ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 69

Document Title:

Argonne National Laboratory – East – 

External Dosimetry

Document Number: ORAUT-TKBS-0036-6 

Revision: 00

Effective Date: 02/09/2006

Type of Document: TBD

Supersedes: None 

Subject Expert: Eugene W. Potter

Document Owner

Approval: Signature on File Approval Date: 02/02/2006

Norman D. Rohrig, TBD Team Leader

Approval: Signature on File Approval Date: 02/01/2006

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 02/06/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 02/09/2006

Stuart. L. Hinnefeld, Health Science Administrator

 New Total Rewrite Revision Page Change

FOR DOCUM ENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REV ISION.
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 2 of 69

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.
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 3 of 69

TABLE OF CONTENTS

SECTION TITLE

6.1 Introduction..............................................................................................................................

PAGE

8

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

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.....................................................................................13

6.3.2 Observed Data Discrepancies ..................................................................................18

6.4 Historical Administrative Practices ..........................................................................................18

6.4.1 Badged Population...................................................................................................18

6.4.2 Badge Exchange Frequency.....................................................................................19

6.4.3 Field-Specific Calibration Factors..............................................................................19

6.4.4 Minimum Reported Dose..........................................................................................19

6.5 Common Issues .....................................................................................................................20

6.5.1 Number of Zero Readings.........................................................................................20

6.5.2 Discrepancies...........................................................................................................21

6.5.3 Missing Entry ...........................................................................................................21

6.5.4 Exposure Geometry..................................................................................................21

6.6 Photon Dose ..........................................................................................................................21

6.6.1 Energy Groups.........................................................................................................21

6.6.2 Calibration Factor.....................................................................................................22

6.6.2.1 Units for Dose Conversion Factors for Dose to Organ Dose.........................22

6.6.3 Missed Dose ............................................................................................................23

6.6.3.1 Limit of Detection........................................................................................23

6.6.3.2 Determination of Missed Dose.....................................................................23

6.6.3.3 Unmonitored Energy Range........................................................................23

6.6.4 Geometry - Angular Dependence..............................................................................24

6.6.5 Uncertainty...............................................................................................................24

6.6.5.1 Film............................................................................................................24

6.6.5.2 Thermoluminescent Dosimeters..................................................................25

6.7 Neutron Dose.........................................................................................................................26

6.7.1 Energy Groups.........................................................................................................27

6.7.1.1 Neutron Energy Spectra..............................................................................27

6.7.1.2 Reported Dose to Energy Groups................................................................29

6.7.2 Calibration Factor.....................................................................................................30

6.7.2.1 Units for Dose Conversion Factors for Dose to Organ Dose.........................30

6.7.3 Missed Dose ............................................................................................................31

6.7.3.1 Limit of Detection........................................................................................31

6.7.3.2 Unmonitored Energy Range........................................................................32

6.7.3.3 Neutron Dose Reconstruction Project..........................................................33

6.7.4 Geometry-Angular Dependence................................................................................33

6.7.5 Uncertainty...............................................................................................................33
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6.7.5.1 Film............................................................................................................33

6.7.5.2 Thermoluminescent Dosimeter....................................................................33

6.7.6 Application of the Neutron Correction Factors ...........................................................34

6.7.6.1 The sequence of application of the neutron correction 

 factors for NTA film is.....................................................................................34

6.7.6.2 Example .....................................................................................................34

6.8 Electron Dose.........................................................................................................................35

6.8.1 Energy Groups.........................................................................................................35

6.8.2 Calibration Factor - Units for Dose Conversion Factors for Dose to Organ Dose........35

6.8.3 Missed Dose - Limit of Detection...............................................................................36

6.8.3.1 Under Reporting of Shallow Dose Equivalent...............................................36

6.8.4 Geometry.................................................................................................................37

6.8.4.1 Angular Dependence ..................................................................................37

6.8.4.2 Exposure Geometry....................................................................................37

6.8.5 Uncertainty...............................................................................................................37

6.8.5.1 Film............................................................................................................37

