At first, ANL-E routinely reported exposures in roentgens, but since mid-1953 the 

laboratory has routinely used milliroentgen and millirem. It is highly unlikely that a record would show 

a dose greater than the quarterly or annual limit without an additional record that indicates an 

overexposure. 

If no date is associated with a dose record or if it is illegible, it is favorable to claimants to use that 

dose in the dose reconstruction. The dose reconstructor should use best judgment to credit the dose 

to the most likely year. 

Corrections were noted in the dose record by lining through the incorrect information. If the record 

was updated and noted, the correction should not be applied again. If there is no obvious notation to 

indicate the incorporation of a correction, the action most favorable to claimants is to incorporate the 

correction in the dose to be used for reconstruction. 

6.5.3 Exposure Geometry

Because little information is available on the exposure geometry for an individual, the standard 

assumption is that all exposures are for anterior-posterior geometry, which is favorable to the claimant

for most organs. In accordance with the NIOSH External Dose Reconstruction Implementation 

Guideline (NIOSH 2007), rotational and isotropic geometries must also be evaluated in accordance 

with that document for bone surface, red bone marrow, esophagus, or lung exposures. The favorable 

to claimant geometry from this comparison should be used for these organs unless workplace 

information indicates that anterior-posterior geometry is applicable for a specific exposure situation. 

6.6 PHOTON DOSE

6.6.1 Energy Groups

The NIOSH IREP POC program contains three photon energy ranges: below 30 keV, 30 to 250 keV, 

and above 250 keV (NIOSH 2007). Separation of the dose from each energy band is necessary. 

Very little spectroscopy data have been found to indicate gamma spectrum in ANL-E work areas. To 

estimate the gamma spectrum to which workers were exposed, facilities were grouped into 

categories. From the information in Attachment B and the worker’s external dosimetry records, the 

dose reconstructor can make an estimate of the energy band that is favorable to the claimant. 

Plutonium facilities have mostly 17-keV photons with some 59-keV photons from 241Am. ORAUTOTIB-0017, Interpretation of Dosimetry Data for Assignment of Shallow Dose (ORAUT 2005b), 

suggests that a correction factor of 0.6 should be applied to the nonpenetrating dose that is 

determined from early film dosimeters and that nonpenetrating doses should be assigned as photons 

less than 30 keV if the employee worked with or around plutonium.

Reactor facilities have dispersed fields of higher energy photons from fission as well as fission and 

activation products. Radioactive materials handling and processing facilities included a wide variety 

of activities. For the purpose of this analysis, the worst case was assumed to be handling of 

irradiated fuel. The gamma energy ranges were therefore taken to be the same as for reactors. For 

electron accelerator facilities [APS, electron linear accelerator (LINAC)], bremsstrahlung photons 

dominate the secondary radiation field. Thick shields of concrete or other materials result in photons 

in the MeV energy range. For proton or positive ion accelerators [Zero Gradient Synchrotron (ZGS), 

Intense Pulsed Neutron Source (IPNS), and Van de Graaff], neutrons generally constitute the greatest 
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hazard. Thick shields that are designed for the neutron hazard eliminate all but the most energetic 

photons. With shielding and safety interlock systems, exposure to the direct beam of ANL-E 

accelerators was rare, and the personal exposure records should document them. However, 

maintenance personnel were exposed to activated accelerator components during repairs, target 

changes, and so forth. One reaction of particular importance is the thermal neutron capture of sodium 

in the concrete of the accelerator shielding as discussed in National Council on Radiation Protection 

and Measurements (NCRP) Report 144 (NCRP 2003). This reaction [23Na(n,γ)24Na] produces a 

radioactive isotope that decays with a 15-hour half-life by emitting gamma rays of 1.4 and 2.8 MeV. 

Therefore, the gamma energy band for accelerators was judged to consist primarily of photons with 

energies above 250 keV. Radiation from X-ray machines and some radioisotopes presented lowenergy photon hazards, but these appear to have been used generally in conjunction with higher

energy sources rather than in distinct facilities. Table 6-6 lists the photon energy range percentages. 

Table 6-6. Selection of photon energies for IREP. 

Facility type

(see Attachment 6B)

Energy band 

(keV) Percentage

Plutonium facilitiesa <30 75

30–250 25

>250 0

Reactors <30 0

30–250 25

>250 75

Radioactive materials <30 0

30–250 25

>250 75

Accelerators <30 0

30–250 10

>250 90

a. Shallow doses can be determined using open window readings 

for the early film badges and the methods in ORAUT (2005b). 

6.6.2 Calibration Factor and Units for Dose Conversion Factors

The early dosimeters were calibrated in roentgens (Strom 1982). It is reasonable to assume that this 

continued until calibration of the Panasonic TLD dosimetry system with DOELAP sources at Pacific 

Northwest National Laboratory (PNNL). The personal dose equivalent Hp(10) is the appropriate unit 

to use for this period. Table 6-7 lists the dose units to use for organ dose conversion factors. 

Table 6-7. Photon dose units for use with 

organ dose conversion factors. 

Years Dose unit

1946–1988 R

1989–present Hp(10)

6.6.3 Missed Dose

Section 2.1.2 of NIOSH (2007) recommends the use of the LOD/2 method for determining missed 

dose. 

6.6.3.1 Limit of Detection

The film badge that was initially developed at the University of Chicago was used at other DOE sites. 

All of these badges used X-ray film in a metal badge holder, and each had an open window and an 

area covered with 1 mm of silver, tin, or cadmium (ORAUT 2006). Information from ANL-E indicated 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 25 of 73

that a cadmium filter was used. A PNNL study of this two-element dosimeter identified a detection 

level of about 40 mR at the upper 95% confidence level for radium gamma radiation (Wilson et al. 

1990). An improved film implemented at Hanford in 1960 (Wilson et al. 1990) reduced this detection 

level to about 15 mR. Limited information is available on the film types and when changes occurred. 

This information has been summarized in Table 6-2. In 1956, the two-element film badge was 

capable of measuring doses from 50 mR to 500 R (Novak 1956). The specifications for the 

multielement film badge (Hanford type) indicated that the low end of the exposure range was to be 

0.025 R (Strom 1982). The LOD of the current Panasonic TLD system is 10 mR for photons. 

Table 6-8 lists the recommended photon dosimeter LODs for the ANL-E dosimeters. 

Table 6-8. Photon dosimeter LODs. 

Years LOD

1946–1959 50 mrem

1960–1988 25 mrem

1989–present 10 mrem

6.6.3.2 Determination of Missed Dose

Determination of missed dose is performed using LOD/2 times the number of zero readings (NIOSH 

2007, Section 2.1.2.2). Except for the period from 1964 to 1975, the number of zero readings can be 

determined directly from the dosimetry data (Section 6.5.1). The missed dose is assumed to have a 

lognormal distribution with central tendency nLOD/2, where n is the number of zero readings. The 

upper 95% dose is nLOD. If the number of zero readings cannot be determined, it must be estimated 

on the assumption that prorated dose limits were not exceeded. Section 6.5.1 of this document and 

Section 2.1.2.3 of NIOSH (2007) discuss this estimate. In this case, the estimate is assumed to have 

a lognormal distribution with central tendency mLOD/2, where m is the median of minimum and 

maximum possible number of zero readings. The upper 95% dose is pLOD, where p is the maximum 

possible number of zero readings. 

6.6.3.3 Unmonitored Energy Range

The two-element film dosimeter ANL-E used was similar to those at other sites. The response of this 

dosimeter is addressed in ORAUT-TKBS-0003, Savannah River Site (ORAUT 2005c), which 

addresses the significant over-response of film to photons of low energy. Dosimeter calibration was 

performed on the open window using radium or 60Co. Therefore, no missed photon dose correction 

factor is appropriate for this dosimetry system. 

The multielement film dosimeter at ANL-E provided better energy response to measure worker dose 

more accurately. Multiple calibration sources were used to characterize the energy response of the 

badge. A sophisticated procedure was developed to interpret the density readings under the open 

window and filters. It appears that corrections were incorporated to prevent missed photon dose. 

Therefore, no missed photon dose correction factor is appropriate for this dosimetry system. 

With the implementation of the Panasonic TLD system, standard DOELAP sources have been used 

for the development of the algorithm and testing. Automated readout has eliminated many of the 

human variables associated with film development and interpretation. Modern data processing has 

allowed individual TLD element correction factors to be stored. Therefore, this system is unlikely to 

have missed photon dose in the energy range to which workers could be exposed. Therefore, no 

missed photon dose correction factor is appropriate for this dosimetry system. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 26 of 73

6.6.4 Angular Dependence

The film dosimeters at ANL-E had various angular responses. Dosimeters were not always exposed 

normally, which resulted in variant responses in relation to actual worker exposure. 

For a parallel beam of photons, the film dosimeters would have experienced an apparent decrease in 

dose with increasing angle from normal incidence due to the greater distance of travel in the filter. 

However, at large angles, there would have been an increase in response because photons would be 

able to expose the film under the filter without passing through it. No data was found for the film 

dosimeters to quantify these effects. 

Quantitative information is available for the Panasonic UD-814AS4 dosimeter (Dolacek 1996). For 

five DOELAP exposure categories, element responses generally decreased as the angle increased. 

This effect was more pronounced at lower photon energies. However, lower energies also caused an 

over-response at small angles. For angles of incidence up to 30°, the ratio of reported dose to 

delivered deep dose ranged from about 0.8 to 1.2 for photons. 

There is insufficient data to identify an angular dependence correction to apply to any of the 

dosimeters. Dose reconstructors should follow current NIOSH guidance. 

6.6.5 Uncertainty

NIOSH (2007) describes methods for quantification of laboratory uncertainty associated with reading 

film and TLDs. These methods provide a statistical treatment of the variability associated with reading 

dosimeters in the laboratory. 

6.6.5.1 Film

ANL-E used film to measure photons between 1946 and 1988. The DuPont 558 film packet (with the 

sensitive 508 film) was used in 1960. The 508 film was the successor to 502 film, and the National 

Research Council (NRC) reported that both films had a useful range from 10 or 20 mR up to 

approximately 10 R (NRC 1989). It is not clear when ANL-E started using 502 film or when it changed 

to 508 film. Hanford changed to 508 film in 1960 (Wilson et al. 1990). Both film types have 

approximately the same reading uncertainty. 

The method in NIOSH (2007) was used to determine the laboratory uncertainty (upper 95% 

confidence dose) for film readings. NRC detailed the method in Film Badge Dosimetry in Atmospheric 

Nuclear Tests (NRC 1989). The discussion of this method cites sensitivity parameters for 502 film. 

The spreadsheet developed for ORAUT-TKBS-0011-6, Rocky Flats Plant – Occupational External 

Dose (ORAUT 2010), was modified with parameters specific to ANL-E. Review of dosimetry 

worksheets indicated that optical density readings were recorded to the nearest 0.01 density units 

(ANL 1966). Therefore, the densitometer reading uncertainty was assumed to be ±0.005 density unit. 

Reviewing ANL-E optical density-to-dose conversion charts from 1962 (Strom 1982, four-element 

badge, reading under cadmium for photons with energy above 150 keV), it was possible to determine 

film sensitivity (0.92 density units/R) for the four-element badge. However, the calibration curve did 

not extend past 200 mR. A curve for the two-element badge extended the range to 5 R (Strom 1982). 

The film sensitivity was estimated to be 0.71 density units/R for the two-element badge. Using these 

parameters, the upper 95% confidence doses for various dosimeter readings were calculated for each 

of the badge types. 

Although the uncertainty is smaller at higher exposures, the NRC methodology recognizes that 

additional uncertainty from variability in calibration, film processing, and reading the calibration curve 

prevents the upper 95% confidence dose from falling below 120% of the reported exposure. This 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 27 of 73

limitation was applied here and affects the estimate of the upper 95% confidence dose above 24 mR 

for the two-element badge and 18 mR for the four-element badge. Because of the 120% limit, the 

upper 95% confidence doses are the same above 24 mR. Table 6-9 lists the upper 95% confidence 

doses. 

Table 6-9. Uncertainty for photon dose. 

Dose

(mR)

Upper 95% confidence photon dose (mR)

DuPont 508 film

two-element badge

1946–1959

DuPont 508 film

four-element badge

1960–1988

DOELAP-accredited TLD

1989–present

10 15 14 20

20 25 24 31

50 60 60 64

100 120 120 120

200 240 240 240

500 600 600 600

1,000 1,200 1,200 1,200

2,000 2,400 2,400 2,400

6.6.5.2 Thermoluminescent Dosimeters

TLDs provided improved photon dosimetry. ANL-E replaced film whole-body dosimeters with TLDs in 

1988 when DOELAP performance testing began. According to the ANL-E Technical Basis Document 

for External Dosimetry (Dolacek 1996), the current beta-gamma TLD (Panasonic UD-814SA4) 

responds within 10% of the delivered dose equivalent for most single radiation fields and within 20% 

for the mixed radiation fields that were DOELAP-tested. The largest positive deviations were for 

mixtures of X-rays and beta particles. The few negative deviations for mixed fields were less than 

10% below the delivered dose. These data are for radiation normally incident to the badge (Dolacek 

1996). 

The standard deviation of the null readings for the DOELAP-accredited Panasonic dosimeter was not 

documented. Therefore, the analysis used the Simplified Dosimetry Uncertainty calculation 

recommended by NIOSH (2007, Section 2.1.1.3.3) where the critical level Lc is the LOD estimated in 

Section 6.6.3.1 of this document. The percent standard error at high air kerma was estimated to be 

10%. This assumption is favorable to claimants for this dosimetry system. Table 6-9 lists the upper 

95% confidence doses. 

6.7 NEUTRON DOSE

There were and are a number of facilities at ANL-E with the potential for neutron exposures. The 

record is not clear concerning early neutron monitoring at ANL-E, but boron-lined PICs were reported 

at most facilities before the early 1950s (NIOSH 2007). It is also unclear from the program 

documentation exactly when monitoring with Kodak NTA film began at ANL-E. Because the initial 

mission of the site was research on nuclear reactors, where neutrons were a recognized hazard, it 

seems reasonable that ANL-E would have been one of the first sites to use NTA film. It was in use by 

1956 and was referenced in the site Radiation Safety Guide (Novak 1956). NTA film continued in 

general use until the late 1980s when albedo TLD dosimeters replaced it (ORAUT 2005a). 

By 1972, Kodak Type 2 film was in use for beta-gamma determinations. According to Strom (1982), a 

procedure that was first written in 1967 and revised in August of 1972 (Procedure 120.0) provided a 

method to determine thermal neutron dose equivalent from the beta-gamma film by comparing the 

open window area and the area under the cadmium filter. Because thermal neutron exposures were 

normally only present in conjunction with high-energy gamma exposures, any extra darkening under 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 28 of 73

the cadmium filter was assumed to be from neutron-gamma reactions in the filter. This method 

required some knowledge of the work being performed during the wear period (Strom 1982). 

In 1972 and 1973, studies of the SSTR neutron dosimeter were undertaken at ZPR-VI where there 

was a concern that the moderated plutonium light element neutrons that were being encountered 

during maintenance operations were not being detected by the NTA film dosimeters. It appears that, 

after a testing period, official doses were recorded from these dosimeters (Dolecek 1977). 

ANL-E switched to TLDs for dose of record in 1989, but some neutron doses were determined using 

TLD albedo dosimeters before this time. A Hankins albedo dosimeter was first employed at the ZPR 

facility around 1977. This dosimeter was followed by an in-house design termed the Hosger albedo 

dosimeter that contained two 6

Li (TLD-600) and two 7Li (TLD-700) elements. It was noted that both of 

these types required knowledge of the neutron energy spectrum to accurately assess the dose 

equivalent. To avoid underestimating the dose equivalent, an overly conservative conversion factor 

was used (Strom 1982). It appears that ANL-E planned to implement a CR-39 track etch dosimeter in 

1989, but there is no indication this occurred. Table 6-10 summarizes the methods ANL-E used to 

assess personnel neutron doses. The dates are approximate. 

Table 6-10. Neutron dose assessment methods.

Type Dates Comment

NTA film 1953–1988 Only capable of reliably detecting neutrons of greater 

than approximately 800 keV

Kodak Type 2 beta-gamma film 1971–1988 Used to assign thermal neutron dose in the presence of 

high-energy gamma doses

SSTRs 1973–1975 Contained 0.5 g U-235, discontinued due to fissile

material accountability and personnel dose concerns

Albedo dosimeters 1977–1982 Hankins type designed for use at the ZPR facility

1982–1989 Hosger, a four-element ANL-E design (2 TLD-600 and 

2 TLD-700), similar in appearance to a self-reading 

pocket chamber

1988–present ANL-E design, TLD-600 and -700 in a polybox inserted 

in the Panasonic dosimeter, similar to a Hankins type

6.7.1 Energy Groups

The measured neutron dose must be divided into energy groups consistent with the dose conversion 

factors in Appendix B of NIOSH (2007). These energy groups and the associated radiation weighting 

factors (wR) from International Commission on Radiological Protection Publication 60 (ICRP 1991) 

are:

Energy range (MeV) wR

<0.01 5

0.01–0.1 10

>0.10–2 20

>2–20 10

>20 5

6.7.1.1 Neutron Energy Spectra

The research did not discover any comprehensive studies of the neutron energy spectra at ANL-E, 

but some facility-specific information was found. Most of this data was in the form of summary 

statements about the average energy, effective energy, or energy range of the neutrons. Some of the 

measurements were not appropriate for dose reconstruction because they were made under unusual 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 29 of 73

operating conditions or in normally unoccupied areas. The ANL-E information was supplemented by 

other examples in the literature for similar facilities. 

Neutrons are a significant concern at only a few facilities at ANL-E. These include reactors, highenergy accelerators, and materials-handling areas where alpha emitters could interact with light 

elements. Facilities such as the 60-Inch Cyclotron, the Electron LINAC, and other low-energy 

accelerators do not present a neutron hazard outside of their shielding under normal operating 

conditions (Coulson, Elwin, and Freeman 1989). Neutrons rarely contribute more than 20% to the 

annual collective dose at accelerator facilities (Coulson, Elwin, and Freeman 1989). 

The ZGS was a 12.5-GeV proton accelerator that operated between 1963 and 1979. No useful 

information about spectra from ANL-E for potentially occupied areas was found. Figure 6-8 in 

ORAUT-TKBS-0010-6, Los Alamos National Laboratory – Occupational External Dose (ORAUT 

2013), provides neutron spectra unfolding results for the ER-1 target at the Los Alamos Meson 

Physics Facility (LAMPF), which uses an 800-MeV proton beam. Given the lack of better information, 

the SPUNIT program result was chosen to simulate neutron spectra at the ZGS. A spectrum was 

generated to visually match that in Figure 6-8 of ORAUT (2013). The Ing and Makra (1978) 

parameterization for the NCRP Report 38 (NCRP 1971) flux-to-dose conversion factor was used to 

generate the energy spectrum of dose. The resulting neutron spectrum per unit lethargy (logarithm of 

energy) is shown in Figure 6-1 along with the integral spectrum of flux and dose equivalent. The 

fraction of dose equivalent in each of the IREP neutron energy intervals (Table 6-11) was determined 

from this spectrum. 

Figure 6-1. Neutron spectrum per unit lethargy and the integral spectrum of flux and dose equivalent 

for the LAMPF ER-1 spectrum.