6.8.5.2 Thermoluminescent Dosimeter....................................................................37

6.8.6 Skin Contamination ..................................................................................................38

6.9 Unmonitored Individuals.........................................................................................................39

6.9.1 In Radiologically Controlled Areas.............................................................................39

6.9.2 Outside of Radiologically Controlled Areas................................................................39

6.10 Extremity Dosimetry ...............................................................................................................39

References.....................................................................................................................................41

Glossary .........................................................................................................................................46

ATTACHMENT A, EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS ............50

ATTACHMENT B, FACILITIES BY BUILDING NUMBER AND SECTION CODE................67
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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..............................................................................19

6-4 Historical dose limits............................................................................................................20

6-5 Dose limits based on exchange frequency............................................................................21

6-6 Selection of photon energies for IREP..................................................................................22

6-7 Photon dose units for use with organ dose conversion factors ..............................................23

6-8 Photon dosimeter LODs.......................................................................................................23

6-9 Uncertainty for photon dose................................................................................................. 25

6-10 Neutron dose assessment methods .....................................................................................26

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

facilities...............................................................................................................................28

6-12 ICRP Publication 60 dose multipliers....................................................................................30

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

neutron energy intervals ......................................................................................................30

6-14 Neutron dose units for use with organ dose conversion factors.............................................31

6-15 Neutron dosimeter LODs .....................................................................................................32

6-16 NTA film unmonitored dose correction factors ......................................................................32

6-17 Uncertainty for NTA film and TLD neutron dose...................................................................34

6-18 Beta dosimeter LODs...........................................................................................................36

6-19 Beta dosimeter thickness and associated underreporting......................................................36

6-20 Uncertainty for beta dosimeters............................................................................................38

B-1 Facilities by building number and section code.....................................................................67

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.............................................................................28

A-1 External Radiation Exposure Record....................................................................................50

A-2 Radiation Exposure Record for Doses Assigned in CY 1989 or Later....................................52

A-3 Individual Meter Record.......................................................................................................53

A-4 Cumulative Exposure Record...............................................................................................54

A-5 Accumulated Exposure Report.............................................................................................55

A-6 Semi-Annual Accumulative External Exposure Report..........................................................56

A-7 Whole Body Radiation Exposure Summary..........................................................................57

A-8 Master Radiation Exposure File ...........................................................................................58

A-9 Master External Radiation Exposure Report.........................................................................59

A-10 Neutron Exposure Report ....................................................................................................60

A-11 Special Exposure Records...................................................................................................61

A-12 Special Film Request...........................................................................................................62

A-13 Special Meter Assignment and Radiation Exposure Report...................................................63

A-14 Special TLD Finger Exposure Report...................................................................................64

A-15 Personnel Dose Equivalent Evaluation – Level 1..................................................................65

A-16 Termination Occupational Exposure Report .........................................................................66
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ACRONYMS AND ABBREVIATIONS

a.u. arbitrary units

AEC U.S. Atomic Energy Commission

AGHCF Alpha Gamma Hot Cell Facility

AMFF Advanced Materials Fabrication 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

C.F.R. Code of Federal Regulations

Ci curie

cm centimeter

CP-1 Chicago Pile 1

CP-2 Chicago Pile 2

CP-3 Chicago Pile 3

CP-3 ́ Chicago Pile 3 Prime

CP-5 Chicago Pile 5

d day

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

EBR-II Experimental Breeder Reactor No. 2

EBWR Experimental Boiling Water Reactor

EEOICPA Energy Employees Occupational Illness Compensation Program Act

ERDA Energy Research and Development Agency

ES&H Environment, Safety, and Health

eV electron volt

g gram

GeV giga electron volt, 1 billion electron volts

hr hour

HVEM High-Voltage Electron Microscope

ICRP International Commission on Radiological Protection

in. inch

IPNS Intense Pulsed Neutron Source

IREP Interactive RadioEpidemiological Program

keV kilo electron volt, 1,000 electron volts

LAMPF Los Alamos Meson Physics Facility 

LANL Los Alamos National Laboratory

LINAC linear accelerator

LOD limit of detection

mA milliampere
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MeV mega electron volt, 1 million electron volts

mg milligram

mm millimeter

mo month

mR milliroentgen

mrad millirad

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

PIC pocket ionization chamber

PNNL Pacific Northwest National Laboratory

pps pulses per second

QF quality factor

qtr quarter

R roentgen

s second 

SSTR solid-state track recorder

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

WBC whole-body counter

wk week

wR weighting factor

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-IV ́ 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