LAMPF ER-1 Neutron Spectrum

0.0

4.0

8.0

12.0

16.0

20.0

1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07 1.E+08 1.E+09

Energy (eV)

Flux/Lethargy

0.0

0.2

0.4

0.6

0.8

1.0

Integral Fraction Flux and Dose Equiv.

flux/lethargy

frac flux

frac dose
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 30 of 73

The 500-MeV IPNS is capable of generating neutrons in the energy range from 0.025 eV to greater 

than 20 MeV. In 1982, outside the biological shielding at the facility, the energy levels of potential 

neutron exposures ranged from 0.1 to 1 MeV (Strom 1982). Low dose equivalent rates, on the order 

of 0.5 mrem/hr, were observed on the roof area due to suspected neutrons streaming through cracks. 

Spectral measurements made with Bonner spheres near the second-floor offices and first-floor shops 

indicated average neutron energies of 0.1 to 0.5 MeV. Skyshine was a problem before 1989 when 

additional shielding was added above the enriched 235U target (Coulson, Elwin, and Freeman 1989). 

To establish the dose equivalent fractions in the IREP neutron energy intervals, the spectrum from 

LAMPF was modified to eliminate neutrons above 30 MeV. At the IPNS, the proton beam bombards 

a thick (greater than the proton range) uranium target which, through cascade and inelastic 

processes, reduces the energy of the neutrons. At LAMPF and at the ZGS, a thin target (less than 

the proton range) was used so that high-energy particles emerged from the target. The resulting dose 

equivalent fraction estimates for the IPNS are also shown in Table 6-11. 

Table 6-11. Percent of NCRP Report 38 (NCRP 1971) dose equivalent in the IREP energy groups at 

ANL-E facilities. 

Energy Group

Percent of dose equivalent

ZPRs

CP-2, CP-3, CP-5, 

Janus,

Argonaut,

Juggernaut ZGS IPNS

Plutonium

chemistry

& metallurgy

No

information 

available

<0.01 MeV (wR=5) 1 20 2 4 0 0

0.01–0.10 MeV (wR=10) 8 5 2 3 10 0

>0.10–2.0 MeV (wR=20) 65 50 9 19 50 100

>2.0–20 MeV (wR=10) 26 25 30 61 40 0

>20 MeV (wR=5) 0 0 57 13 0 0

The Argonne Tandem-Linear Accelerator System (ATLAS) is a 17-MeV/amu heavy-ion accelerator. 

Direct neutron dose rates are low under normal experimental conditions. However, access is 

restricted when accelerating certain light ions. No skyshine has been observed (Coulson, Elwin, and 

Freeman 1989). No spectral information was found. 

At the 4.5-MeV Dynamitron, accelerating light ions produces neutrons in accessible areas. A rate of 

0.2 mrem/hr/mA of deuterons outside of the experimental area was reported (Coulson, Elwin, and 

Freeman 1989). No spectral information was found. 

A 1966 shielding study at the 12-MeV Tandem Van de Graaff Accelerator used 6.5 MeV as the 

effective energy to determine the flux-to-dose equivalent conversion. However, the authors 

understood that this energy was an overestimate because the spectrum would have been degraded in 

the potentially occupied areas that were being measured (Dyer and Mundis 1966). 

The APS is an electron-positron accelerator that has been in full operation since 1995. Electrons are 

accelerated to an energy of 650 MeV in a LINAC and then raised to an energy of 7 GeV in a booster 

synchrotron. The electrons are then injected into a large-diameter storage ring. Calculations and 

measurements by ANL-E personnel indicate that dose rates are low and that there are no measurable 

neutron dose equivalents outside the shielded areas (DOE 2003; ANL 1994; Moe et al. ca. 1994). 

Some of the highest neutron exposures probably occurred at the ZPR facilities (see Attachment B). 

The ZPRs were large critical assemblies fueled with uranium or plutonium. Spontaneous fission from 

240Pu, other isotopes, and (α,n) reactions in light elements led to significant doses near the reactors 

even when shut down. The reactors were arranged in two halves that were split for loading and to set 

up experimental apparatus. Dose equivalent rates of 300 to 400 mrem/hr were seen between the 

halves. Even with temporary shielding in place, personnel between the halves experienced significant 

dose equivalent rates, especially when working near a gap in the shielding, which was necessary to 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 31 of 73

load fuel elements and perform other work. Although the fission spectrum was somewhat moderated 

by the components of the reactor, the flux of neutrons in the intermediate energy ranges was 

significant. There were many different ZPR configurations at ANL-E (and some at ANL-W). This 

analysis uses the ZPR-6 Assembly 7, but the energy spectra should not differ significantly among the 

ZPRs. A plot of the NCRP Report 38 (NCRP 1971) dose equivalent rate for the measured spectrum 

from neutrons above a cutoff energy as a function of the cutoff energy was examined (Yule 1971), 

and the fraction of the dose equivalent in the IREP energy intervals was then estimated. The results 

are in Table 6-11. 

Some other reactors that have operated at ANL-E (CP-2, CP-3, CP-5, Janus, Argonaut, and 

Juggernaut) had similar neutron spectra to the Materials Test Reactor at the Idaho National 

Engineering and Environmental Laboratory (INEEL) (Lennox and Kelber 1956; Folkrod, Moon, and 

Saluja 1961). Therefore, the recommendations in ORAUT-TKBS-0007-6, Idaho National Laboratory 

and Argonne National Laboratory-West – Occupational External Dosimetry (ORAUT 2011), are 

adopted for use here. Similarly, the recommendations in the Los Alamos TBD (ORAUT 2013) are 

adopted for plutonium chemistry and metallurgy. Finally, if no information is available about where the 

exposure occurred, or if it occurred in multiple facilities, it is recommended that all of the dose 

equivalent be assigned to neutrons in the >0.10- to 2.0-MeV energy interval. Table 6-11 presents 

these results. 

6.7.1.2 Reported Dose to Energy Groups

NIOSH (2007) requires that the neutrons be apportioned into the energy groups in ICRP 

Publication 60 (ICRP 1991). Because little neutron spectra information was available, the analysis 

assumed that the reported neutron doses were calculated using the quality factors (QFs) from NCRP 

Report 38 (NCRP 1971). The Report 38 QFs were averaged over the ICRP Publication 60 energy 

groups. A dose multiplier was then calculated for each energy group for conversion from the 

Report 38 dose equivalents to Publication 60 equivalent doses. Table 6-12 lists the dose multipliers 

for the energy ranges. 

Table 6-12. ICRP Publication 60 dose multipliers 

(ICRP 1991). 

ICRP 60 energy group 

(MeV)

Average

NCRP 38 QF

ICRP 60

dose multiplier

<0.01 (wR=5) 2 2.5

0.01–0.1 (wR=10) 5 2.0

>0.10–2 (wR=20) 10 2.0

>2–20 (wR=10) 7.5 1.3

>20 (wR=5) 4.0 1.2

The ICRP Publication 60 dose multipliers from Table 6-12 were used with the estimated neutron 

spectra in Table 6-11 to develop multipliers that can be used to determine the neutron equivalent 

dose for each IREP energy interval. The corrected neutron dose that is reported in the claimant’s 

dose record should be multiplied by these factors to determine the ICRP (1991) neutron dose for each 

neutron energy interval. Since 1960, when neutron doses began to be recorded routinely, the QFs for 

thermal and fast neutrons were usually 3 and 10, respectively (Dolecek 1981). Thermal neutron 

doses could have been measured by neutron PICs or beta-gamma film badges. In general, however, 

it is not possible to separate the thermal and higher-energy dose equivalents in the records. Energies 

above 800 keV were measured with NTA film. Corrections of the NTA film for neutrons between 

thermal and 800 keV are discussed separately. The current albedo dosimeters respond to all 

energies but are less sensitive to high energies. Table 6-13 lists the recommended multipliers to be 

applied to the reported neutron doses. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 32 of 73

Table 6-13. Multipliers recommended to assign ICRP Publication 60 equivalent dose to the IREP 

neutron energy intervals.

Energy Group

Multipliers used to assign to IREP neutron energy intervals

ZPRs

CP-2, CP-3, 

CP-5, Janus,

Argonaut,

Juggernaut ZGS IPNS

Plutonium

chemistry

& metallurgy

No 

information 

available

<0.01 MeV (wR=5) 0.025 0.5 0.05 0.1 0 0

0.01–0.10 MeV (wR=10) 0.16 0.1 0.04 0.06 0.2 0

>0.10–2.0 MeV (wR=20) 1.3 1.0 0.18 0.38 1.0 1.91

>2.0–20 MeV (wR=10) 0.34 0.33 0.39 0.79 0.52 0

>20 MeV (wR=5) 0 0 0.68 0.16 0 0

6.7.2 Calibration Factor and Units for Dose Conversion Factors

Early calibration information is limited, but ANL-E appears to have calibrated neutron film with 

unmoderated plutonium-beryllium (PuBe) neutron sources by 1958 (Strom 1982; Tedeschi 1958). A 

curve of neutron film correction factors versus effective neutron energy determined with a DvorakDyer sphere was published in a 1965 study of the 50-MeV injector of the ZGS (Steele 1965). At an 

energy of 800 keV, the correction factor for a PuBe calibration was 1.6. At 2-MeV effective energy, 

the correction factor was about 1.2, which suggests that the neutron dose equivalent for fission 

spectrum and below should be multiplied by an average of 1.4 to account for the PuBe calibration. 

This correction for the measured spectrum is accounted for as part of the unmonitored energy range

correction discussed in Section 6.7.3.2. 

The reported doses for the SSTRs and interim albedo dosimeters that were used before 1989 are 

assumed to represent ambient dose. During 1976 to 1977, SSTRs and Hankins albedo dosimeters 

compared favorably with time-motion studies at the ZPR facility (Dolecek 1977). In many cases, NTA 

film was worn in conjunction with these dosimeters. 