§ section 
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6.1 INTRODUCTION

Technical basis documents (TBDs) and site profile documents are general working documents that 

provide guidance concerning the preparation of dose reconstructions at particular sites or categories 

of sites. They will be revised in the event additional relevant information is obtained about the 

affected site(s). These documents may be used to assist the National Institute for Occupational 

Safety and Health (NIOSH) 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 [probability of causation] guidelines established under subsection 

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

dose reconstruction regulation) define “performance of duty” for DOE employees with a covered 

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

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

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

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

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

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

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

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the Naval 

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

6.1.1

Argonne National Laboratory was established on July 1, 1946 and this TBD is intended to cover since 

that date. The work was a continuation of that done by the Metallurgical Laboratory of the University 

of Chicago beginning in 1941 which is an Atomic Weapons Employer under EEOICPA. The job 

locations did not change until land and buildings were acquired for the laboratory. 

Purpose
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The purpose of this TBD is to describe the external dosimetry systems and practices at Argonne 

National Laboratory-East (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

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 the areas of 

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. It contains supporting documentation to 

assist in the evaluation of occupational external doses from these processes in accordance with the 

External Dose Reconstruction Implementation Guideline (NIOSH 2002).

Scope 

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 the EEOICPA legislation. 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 used to prepare 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 used initially 

in these badges was 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 mid1950s some film badge rings (provided by a contactor) 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 

(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 contactor 

reported to process and read Kodak NTA film. Due to the labor needed 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 gamma dose measured for the same period (Dolecek 1981).
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Table 6-1. Summary of beta-gamma film-processing history . 

Film processor Dates

ANL-E 1942–1954

Tracerlab Mar 1954–Jun 1958

Landauer Jul 1958–Jun 1960

Atomic Film Badge Corporation Jul 1960–Dec 1964

Tracerlab Jan 1965–Sep 1965

ANL-E Oct 1965–1988

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

from its contactor, 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 the one 

used at the National Reactor Testing Station (NRTS) in Idaho and at Hanford in Washington State

and 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 referring 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 two albedo neutron designs (Hankins and 

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

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

of low-energy photon radiation (ANL 1996). This dosimeter passed Department of Energy 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 containing Harshaw TLD-600 and TLD-
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 11 of 69

700 chips in a polybox is inserted into the ANL-E holder for workers with the potential for neutron 

exposure (ANL 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 as the official dose, the readings were 

recorded in the individual dosimetry records up to 1964. Film badges and PICs were stored at the 

entrances to the different facilities. 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 pocket chamber) (McDowell-Boyer 2005). 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 years given are approximate. In 

some cases, changes probably were 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 the U.S. Department of Labor requests a worker's dosimetry records, the ANL-E Chemical 

Engineering Division searches for the relevant records. For external dosimetry, there are three 

records sources that might contain the needed information. The Worker Protection System is a 

database of employees between 1945 and the present. A second database (dBase III plus) contains 

the external dosimetry records for persons employed between 1989 and the present. Last, there are 

hard-copy records of the external doses for persons who were employed between 1942 and 1988. 

From these sources, an 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 

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

summary is provided (Figure A-2) for those years. Requests for the detailed information have 

resulted in ANL-E providing 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.

The pre-1989 record summaries 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 of density 1 g/cm3

. 

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

tissue of density 1 g/cm3

. 

• Code M for years in which the dose was less than the minimum sensitivity of the measuring 

device.

• Code - [hyphen] or NM for unmonitored periods; some records indicate unmonitored periods 

with a statement such as “Not monitored 1960-1962”. 
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• Code * (asterisk) after the dose when the only monitoring was by a direct reading dosimeter; 

additional codes used are explained on the report. 

Table 6-2. ANL-E external dosimetry history. Year Holder Beta/gamma Neutron Extremity

Detector Filtration Processor Detector Processor Detector Processor

Deep Shallow

1945 University

of

Chicago

(two

filter) 

Film,

Type

Unknown

1 mm Cd Open

(28 

mg/cm2

Wrapper)

ANL Film ANL

1946

1947

1948

1949

1952 “catastrophic” 

1953 film added Kodak 

NTA

ANL Film rings

Ring, 

wrist, & 

head films

R. C.

Scientific

& Others

Wrist, head 

same as 

whole body

1954

1955 Tracerlab,

1956 Inc. 

1957

1958 Landauer,

1959 DuPont 553 Inc. 

1960 See text DuPont 558 See text See text Atomic

Film Badge

Corp.

Kodak 

NTA

Atomic

Film Badge

Corp.

1961

1962 Hanford

Type

(four

filter) 

DuPont 558 0.0191” 

Al

0.005” 

Ag

0.0395” 

Cd

Open

1963 Window ANL Kodak 

NTA

ANL

1964

1965 DuPont 553

1966 DuPont 558

1970 Kodak Type 3 TLD ring ANL

1971 Kodak Type 2

1973 Kodak 

NTA &

SSTR

1974

1975

1976

1977 Kodak 

NTA &

Albedo 

TLD

1978 Current 

TLD

Ring

1979-

1987

1988 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 TLD

Harshaw

TLD 

600/700

1989

1990

1991

1992

1993

1994

2004

The 1989 and later records summaries (Radiation Exposure Record) report external dosimetry results 

as: 

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

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

• Code NM for not monitored

• Code ND for not detectable

The reported values for the early years when the dosimeters were calibrated to exposure assume 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.
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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 2005).

6.3.1

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 hand-entered forms. From 1964 to 

1975, a handwritten form was again used and only shows 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.

Dosimetry Records Systems

An individual worker’s records can contain one or more of the following depending on the years of 

employment. Dates given are approximate. Attachment A contains a redacted sample with personal 

identifiers removed from each record. 

Individual Meter Record (1942 to mid-1953) – Hand Entered (Figure A-3)

• Each card covers 1 wk. 

• The values are in roentgen rather than milliroentgen. 

• Daily readings for PICs and a weekly total, number of times worn, and Discg (entry placed 

here for accidental discharge).

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

• Location worn (e.g., Site A).

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

• Results for each wear period.

• The values are in milliroentgens (γ) and arbitrary units (β). 

• From mid-1953 (fiscal year 1954) to 1956, the forms are for one fiscal year. From 1956 to 

1961 they are by 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 dose accumulated 

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.
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• For betas, when the change from arbitrary units (a.u.; see Section 6.8.2.1) 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 31st and read late.

• The forms were divided into 4 qtr of 13 wk each. Each quarter was subtotaled.

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

nonpenetrating (a.u. or beta) doses. 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 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.

• Biweekly total (BIWK TOTAL) and running totals for the calendar year (CAL YR) and last 14 

weeks (QTD TOTAL) are given. 

• The values are in milliroentgens (γ), mrad (β), and mrem (neutron). 

• 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 may 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 mo of data.

• The values are in milliroentgens (γ), mrad (β), and mrem (neutron). 

• 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. 
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• 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 (WK DOS, i.e., PIC) results. 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 is 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, gamma, and neutron).

• The values are in mrem. 

• 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 wide computer-paper form is copied for 

inclusion in the hard-copy records. The worker’s name is cut off, but the social security 

number shows. 

• 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 rover, gamma, and neutron), and Remarks (RE). A common remark was 

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

• The values are in milliroentgens (γ), mrad (β), and mrem (neutron). 

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

is included in the record, the moving-quarter value will always be the forth quarter total.