ANL-E adopted the Panasonic TLD system in use at ANL-W for beta-gamma dosimetry in 1989. 

However, a different albedo neutron insert was used that was not based on the ANL-W Panasonic 

TLDs. This albedo design, using Harshaw TLD-600 and TLD-700 chips, has not been accredited by 

DOELAP. The site’s TBD for external dosimetry (Dolacek 1996) states, “Calibration of the ANL-E 

albedo dosimeter has been conducted at various locations both on-site and off-site,” and goes on to 

state the LOD for exposure to moderated 252Cf. These statements appear to indicate the albedo TLDs 

at ANL-E were calibrated with DOELAP exposure standards in preparation for DOELAP performance 

testing. Therefore, the deep dose equivalent Hp,slab(10) is appropriate for this dosimeter. Table 6-14 

lists the dose units for use with organ dose conversion factors. 

Table 6-14. Neutron dose units for use with 

organ dose conversion factors. 

Years Neutron dose unit

1946–1988 Ambient dose, H*

(10)

1989–

present

Deep dose equivalent, Hp,slab(10)

6.7.3 Missed Dose

6.7.3.1 Limit of Detection

The LOD for NTA film was not well documented, but most references are to 50 mrem (Steele 1965). 

It appears that, even when beta-gamma film processing and reading was contracted, ANL-E generally 

processed and read the neutron films (Strom 1982). An exception was between 1960 and 1964 when 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 33 of 73

Atomic Film Badge Corporation processed the NTA films. There is some indication that this 

contractor reported all read doses for films with more tracks than the control films (Strom 1982). It is 

not clear what this minimum reporting level was, but it could have been below 50 mrem. When betagamma film was used to determine thermal neutron dose, the reported minimum was 50 mrem. 

Doses less than this value were interpreted as gamma exposures, which Procedure 120.0 claimed 

was conservative (Strom 1982). This also appears to have been the reporting threshold for neutron 

dose equivalent from Kodak NTA film from 1965 to 1972 (Strom 1982). 

No LOD was reported for the SSTRs (Yule, Armani, and Gold 1972; Yule 1973). The LOD was 

probably less than 50 mrem, but is assumed to be the same as NTA film for dose reconstruction 

purposes. 

Of the interim albedo dosimeters, only the Hosger dosimeter has a documented LOD, which was 

20 mrem at an average energy of 2.4 MeV (Strom 1982). Because the LOD would be lower for the 

lower energy spectra likely to be encountered by most workers, this value is favorable to claimants for 

these dosimeters. 

In 1989, ANL-E started using an albedo TLD inserted into the Panasonic holder for the dose of record. 

The stated LOD for this system is 10 mrem (Dolecek 1996). Table 6-15 lists the neutron dosimeter 

LODs. The dates are approximate. 

Table 6-15. Neutron dosimeter LODs. 

Type Dates LOD

NTA film 1953–1988 50 mrem

Kodak Type 2 beta-gamma film 1971–1988 50 mrem (thermal neutron dose)

SSTRs 1973–1975 50 mrem

Albedo dosimeters 1977–1982 Hankins 20 mrem

1982–1989 Hosger 20 mrem

1988–present 10 mrem

6.7.3.2 Unmonitored Energy Range

NTA film is a poor detector of neutron energies below 500 to 800 keV (Griffith et al. 1979; Wilson et al. 

1990). In a 1965 study at the 50-MeV injector of the ZGS, ANL-E recognized that dosimetry results 

below an effective energy of 1 MeV (determined with a 12-in. Dvorak-Dyer paraffin sphere) should 

have a correction factor applied to the PuBe calibration. Below 300 keV, this correction factor was 

estimated to be so large that other methods of determining dose equivalent would be necessary. In 

addition, the Steele (1965) stated that chronic low-level exposures in areas where personnel were

sometimes located were not recorded by the film. 

A 1966 investigation pointed out that the majority of neutrons in areas that were occupied by 

personnel at the 4.5-MeV and 2-MeV Van de Graaff accelerators had energies between 100 keV and 

1 MeV. NTA film measurements in the facilities showed dose equivalents of less than 5% of the 

values from instrument measurements (Till 1966). 

In 1973, a workforce reduction in the dosimetry group led to a proposal to eliminate NTA film with the 

exception of the ZGS facility, which could indicate that NTA film that had been deployed to other 

areas generally produced only background results. A response by site health physicists pointed to an 

example of a measurable neutron exposure at the Tandem Van de Graaff facility. In another 

example, SSTRs at the Building 316 ZPR facility detected neutrons where NTA film did not. The 

decision to discontinue NTA film was reversed at least partially due to accidental criticality 

considerations (Bleiler 1973b). 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 34 of 73

In spite of the recognized shortcomings of NTA film, documentation was not found to indicate that any 

compensating measures were taken by the external dosimetry program. Before 1989, dose from 

neutrons below approximately 800 keV was not reliably detected. To determine the magnitude of 

potentially missed dose, the correction factors in Table 6-16 were estimated based on experience. 

Factors for the ZPR and plutonium facilities should only be applied when the record indicates work 

was performed in those facilities. If the exact work location is unknown, but the record indicates that 

no work was performed with plutonium (e.g., participation in bioassay monitoring for radionuclides 

other than plutonium or americium), a factor of 1.5 would be favorable to claimants for other facilities. 

Table 6-16. NTA film unmonitored dose correction 

factors. 

Facility

NTA film unmonitored 

dose correction factor

ZGS, IPNS 1.25

Other accelerators 1.5

CP-3, CP-3ʹ, CP-5 reactors 1.5

ZPR Pu-fueled reactors 4

Pu-handling facilities 4

The ANL-E TLD albedo neutron dosimeter systems were calibrated using variously moderated 

spectra. These dosimeters over-respond to the lower (relative to the calibration spectrum) energy 

spectra most likely in operational areas. There is no need for a neutron dose correction for an 

unmonitored energy range for these dosimeters. 

6.7.3.3 Neutron Dose Reconstruction Project

A 1982 survey (Strom 1982) indicated that, similar to Rocky Flats, not all neutron films that were 

developed were read. Before 1960, neutron films were apparently only read if the gamma dose was 

100 mrem or more. This correction should be accounted for by adding missed dose for each zero 

reading where there is an indication that neutron monitoring occurred. 

6.7.4 Angular Dependence

A 1965 study of neutron field characteristics at the 50-MeV proton injector of the ZGS indicated that 

NTA film response decreased from 84% of the calculated dose equivalent when the film was 

perpendicular to the source to 57% when the film was at 90°. The effective energy of the neutrons 

was 1.18 MeV, and the dose equivalent rates were high (174 rem/hr). The contribution of scattered 

neutrons to the dose equivalent at the point of measurement was determined to be 24% with an 

effective energy of 0.49 MeV (Steele 1965). The location of these measurements was not in an area 

occupied by personnel during operation of the accelerator. Due to multi-scattering, NTA film response 

in the normally occupied areas was likely to be less angular dependent. Therefore, with the possible 

exception of accidental exposures in high-dose areas, no correction for angular dependence is 

deemed necessary. 

No information is available for the albedo TLDs ANL-E used. 

6.7.5 Uncertainty

6.7.5.1 Film

A 1962 review of commercial film badge services at ANL-E indicated that Atomic Film Badge 

Corporation was the only contractor to read neutron films. From February 27 to April 10, 1961, 

418 control films that were exposed to 300 mrem from a PuBe source were processed. Of those, 

199 films (47.6%) were not within ±25% of the expected value (Strom 1982). This same contractor’s 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 35 of 73

performance on gamma control films showed only 3.1% that were not within ±25% of the expected 

value. 

Many authors have expressed concern over latent image fading for Kodak NTA film, especially in 

high-humidity environments. Some European manufacturers sealed their NTA-type film in aluminum 

or an aluminized plastic to reduce the effects of humidity (AEC 1973). No NTA fading data was found 

for ANL-E. However, fading is not expected to be significant for a monthly exchange frequency. 

In 1980, ANL-E participated in performance testing by the University of Michigan of the NTA film 

dosimeter (Neal 1980). This data showed an average standard deviation of 10% for high doses (301

to 5,000 mrem) of 252Cf. However, for lower doses (100 to 300 mrem) the average standard deviation 

was 64%. The standard deviation of both ranges was 48%. Because this performance might not 

have been typical for earlier periods, a standard deviation of 50% was selected for this analysis. 

Again the simplified uncertainty analysis of NIOSH (2007) was used. The results are shown in 

Table 6-17. 

Table 6-17. Uncertainty for NTA film and TLD 

neutron dose. 

Dose

(mR)

Upper 95% confidence

neutron dose (mrem)

NTA film

Hosger &

other TLDs

Current 

TLD

5 55 25 15

10 61 31 21

20 74 42 34

50 120 82 77

100 210 160 150

200 400 300 300

500 990 800 760

1000 2,000 1,500 1,500

2000 4,000 3,000 3,000

6.7.5.2 Thermoluminescent Dosimeter

The LOD of the Hosger albedo TLD was 20 ±10 mrem at the 95% confidence level with an average 

spectral energy of 2.4 MeV (Strom 1982). The standard deviation was therefore 26% (50% divided by 

1.96). The simplified method was used to calculate the upper 95% confidence doses. No data were 

found for the other interim albedo dosimeters. Until such data are found, the values for the Hosger 

albedo TLD should be used. 

The current albedo system did not undergo DOELAP testing in 1988 or 1991. The lower LOD for this 

badge is 10 mrem in normal personnel dosimetry practice (Dolacek 1996). Because this dosimeter 

involves the reading of individual TLD chips, like the Hosger dosimeter, the same standard deviation 

for high doses was assumed (26%). Again, the simplified method was used to calculate the upper 

95% confidence doses. The results for both NTA film and TLDs are presented in Table 6-17. 