Master External Radiation Exposure Report (1989 to 2002) – 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.
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• Column headings are location code (SECTION), BADGE TYPE, ISSUE DATE, RETURN 

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

deep and neutron), 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 (also 

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 values are in mrem. 

• 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 the 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 values are in 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, i.e., extra badge worn on wrist or body.

• The values are in milliroentgens and a.u.

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

• Information includes date, type (film badge or dosimeter), area, time (worn), position, mR 

under filter, 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 is used for 

one badge only.

• Includes date(s) and time(s) worn, area (building), position worn, type of film (beta-gamma or 

neutron), and remarks. The remarks can indicate the reason for the badge.
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 17 of 69

• The results are in milliroentgens, a.u., and millirem (neutron).

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

(Figure A-13) 

• The records contain supplemental dosimeter readings, i.e. beta-gamma badge worn on right 

wrist.

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

were noted corresponding to the dosimeters in use over the years: Two-element beta-gamma, 

four-element beta-gamma, NTA film, and albedo TLDs.

• The values are typically in milliroentgen, mrad, and millirem (neutron) and are 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 (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 beta-gamma, 

gamma, and neutron exposures. 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 is included to indicate that more 

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

• Doses are in 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).

• 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, the level in millirem, the type, whether the dose is valid (yes/no), 

and whether it was anticipated (yes/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.
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• 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 workers terminated employment 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 

received in shipment. The individuals’ detail cards were punched with code 89 and zero doses were 

recorded. 

6.3.2

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 to not double count the doses for these periods.

Observed Data Discrepancies

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

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

6.4 HISTORICAL ADMINISTRATIVE PRACTICES

6.4.1

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

individuals entering an active area were required to wear personnel monitoring devices found at the 

entrance. Areas spelled out as active included Building 330 (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 containing neutron film were required at CP-5 

(ANL 1956a).

Badged Population

Information developed in a 1982 survey for a DOE health and mortality study (Strom 1982) indicates 

that early on everyone was badged. By 1965, nearly all employees were still badged. By the early 

1970s, the site health physicists assigned badges based on the exposure potential. By 1982 it was 

noted that approximately one-third of the workers were badged.

In the 1973 to 1984 revisions of the site Health and Safety Manual (ANL 1973), 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 producing 

ionizing radiation was operated. Each person entering 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).
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 19 of 69

6.4.2

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

was 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 twenty special cases at the 

new Chemistry Building at the University of Chicago (Rose 1951).

Badge Exchange Frequency

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/yr) (Bleiler 1968a, 1968b, 1970, 1973a, 1974).

Due to a work force reduction in the Personal Monitoring Group biweekly exchanges were eliminated 

and all film badges were placed on a 4-wk 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 may 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 done 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 are exchanged quarterly at that facility (ANL 1994).

Table 6-3 provides the claimant-favorable default dosimetry exchange frequencies to be used for 

dose reconstructions.

Table 6-3. Default dosimeter exchange frequencies. 

Years General population Others

1945–1959 Weekly

1960–1973 Biweekly

1974 4-Week (13 periods/yr)

1975–2005 Monthly (12 periods/yr) 1994 APS quarterly

6.4.3

No workplace-specific calibration factors have been found. 

Field-Specific Calibration Factors

6.4.4

The specifications for the multi-element 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).

Minimum Reported Dose

The specifications for the Atomic Film Badge Corporation multi-element 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. Any 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.
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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 dosimeter described is not the Panasonic UD-808AS tested by DOELAP in 1992, 

but a Hankins-type (one TLD-600 and one TLD-700 enclosed in a polybox) (ANL 1996). 

6.5 COMMON ISSUES

6.5.1

The available dosimetry records do not provide individual dosimeter results from 1964 to 1975. As 

mentioned above, there is some overlap in the dose records for 1964; so 1964 detail might be 

available for some workers. In many cases it will therefore be necessary to estimate the dosimeter 

exchange frequency from 1964 to 1975. It appears that ANL-E did not generally use different 

exchange frequencies based on job categories, so assumption of a single frequency by time 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/yr). For 1974 the exchange 

frequency should be assumed to be every 4 wk (13 exchanges/yr). For 1975 the exchange frequency 

should be assumed to be monthly (12 exchanges/yr). For other years, determining the number of 

zero readings should be straightforward.