6.7.6 Application of the Neutron Correction Factors

The sequence of application of the neutron correction factors for NTA film is:

1. Calculate the missed dose (Section 6.7.3.1), if applicable; 

2. Apply the correction factor for unmonitored energy range (Section 6.7.3.2); and
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 36 of 73

3. Apply the correction for the neutron quality factor for each IREP energy range (Table 6-13, 

Section 6.7.1.2). 

Example

An ANL-E worker in Building 200 (plutonium chemistry facility) is monitored by NTA film for 6 months 

in 1960 and receives doses of 0, 0, 100, 300, 0, and 0 mrem. 

1. The missed dose is LOD/2 (25 mrem) × 4 = 100 mrem. The total for the period is 100 (missed 

dose) + 100 + 300 = 500 mrem. 

2. Because the neutron film did not respond to energies below about 800 keV, the correction for 

the unmonitored energy range is 500 × 4 (Table 6-16) = 2,000 mrem. 

3. The doses for each IREP energy group from Table 6-13 for a plutonium chemistry and 

metallurgy facility are:

– <0.01 MeV, 2,000 × 0 = 0 mrem, 

– 0.01 to 0.10 MeV, 2,000 × 0.2 = 400 mrem, 

– >0.10 to 2 MeV, 2,000 × 1 = 2000 mrem, 

– >2 to 20 MeV, 2,000 × 0.52 = 1040 mrem, and

– >20 MeV, 2,000 × 0 = 0 mrem. 

6.8 ELECTRON DOSE

The earliest published maximum permissible exposure limits at ANL-E recognized the potential 

hazards of beta exposures of the skin (Nickson 1946). The pioneering work there in fabricating 

uranium fuel elements and processing irradiated fuel could have resulted in significant beta 

exposures. Work since the early days has involved a wide range of activities with different natural 

and man-made isotopes. Beta exposures, including exposures to high-energy betas (more than 

1 MeV), cannot be ruled out. The dose reconstructor should be aware that early results from ring and 

wrist dosimeters might have been reported on separate forms and might need to be considered for 

skin dose to the hands and forearms. 

6.8.1 Energy Groups

Although documentation is lacking, it appears that for the two-element badge (1946 to approximately 

1960) the open window doses were nearly always equal to or greater than the shielded window 

doses, indicating that the open window reading included both penetrating and nonpenetrating 

radiation. The exceptions are probably due to the reading uncertainty. The four-element badge 

(approximately 1960 to 1989) used an algorithm to determine beta dose. The algorithm also 

determined photon dose below 150 keV (termed “X-ray”). It appears that this dose was added to the 

photon dose above 150 keV (termed “gamma”) to determine the penetrating dose. Therefore, 

penetrating and nonpenetrating (beta only) radiations are reported separately after approximately 

1960. No detail on the reporting of early film or TLD ring dosimeters was found. The current TLD 

system reports the shallow or skin dose as the sum of the penetrating and nonpenetrating doses. 

ORAUT (2005b) suggests that a correction factor of 0.6 should be applied to the nonpenetrating dose 

from early film dosimeters and that nonpenetrating doses should be assigned as greater than 15 keV 

electrons or as photons less than 30 keV if the employee worked with or around plutonium. 

All measured and missed nonpenetrating doses that are considered electrons should be assigned to 

the >15 keV IREP energy group. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 37 of 73

6.8.2 Calibration Factor and Units for Dose Conversion Factors

During the use of the two-element film badge, the term “arbitrary units” (a.u.) (or sometimes “other”) 

was used to report the open window dose through 1960. This term was an acknowledgement that the 

open window reading could not be correctly interpreted without knowing the energy of the beta or soft 

X-ray that exposed it. The May 1958 specifications for film badge service defined arbitrary unit as a 

unit used to compare the density of a film in the badge window position that was caused by exposure 

to unknown beta-gamma or X-ray radiation, or both, to the equivalent density of a film in the badge 

window position that would be caused by exposure to a known 60Co or 226Ra source in milliroentgen

(Strom 1982). The arbitrary units were later correlated with an exposure to uranium by applying a 

factor of 1.6 based on information from Landauer (ca. 1951). Although the open window in these 

dosimeters over-responded to low-energy photons, the records might underreport the doses from 

electrons (see Section 6.8.3.1). Therefore, open window readings that were reported as arbitrary 

units through 1960 should be multiplied by 1.6 to obtain beta doses in millirem, or by the factor 

applicable to a specific exposure situation as indicated in an individual dose record (typically hand 

written). Beta doses reported in units of millirem or millirep do not need to be adjusted. 

By 1961 the practice of reporting in arbitrary units ended, and the specifications for the four-element 

film badge (Hanford type) indicated that the readings in the open window were to be reported in 

density units as well as beta units (Strom 1982). These later film dosimeters at ANL-E appear to have 

been calibrated with uranium starting at least in 1963 (Pingel and Gray 1963). The wR for electrons at 

all energies is 1 (ICRP 1991). Therefore, reported beta doses are equivalent to millirem. 

The current ANL-E beta-gamma dosimeter, which is based on the Panasonic UD-814AS4, is 

accredited for the uranium slab geometry. However, the algorithm uses the ratio of the net beta 

response (after subtracting the contribution from photons) on elements 1 and 2 to determine beta 

energy. A calibration factor is estimated from the ratio (Dolacek 1996). 

6.8.3 Missed Dose and Limit of Detection

In 1956, the two-element film badge was capable of measuring doses from 50 mR to 500 R 

(Novak 1956). No separate LOD was given for beta radiation, so it was assumed to be the same. 

The specifications for the multielement film badge (Hanford type) indicated that the low end of the 

exposure range was to be 25 mR and included beta radiation (Strom 1982). The LOD of the current 

Panasonic TLD system is 30 mrem for shallow dose equivalent (Dolacek 1996). Table 6-18 lists the 

beta LODs for the ANL-E dosimeters. Specialized extremity dosimetry at ANL-E included film and at 

least two designs of TLD rings (see Section 6.10). No published LODs were found for the early 

dosimeters. The LOD for the current TLD ring is 10 mrem (Dolacek 1996). 

Table 6-18. Beta dosimeter LODs. 

Years LOD

1946–1959 50 mrem

1960–1988 25 mrem

1989–present 30 mrem whole body

10 mrem TLD ring

Under-Reporting of Shallow Dose Equivalent

Due to the thickness of the covering material and the thickness of early beta detectors, early betamonitoring systems under-reported the dose for a depth of 7 mg/cm2

. A general analysis of the 

under-response of beta dosimeters was done for INEEL (ORAUT 2011). Because the dosimeters at 

the two sites were similar and few details are available for ANL-E, the INEEL analysis was used in 

part here. Table 6-19 provides the fraction of recorded beta dose for the various dosimeters. To 

determine the corrected beta dose, the dose reconstructor should divide the nonpenetrating results by 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 38 of 73

the values in the last column of Table 6-19. This result will probably be an overestimate because the 

beta calibration involved but undoubtedly did not consider a similar correction. This value is used 

directly for the shallow dose equivalent [Hp(0.07)]. 

Table 6-19. Beta dosimeter thickness and associated under-reporting. 

Dosimeter

system Period

Cover

(mg/cm2)

Detector thickness

(mg/cm2)

Estimated fraction 

of dose reported

Two-filter film

Early film ring

1946–1959 50 50 0.49

Multifilter film 1960–1988 100 50 0.35

Old TLD ring 1970–1977 104 240 0.21

Current TLD ring 1978–present 18 89 0.56

Panasonic TLD 1989–present 16 15 0.78

6.8.4 Angular Dependence and Exposure Geometry

The sensitive dosimeter elements were mounted in a dosimetry badge. The assembled badge is 

expected to display a severe angular dependence to beta exposure, but in most cases a worker’s 

normal movements tend to average out some of this dependence (DOE 2001). 

The element responses of the Panasonic dosimeter were not specifically tested. For low-energy 

(K-17) X-rays, the element responses of the Panasonic dosimeter generally decreased as the angle 

between the incident radiation and the plane perpendicular to the TLD increased from 0°. Based on

DOE (2001), no angular correction factor is proposed. 

The current TLD ring dosimeter consists of two to four TLD-700 elements around the circumference of 

the ring. Exposures were made to X-rays of various energies to test the response of the elements to

the direction of the radiation (Dolacek 1996). Because the angle of exposure is more or less random 

with this arrangement, no correction is possible. 

Due to exposure geometry, recorded beta doses might significantly overestimate or underestimate the 

dose to the skin at the cancer diagnosis location. Exposure geometry must be analyzed on a caseby-case basis. No site-wide correction is possible with the available information. 

6.8.5 Uncertainty

The method in NIOSH (2007) was used to determine the uncertainty (upper 95% confidence dose) for 

film readings. This method is based on a statistical discussion in Film Badge Dosimetry in 

Atmospheric Nuclear Tests (NRC 1989). 

6.8.5.1 Film

ANL-E used various films as described in Section 6.6 to measure beta dose with the whole-body 

dosimeter between 1946 and 1988. A similar uncertainty estimation methodology was used to 

develop a spreadsheet that matched the illustration given in NRC (1989). No beta calibration curves 

were found for the two-element badge. Beta dose was determined for the four-element badge by 

subtracting the density behind the aluminum filter from the density behind the open window and then 

correcting for possible X-ray exposure. Review of the Kodak Type 2 calibration curve for uranium 

betas from 1981 determined film sensitivity (Strom 1982). A saturation density for 502 film was 

assumed. Using this approach, the upper 95% confidence doses for various beta doses were 

calculated. A limit of 120% was applied as discussed in Section 6.6.5.1. This limit affects the upper 

95% confidence dose at 100 mrem and above. Table 6-20 lists these upper 95% confidence doses. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 39 of 73

Table 6-20. Uncertainty for beta dosimeters. 