Number of Zero Readings 

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 417 mrem/mo. The minimum number of months in 

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

months would be seven, and the minimum would be zero 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 seven (NIOSH 2002). Table 6-4 lists historical dose limits. 

Table 6-4. Historical dose limits.

Period Daily or weekly limit Quarterly or annual limits:

1945–1955 Beta & gamma, 100 mrem/d

(Nickson 1946, Gilbreath undated)

Not applicable

1956–1972 Gamma 300 mrem/w k (AEC 1958)

Beta & gamma (sum), 500 mrep/w k (ANL 1956a) 

3 rem/qtr, 15 rem 

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

1973–1988 Not applicable 3 rem/qtr (A NL 1973)

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

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

(DOE 1988)

1993-2005 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 

(2002), it is possible to develop a claimant-favorable estimate of the number of zeros for calculation of 

the missed dose.
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 21 of 69

6.5.2

If the employee’s record contains discrepancies, it is claimant-favorable 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. Since mid-1953 the laboratory

has routinely used milliroentgens and millirem. It is highly unlikely that a record would show a dose 

greater than the quarterly or annual limit without an additional record indicating an overexposure.

Discrepancies 

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

1945–1955 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)

1989–2005 5 1 (a) (a) (a) 0.417 1.25

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

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

the dose reconstruction. The dose reconstructor should use his or her 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 claimant-favorable action is to incorporate the correction 

in the dose used for reconstruction.

6.5.3

If the dosimetry history contains a missing entry, this probably indicates that the individual missed the 

dosimeter exchange and that the next dosimeter includes the dose from both exchange periods. A 

less likely but claimant-favorable assumption is that the badge was lost and no dose was assigned for 

that period. Dose should be assigned for that period using dosimetry data preceding and following 

that period in consideration of the approach of Watson et al. (1994).

Missing Entry 

6.5.4

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

assumption is that all exposures are for anterior–posterior geometry.

Exposure Geometry 

6.6 PHOTON DOSE

6.6.1

The NIOSH IREP probability of causation program contains three photon energy ranges: Below 30 

keV, 30 to 250 keV, and above-250 keV (NIOSH 2002). Separation of the dose from each energy 

band is necessary. 

Energy Groups 

Very little spectroscopy data to indicate gamma spectrum in ANL-E work areas have been found. 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 a claimant-favorable estimate of the energy band. Plutonium facilities 

have mostly 17 keV photons with some 59 keV photons from 241Am. ORAUT (2005b) suggests that a 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 22 of 69

correction factor of 0.6 should be applied to the non-penetrating dose determined from early film 

dosimeters and that non-penetrating 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 LINACs), 

bremsstrahlung photons dominate the secondary radiation field. Thick shields of concrete or other 

materials will result in photons in the MeV energy range. For proton or positive ion accelerators (ZGS, 

IPNS, Van de Graaff) neutrons generally constitute the greatest hazard. Thick shields designed for 

the neutron hazard will 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, etc. One reaction of particular 

importance is the thermal neutron capture of sodium in the concrete of the accelerator shielding 

(NCRP 2003). This reaction [

23Na(n,γ)24Na] produces a radioactive isotope that decays with a 15-hr

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 low-energy photon hazards, but these appear to 

have been generally used 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. 

6.6.2

Shallow doses may be determined using open w indow readings 

for the early film badges and the methods in ORA UT (2005b)

6.6.2.1 Units for Dose Conversion Factors for Dose to Organ Dose

Calibration Factor

The early dosimeters were calibrated in roentgens (R) (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. 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 23 of 69

Table 6-7. Photon dose units for use with 

organ dose conversion factors. 

Years Dose unit

1945–1988 R

1989–2005 Hp(10)

6.6.3

Section 2.1.2 of NIOSH (2002) recommends the use of the LOD/2 method for determining missed 

dose.