Dose

(mR)

Upper 95% confidence beta dose (mrem)

Kodak Type 2 film DOELAP-accredited TLD

10 25 40

20 35 50

50 66 82

100 120 140

200 240 250

500 600 600

1,000 1,200 1,200

2,000 2,400 2,400

6.8.5.2 Thermoluminescent Dosimeter

TLDs provided improved beta dosimetry. ANL-E replaced film whole-body dosimeters with TLDs in 

1988 when DOELAP performance testing began. According to the site’s TBD (Dolacek 1996), the 

current beta-gamma TLD (Panasonic UD-814SA4) responds within 20% of the mixed radiation fields 

that were tested by DOELAP. The largest positive deviations were for mixtures of X-rays and beta 

particles. The few negative deviations for mixed fields were less than 10% below the delivered dose. 

These data are for radiation normal to the badge (Dolacek 1996). 

The standard deviation of the null readings of the DOELAP-accredited Panasonic dosimeter was not 

documented. The analysis therefore used the simplified dosimetry uncertainty calculation from 

NIOSH (2007, Section 2.1.1.3.3), where the critical level Lc is the estimated LOD, as described in 

Section 6.6.3.1. The percent standard error at high air kerma was estimated to be 10%. This 

assumption is favorable to claimants for this dosimetry system. Table 6-20 lists the upper 95% 

confidence doses. 

6.8.6 Skin Contamination

Skin contamination incidents have occurred throughout the history of the site. Skin decontamination 

methods and procedures were generally well documented, but the specific forms for reporting were 

not. These could have been covered in lower-level procedures that were not available. It can be 

assumed that the practices were consistent with other DOE sites where skin contaminations were 

documented in the records. The Environment, Safety and Health Manual currently requires recording 

of the nonuniform shallow dose equivalent to the skin from contamination on the skin if the dose is 

equal to or greater than 2% of the specified limit for the skin in 10 CFR 835.702(b) (ANL 2005b). 

By 1956, beta-gamma hand and foot monitoring was done with Geiger counters and alpha hand 

monitoring was done with a proportional counter. It appears that these were permanently installed 

devices in addition to portable monitors (Novak 1956). 

Natural and depleted uranium were used and processed at ANL-E, and their progeny was potentially 

contained in the material and would result in beta exposures to the skin. Other high-energy beta 

emitters such as 90Sr/90Y were present at ANL-E. In addition, beta-emitting fission products and 

accelerator-produced isotopes were widely used at the site. 

The area of the contamination might not be available, in which case dose reconstructors should 

estimate it in accordance with Section 2.3.3 of NIOSH (2007). The contamination report will probably 

not indicate the length of time that the contamination was present on the skin. A reasonable 

assumption that is favorable to claimants is that the contamination was present for 4 hours because, 

for example, an individual could have received contamination at the beginning of the shift, not taken a 

midmorning break, and discovered the contamination upon monitoring when leaving the production 

area at lunch. Once the contamination was discovered, initial decontamination was performed locally, 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 40 of 73

which normally resulted in removal of most of the contamination. If these initial efforts failed to 

remove enough of the contamination, the individual was taken to the occupational medicine group 

where further decontamination efforts were made under medical supervision (Novak 1956; 

ANL 1973–1984). 

Dose reconstructors could use software such as VARSKIN (recommended in NIOSH 2007) or other 

appropriate means for this calculation. 

6.9 EXPOSURES OUTSIDE RADIOLOGICALLY CONTROLLED AREAS

For individuals who worked outside the radiologically controlled areas, ambient on-site external 

exposure is an appropriate estimate of exposure, and dose should be assessed in accordance with 

ORAUT-TKBS-0036-4 (ORAUT 2006) rather than this document. Available information should be 

evaluated (such as job titles, work locations, presence or absence of dose records, etc.) to determine 

an individual’s likelihood of having worked in a radiologically controlled area. For workers who would 

not have been required to enter radiologically controlled areas based on such information (for 

example, clerks, typists, janitors, cafeteria workers, etc.), ambient external dose should be applied. If 

records are inadequate to determine whether an individual’s occupation would have involved work in 

a radiologically controlled area, or if monitoring records exist for an individual, they should be 

assumed to have been radiological workers, and dose should be assigned in accordance with the 

guidance of this document. 

6.10 EXTREMITY DOSIMETRY

Monitoring exposure to the skin was important very early in ANL-E’s history and, while maximum 

permissible levels were established (Nickson 1946) it is not clear what monitoring methods were in 

use during what periods. In the early years of the site, body badges were apparently fitted with 

alternative fastening devices and worn on the wrist, leg, or head. 

Monthly radiation safety activity reports from the D-211 Cyclotron Facility indicated that commercial 

finger ring film service was initiated for a 13-week trial basis in 1954 (Okolowitz and Jezik 1954). A 

nonmagnetic film badge holder from Tracerlab was issued to all cyclotron workers in November 1954 

after testing in the magnetic field. Wrist and head monitors were worn for an evolution involving a 235U 

target. The finger ring film service from R. C. Scientific Instrument Company expired in January 1955. 

It is not clear if this is the same ring referred to above (Okolowitz and Jezik 1955). 

The 1956 Radiation Safety Guide indicates that wrist film badges were available (Novak 1956). The 

badge was likely similar to the whole-body film badge (two-element design with an open window and 

a 1-mm cadmium filter). Wrist badges with neutron film were available and were required at CP-5 

(Novak 1956). 

TLD ring badges of an undetermined type were investigated for use in the Building 350 plutonium 

“hoodline” during all of 1967. The rings were calibrated with plutonium fuel using an extrapolation 

chamber. Depending on the type of work, either one or two rings (right and left) were worn. Workers 

continued to wear wrist dosimeters, and the results were to be compared (Steele and Allen 1967). 

The results and conclusions of the study have not been found. 

The current extremity monitoring system consists of ring dosimeters. The dosimeters are issued to 

personnel who have the potential for significant hand exposure including those who work with X-ray 

diffraction equipment. However, this equipment is easily shielded and such exposures are assumed 

to have been rare. The TLD is a Harshaw TLD-700, and each chip measures 0.125 by 0.125 by 

0.035 in. Each ring has four cavities spaced at 90° intervals around its circumference into which TLD 

chips can be placed. Normally two or four chips are used (Dolacek 1996). Curves of net counts 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 41 of 73

versus exposure show that the difference in response of the four elements is greatest at the lower 

energies, as could be expected (Dolacek 1996). Dolacek (1996) is not clear about how the dose of 

record is determined. However, a records review indicates that, if one chip reading is over 10 mrem, 

a non-zero dose is recorded in the record. 

6.11 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 42 of 73

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ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England. [SRDB Ref ID: 11517]

Ing, H., and S. Makra, 1978, Compendium of Neutron Spectra in Criticality Accident Dosimetry, 

Technical Report Series No. 180, International Atomic Energy Agency, Vienna, Austria. 

Landauer (R. S. Landauer, Jr. and Company), ca. 1951, “Conversion Factors for “Other” Readings of 

Film Badge Reports,” Park Forest, Illinois. [SRDB Ref ID: 95910]

Lennox, D. H., and C. N. Kelber, 1956, Summary Report on the Hazards of the Argonaut Reactor, 

ANL-5647, University of Chicago, Argonne National Laboratory, Lemont, Illinois, December. 

[SRDB Ref ID: 17204]

Luck, W. D., 2002, “Information Concerning NIOSH Requests,” letter to G. Lane (U.S. Department of 

Energy, Chicago Operations Office), University of Chicago, Argonne National Laboratory, 

Argonne, Illinois, May 3. [SRDB Ref ID: 13106]

Mendelstein, L., 1951, “Health Physics Operations Committee Report - March 26, 1951 to May 27, 

1951,” memorandum to J. E. Rose, University of Chicago, Argonne National Laboratory, 

Argonne, Illinois, June 8. [SRDB Ref ID: 16726]

Moe, H. J., J. H. Vacca, V. R. Veluri, and M. White, ca. 1994, “Radiation Measurements at the 

Advanced Photon Source (APS) Linear Accelerator.” [SRDB Ref ID: 17201]

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

Neutron Radiation, Report 38, Bethesda, Maryland. 

NCRP (National Council on Radiation Protection and Measurements), 2003, Radiation Protection for 

Particle Accelerator Facilities, Report 144, Bethesda, Maryland. 

Neal, P. E., 1980, “Personnel Dosimetry Records Survey,” letter to Chicago Operations and Regional 

Office (U. S. Department of Energy), University of Chicago, Argonne National Laboratory, 

Argonne, Illinois, May 19. [SRDB Ref ID: 16768]

Nickson, J. J., 1946, “Maximum Permissible Levels Employed at the Argonne National Laboratory,” 

memorandum to Colonel W. D. Fleming, University of Chicago, Argonne National Laboratory, 

Chicago, Illinois, October 23. [SRDB Ref ID: 8255]

NIOSH (National Institute for Occupational Safety and Health), 2007, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 3, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 21. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 45 of 73

NIOSH (National Institute for Occupational Safety and Health), 2010, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October 5. 

Novak, J. R. (editor), 1956, Radiation Safety Guide, University of Chicago, Argonne National 

Laboratory, Lemont, Illinois, June. [SRDB Ref ID: 16973]

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

National Academy of Sciences, National Academy Press, Washington, D.C. 

Okolowitz, K. G., and R. M. Jezik, 1954, “Monthly Summary of D-211, Radiation Safety Activities in 

Connection with the Cyclotron Program for the Period of October 3 - 30, 1954,” memorandum 

to J. R. Novak, University of Chicago, Argonne National Laboratory, Argonne, Illinois, 

November 3. [SRDB Ref ID: 16704, p. 11]

Okolowitz, K. G. and R. M. Jezik, 1955, “Monthly Summary of D-211, Radiation Safety Activities, in 

Connection with the Cyclotron Program for the Period of November 28, 1954, to January 1, 

1955,” memorandum to J. R. Novak, University of Chicago, Argonne National Laboratory, 

Argonne, Illinois, January 10. [SRDB Ref ID: 16704, p. 13]

ORAUT (Oak Ridge Associated Universities Team), 2005a, “Information Needed for External 

Dosimetry TBD Section for ANL-E,” Oak Ridge, Tennessee, September 21. [SRDB Ref ID: 

19636]

ORAUT (Oak Ridge Associated Universities Team), 2005b, Interpretation of Dosimetry Data for 

Assignment of Shallow Dose, ORAUT-OTIB-0017, Rev. 01, Oak Ridge, Tennessee, 

October 11. 

ORAUT (Oak Ridge Associated Universities Team), 2005c, Savannah River Site ORAUT-TKBS-0003, 

Rev. 03, Oak Ridge, Tennessee April 5. 

ORAUT (Oak Ridge Associated Universities Team), 2006, Argonne National Laboratory-East – 

Occupational Environmental Dose, ORAUT-TKBS-0036-4, Rev. 00, Oak Ridge, Tennessee, 

February 9. 

ORAUT (Oak Ridge Associated Universities Team), 2010, Rocky Flats Plant – Occupational External 

Dose, ORAUT-TKBS-0011-6, Rev. 02 PC-1, Oak Ridge, Tennessee, October 20. 

ORAUT (Oak Ridge Associated Universities), 2011, Idaho National Laboratory and Argonne National 

Laboratory-West – Occupational External Dosimetry, ORAUT-TKBS-0007-6, Rev. 03, 

Oak Ridge, Tennessee, April 19. 

ORAUT (Oak Ridge Associated Universities Team), 2013, Los Alamos National Laboratory – 

Occupational External Dose, ORAUT-TKBS-0010-6, Rev. 03, Oak Ridge, Tennessee, 

March 21. 

Pingel, J. H., and P. C. Gray, 1963, “Dose Rates from Uranium for Film Calibration,” memorandum to 

P. Tedeschi, University of Chicago, Argonne National Laboratory, Argonne, Illinois, 

December 31. [SRDB Ref ID: 16804]
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 46 of 73

Steele, T. A., 1965, Use of Personal Monitoring Neutron Film Near the 50 MeV Injector of the ZGS, 

University of Chicago, Argonne National Laboratory, Argonne, Illinois, October 12. [SRDB Ref 

ID: 17280]

Steele, T. A., and T. M. Allen, 1967 “Building 350 Extremity Exposures as Determined by 

Thermoluminescent Finger Dosimeters,” memorandum to W. E. Bleiler, University of Chicago, 

Argonne National Laboratory, Argonne, Illinois, January 25. [SRDB Ref ID: 17216]

Strom, D. J., 1982, “Argonne Trip Report and Dosimetry Program Evaluation Data,” memorandum to 

D. Robie, University of North Carolina at Chapel Hill, Chapel Hill, North Carolina, 

November 24. [SRDB Ref ID: 12632, p.3] 

Tedeschi, P., 1958, “Progress Made on Calibration of Eastman Kodak Neutron Monitoring Film 

Type A for 14 MeV Neutrons,” memorandum to J. R. Novak, University of Chicago, Argonne 

National Laboratory, Argonne, Illinois, September 22. [SRDB Ref ID: 17198]

Till, H. A., 1966, “Evaluation of Bendix Tissue Equivalent Dosimeters,” memorandum to J. R. Novak, 

University of Chicago, Argonne National Laboratory, Argonne, Illinois, August 26. [SRDB Ref 

ID: 17188]

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

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

Physics, volume 67, number 4, pp. 402–405. [SRDB Ref ID: 8601]

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, Battelle Memorial 

Institute, Pacific Northwest Laboratory, Richland, Washington, September. [SRDB Ref ID: 

4793]

Yule, T. J., 1971, “II-15. Spectrum Measurement Between the Halves of ZPR-6 Assembly 7,” Applied 

Physics Division Annual Report, July1, 1970, to June 30, 1971, ANL-7910, University of 

Chicago, Argonne National Laboratory, Argonne, Illinois. [SRDB Ref ID: 18242]

Yule, T. J., 1973, “Report on the Fourth Workshop on Personnel Neutron Dosimetry Held at Miami 

Beach, June 13-15, 1973,” memorandum to distribution, University of Chicago, Argonne 

National Laboratory, Argonne, Illinois, June 19. [SRDB Ref ID: 17274]

Yule, T. J., R. J. Armani, and R. Gold, 1972, “III-14. The Adequacy of Solid-State-Track-Recorder 

Fast-Neutron Personnel Dosimeters for Use in ZPR Environments,” Applied Physics Division 

Annual Report, July 1, 1971, to June 30, 1972, ANL-8010, University of Chicago, Argonne 

National Laboratory, Argonne, Illinois. [SRDB Ref ID: 18243]
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 47 of 73

GLOSSARY

albedo dosimeter

Thermoluminescent dosimeter that measures the thermal, intermediate, and fast neutrons 

scattered and moderated by the body or a phantom from an incident fast neutron flux. 

arbitrary unit (a.u.)

Unit for comparing the density of a film in the badge window position from exposure to 

unknown beta-gamma radiation or x-rays, or both, to the equivalent density of a film in the 

badge window position from exposure to a known 60Co or 226Ra source in milliroentgen units. 

beta dose

Designation (i.e., beta) on some records for external dose from beta and less-energetic X-ray 

and gamma radiation, often for shallow dose or dose to the lens of the eye.

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 absorbed dose (Dd) 

Absorbed dose in units of rem or sievert at a depth of 1 centimeter (1,000 milligrams per 

square centimeter). 

deep dose equivalent [Hd, Hp(10)] 

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

per square centimeter). See 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. 

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

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

units of roentgens, rad, rep, or grays.

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

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See albedo dosimeter, film dosimeter, neutron film dosimeter, 

pocket ionization chamber, and thermoluminescent dosimeter. 
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dosimetry

Measurement and calculation of internal and external radiation doses.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

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.

extremities

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. 

field calibration

Dosimeter calibration based on radiation types, intensities, and energies in the work 

environment. 

film

(1) In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. (2) X-ray film.

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.

fission

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons.

gamma ray, particle, or photon (γ)

See gamma radiation. 

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. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 49 of 73

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, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.

minimum reporting level

Level below which an analytical dose is not recorded in the worker’s dose record, usually 

based on a site-specific policy decision. The recording level is not necessarily the same as 

the minimum detectable amount or activity for that measurement. Also called less-than value, 

minimum reportable dose, minimum recordable or recording dose, recording level, and 

reporting level.

neutron (n)

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron film dosimeter

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

nuclear track emulsion, type A (NTA)

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

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

power magnification. The number of tracks in a given area is a measure of the dose from that 

radiation. 

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide. 

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 dosimeter

See pocket ionization chamber. 

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 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 50 of 73

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. 

personal dose equivalent, slab phantom [Hp,slab(d)] 

Personal dose equivalent in relation to a slab phantom. See phantom and slab phantom. 

phantom

Any structure that contains one or more tissue substitutes (any material that simulates a body 

of tissue in its interaction with ionizing radiation) and is used to simulate radiation interactions 

in the human body. Phantoms are primarily used in the calibration of in vivo counters and 

dosimeters. See slab phantom. 

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

photon X-ray

Electromagnetic radiation of energies between 10 keV and 100 keV whose source can be an 

X-ray machine or radioisotope. 

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 

to be able to read the dosimeter. Also called pencil, pocket pencil, pencil dosimeter, and 

pocket dosimeter. 

quality factor, Q

Principal modifying factor (which depends on the collision stopping power for charged 

particles) that is employed to derive dose equivalent from absorbed dose. The quality factor 

multiplied by the absorbed dose yields the dose equivalent.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation. 

radioactive

Of, caused by, or exhibiting radioactivity.

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. See radionuclide. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

rem

A unit of dose equivalent equal to the product of the number of rad absorbed and the quality 

factor. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 51 of 73

roentgen (R or r)

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.

rover dosimeter or rover

Pencil dosimeter that was not permanently assigned. Typically worn by visitors or personnel 

who were not normally assigned to an area. 

shallow dose equivalent [Hs, 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. 

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

slab phantom

Phantom in the form of a slab of plastic typically about 30 or 40 centimeters square by 

15 centimeters deep. See phantom. 

skin dose

See shallow dose equivalent. 

thermoluminescence

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

thermoluminescent dosimeter (TLD)

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

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

a measurement of absorbed dose.

U.S. Atomic Energy Commission (AEC)

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

U.S. Department of Energy. 

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

See X-ray radiation.

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. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 52 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

LIST OF FIGURES

FIGURES TITLE PAGE

A-1 External Radiation Exposure Record ..................................................................................... 53

A-2 Radiation Exposure Record for Doses Assigned in CY 1989 or Later .................................... 55

A-3 Individual Meter Record ......................................................................................................... 56

A-4 Cumulative Exposure Record................................................................................................. 57

A-5 Accumulated Exposure Report............................................................................................... 58

A-6 Semi-Annual Accumulative External Exposure Report ........................................................... 59

A-7 Whole Body Radiation Exposure Summary............................................................................ 60

A-8 Master Radiation Exposure File ............................................................................................. 61

A-9 Master External Radiation Exposure Report........................................................................... 62

A-10 Neutron Exposure Report....................................................................................................... 63

A-11 Special Exposure Records ..................................................................................................... 64

A-12 Special Film Request .............................................................................................................65

A-13 Special Meter Assignment and Radiation Exposure Report ................................................... 66

A-14 Special TLD Finger Exposure Report..................................................................................... 67

A-15 Personnel Dose Equivalent Evaluation – Level 1 ................................................................... 68

A-16 Termination Occupational Exposure Report........................................................................... 69
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 53 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-1. External Radiation Exposure Record (page 1 of 2). 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 54 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-1. External Radiation Exposure Record (page 2 of 2). 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 55 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-2. Radiation Exposure Record for Doses Assigned in CY 1989 or Later.
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 56 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-3. Individual Meter Record. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 57 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-4. Cumulative Exposure Record. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 58 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-5. Accumulated Exposure Report. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 59 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-6. Semi-Annual Accumulative External Exposure Report. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 60 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-7. Whole Body Radiation Exposure Summary. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 61 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-8. Master Radiation Exposure File.
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 62 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-9. Master External Radiation Exposure Report. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 63 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-10. Neutron Exposure Report. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 64 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-11. Special Exposure Records. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 65 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-12. Special Film Request. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 66 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-13. Special Meter Assignment and Radiation Exposure Report. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 67 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-14. Special TLD Finger Exposure Report.
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 68 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-15. Personnel Dose Equivalent Evaluation – Level 1. 
Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 69 of 73

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-16. Termination Occupational Exposure Report. 
ATTACHMENT B

FACILITIES BY BUILDING NUMBER AND SECTION CODE

Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 70 of 73

Table B-1. Facilities by building number and section code.

a,b

Building

Section codes

Facility name

Years of 

operation

(approximate) Additional information

Facility type

1980 Present

Site A CP-2 1943–1954 CP-1 moved from West Stands and reassembled 

at Site A. Graphite moderated.

Reactor

Site A CP-3, CP-3ʹ 1944–1954 CP-3 was heavy-water moderated, natural 

uranium fuel. Natural U fuel replaced with 

enriched and renamed CP-3ʹ 1950.

Reactor

25 AA Argonaut 1957–1980 University research reactor. Light-water

moderated and cooled, graphite moderated. 

Enriched U-235 fuel (20%).

Reactor

40 Chemistry Hot Lab 1948– Radioactive material stored there awaiting 

disposition (1994)

Radioactive 

Materials

200 JA, JG, JM, 

JR

JG, JM, VA Chemistry Building 1951– Characterization of isotopes and nuclides. High, 

moderate, and low-level cells. Only one wing 

(MWing Addition) still used for studies of 

radioactive materials; designed to work with Pu 

safely.

Radioactive 

Materials

202 GE, GF, GJ GE, GN, 

VE, VK

Biology Building

Janus Reactor 1964–1992

1952– Low and high gamma irradiation rooms. Neutron 

Irradiation Facility (1959). Radiation effects 

studies 1950s to 1970s.

Reactor 

(1964-1992)

Radioactive 

Materials

203 ED, EF, 

EG, EL

ED, EF, EG, 

VH

Physics Building

Initially housed Dynamitron 

and Van de Graaff accelerators

1952– Many modifications to add new accelerator 

facilities, such as the Tandem Van de Graaff. 

ATLAS added to the north side of building in 

1982. First ion beam accelerated 1985. 1 Ci 

PuBe source stored there (1994)

Accelerator

205 YI VC Chemical Technology 1951– Nuclear fuel processing and fuel recovery 

studies. Proof-of-breeding experiments. G and 

K Wing Laboratories are current nuclear facilities.

Radioactive 

Materials

206 EA EA, VF Engineering Development Lab 1953– Sodium studies. Half-scale mock up of EBR-II 

that was built at ANL-W. Temporary hot cell 

facility added 1955.

Radioactive 

Materials

208 EY EY Reactor Engineering Building 1952– Mostly office space. A Van de Graaff (F Wing) 

added 1956.

Accelerator
ATTACHMENT B

FACILITIES BY BUILDING NUMBER AND SECTION CODE (continued)

Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 71 of 73

Building

Section codes

Facility name

Years of 

operation

(approximate) Additional information

Facility type

1980 Present

211 JH, JK JH, JK, JR, 

VB

Cyclotron Building or Low 

Energy Accelerator Facility 

1951– 60-Inch Cyclotron. Caves to process irradiated 

materials. Major expansion 1962 adding a 3-

MeV Van de Graaff and a 20-MeV high-current 

electron LINAC. 1992 characterization indicated 

Co-60 was most prevalent isotope.

Accelerator

212 GA, GB, 

GC, GD

GA, GB, 

GC, GD, VD

Fuel Technology Center

AGHCF

1962– HVEM formally dedicated 1981. AGHCF 

constructed in 1963 to contain Pu in N2. HVEM 

removed 2001. AGHCF and the AMFF are 

current nuclear facilities.

Radioactive 

Materials

223 GK, GL GL Material Science Division 1968– Mainly used for the development of 

superconductors. 1,200 Ci Co-60 stored there in 

1994.

Radioactive 

Materials

301 Physics and Metallurgy Hot 

Lab

1951– Fuel analysis and processing activities. Five 

cave facilities. Principal hot lab for reactor 

development, reactor fuel studies, and handling 

of highly radioactive materials. In 1998, parts of 

building were not controlled areas, used as office 

space. The dominant floor contaminant was Cs137.

Radioactive 

Materials

303 MWSF ~1950– Current nuclear facility. No history located on 

previous uses.

Radioactive 

Materials

306 MG, MJ MH, MJ, VM Waste Management 

Operations

(Decontamination Shop)

1950– Size nearly doubled in 1961. Two caves and six 

retention tanks. Size reduction and packaging 

for shipment. Laundry facilities. Current nuclear 

facility.

Radioactive 

Materials

308, 309 MA MA Energy Technology 1956– High bay for EBR-II project (partial mock-up). 

Sodium loop tests for EBR-II and the Integral 

Fast Reactor.

Radioactive 

Materials

310 MK, MT Experimental Waste 

Processing, Storage, and 

Shipping

1948– Similar function to Build 306. Dry waste 

incineration. Retention tanks in basement. 

Gamma Irradiation Facility added in 1954. Spent 

fuel rods used in food irradiation experiments. 

Several additions in 1970s to support EBR-II fuel 

cycle facility at ANL-W

Radioactive 

Materials

314 ST ST ZPR Assembly 1968– Fast Neutron Generator. Considered part of the 

ZPR complex.

Radioactive 

Materials
ATTACHMENT B

FACILITIES BY BUILDING NUMBER AND SECTION CODE (continued)

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Building

Section codes

Facility name

Years of 

operation

(approximate) Additional information

Facility type

1980 Present

315, 316 SN, SP SP, VP ZPR Assembly 1950 (316)–

1962 (315)–1982

Additional cell added 1956. In all there were 

seven ZPRs and the ATSR in 316. 315 

Constructed in 1962 to house two ZPRs. Part of 

a larger complex of buildings (314/315/316) used 

for storage of materials waiting disposal in 1994. 

Current nuclear facility.

1950-1982 

Reactor

1983 to 

present

Radioactive 

Materials

317 MD Above- and below-grade waste 

storage

Unknown–1994 The facilities in 317 Area included the southwest, 

south-middle, and southeast in-ground concrete 

vaults, the map tube facility, the deep vault, the 

north vault, a concrete storage pad, the waste 

baler building, and a graveled container storage 

area. 317B current nuclear facility (below-grade 

storage facility).

Radioactive 

Materials

329, 330 QM CP-5 1954–1979 Wing additions between 1958 and 1962. Reactor 

room, control room, laboratories, and fuel 

storage. Nuclear physics, materials research, 

and biological studies. Decontamination and 

decommissioning of reactor completed 2000.

Reactor

331 QW QW EBWR 1956–1967 Reactor removed from shell in 1995 during 

decontamination and decommissioning. Current 

nuclear facility, now used for storage of low-level 

and transuranic waste.

1956-1967 

Reactor

Radioactive 

Materials

335 Juggernaut Reactor 1962–unknown Light-water moderated and cooled, graphite 

moderated. Highly enriched 235U fuel (93%).

Reactor

340 MP Experimental Animal Quarters 1950– Housed animals irradiated in Building 202. Bone 

cancer research 1963 to 1980.

Radioactive 

Materials

350 SH SR, VN Fuel Fabrication Facility

(now New Brunswick 

Laboratory, a DOE facility)

1952–1979 1959 became the first large-scale plant for 

making fuel from Pu. Decommissioned in 

1978-1979.

Radioactive 

materials

360-390 SA, SB, SC, 

SD, SE, SF, 

SG, SK, 

SM, SQ

ZGS 1963–1979 16 Buildings in the 360 area. Some of same 

facilities now used for the IPNS (see below). 2 Ci 

Pu–Be-239 stored in 360 in 1994

High-energy 

accelerator
ATTACHMENT B

FACILITIES BY BUILDING NUMBER AND SECTION CODE (continued)

Document No. ORAUT-TKBS-0036-6 Revision No. 01 Effective Date: 10/16/2014 Page 73 of 73

Building

Section codes

Facility name

Years of 

operation

(approximate) Additional information

Facility type

1980 Present

366 SJ AWA 2000– Research facility for new accelerator design. The 

initial phase designed to provide short 100-nC 

bunches of electrons at 20 MeV. The facility is 

housed in a concrete bunker having 6- to 8-foot 

thickness. The designed safety operating 

envelope for the facility is 400 nC, 30 pulses per 

second, at 20 MeV.

Accelerator

361, 391, 

375

SA, SB, SC, 

SD, SE, SF, 

SG, SK, SQ, 

SS, SU, SV, 

SX, SZ, WC

IPNS 1981– Uses some of the same equipment and facilities 

as the ZGS. A number of 239PuBe sources 

stored in 375 in 1994

High-energy 

accelerator

400, 410, 

411, 412, 

415, 420

RA, RB, RD, 

RE, RF, RG, 

RH, RJ, RK, 

RL, RN, RP, 

RZ, RV

APS 1994– Accelerator

a. See the Acronyms and Abbreviations for facility names. 

b. Blank = none.