Missed Dose

6.6.3.1 Limit of Detection 

The film badge initially developed at the University of Chicago was used at other U.S. Atomic Energy 

Commission (AEC, a DOE predecessor agency) sites. All of these badges used X-ray film 

surrounded by a metal badge holder, and each had an open window and an area covered with 1 mm 

of silver, tin, or cadmium (ORAUT 2004a). Information found at ANL-E indicated 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.

Only 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 (ANL 1956a). The specifications for the multi-element 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

1945–1959 50 mrem

1960–1988 25 mrem

1989–2005 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 

2002, 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 

log normal 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 and Section 2.1.2.3 of 

NIOSH (2002) discuss this estimate. In this case, the estimate is assumed to have a log normal

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 used at ANL-E is similar to those used at other sites. The response 

of this dosimeter is addressed in the Savannah River TBD (ORAUT 2003). These documents 

address the significant over-response of film to low photon energies. Dosimeter calibration was 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 24 of 69

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

factor is appropriate for this dosimetry system.

The multi-element 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.

6.6.4

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

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

Geometry - Angular Dependence 

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 traveled 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 (ANL 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 OCAS guidance. 

6.6.5

NIOSH (2002) 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.

Uncertainty 

6.6.5.1 Film 

ANL-E used film to measure photons between 1945 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 reports that both films have 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.
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 25 of 69

The method in NIOSH (2002) 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 the Rocky Flats external dosimetry TBD (ORAUT 2004b) 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 is 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 Cd 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 reduced at higher exposures, the NRC methodology recognizes that 

additional uncertainty contributed by 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 limitation has been 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

1945-1959

DuPont 508 film

Four-element badge

1960-1988

DOELAP-accredited TLD 

1989-2005

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, 

the current beta-gamma TLD (Panasonic UD-814SA4) responds within 10% of the delivered dose 

equivalent for most single radiation fields and within 20% of the mixed radiation fields tested by 

DOELAP (ANL 1996). 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 (ANL 1996).

The standard deviation of the null readings associated with the DOELAP-accredited Panasonic 

dosimeter was not documented. Therefore, the analysis used the Simplified Dosimetry Uncertainty 

calculation recommended by NIOSH (2002, 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 kermas was 

estimated to be 10%. This is a claimant-favorable assumption for this dosimetry system. Table 6-9 

lists the upper 95% confidence doses. 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 26 of 69

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 2002). 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 (ANL 1956a). NTA film continued in 

general use until the late 1980s, when albedo TLD dosimeters replaced it (McDowell-Boyer 2005).

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

procedure originated 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 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 solid state track recorder (SSTR) neutron dosimeter were 

undertaken at the Zero Power Reactor (ZPR) VI facility where there was a concern that the 

moderated plutonium light element neutrons 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. However, 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 that this system was brought on line. Table 6-10 summarizes the 

methods employed by ANL-E to assess personnel neutron doses. The dates shown 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 selfreading pocket chamber

1988–2005 ANL-E design, TLD-600 and -700 in a polybox 

inserted in the Panasonic dosimeter, similar to a 

Hankins type
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 27 of 69

6.7.1

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

factors provided in Appendix B of NIOSH (2002). These energy groups and the associated radiation 

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

60 (ICRP 1991) are:

Energy Groups 

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, effective, or energy range of the neutrons. Some of the 

measurements were not appropriate for dose reconstruction because they were made under unusual 

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 limited number of facilities at ANL-E. These include 

reactors, high-energy accelerators, and materials-handling areas where alpha emitters could interact 

with light elements. Facilities such as the 60-Inch Cyclotron, the Electron Linear Accelerator (LINAC), 

and other low-energy accelerators do not present a neutron hazard outside of their shielding under 

normal operating conditions (Coulson 1989). Neutrons rarely contribute more than 20% to the annual 

collective dose at accelerator facilities (Coulson 1989). 

The Zero Gradient Synchrotron (ZGS) was a 12.5-GeV proton accelerator that operated between 

1963 and 1979. No useful spectral information from ANL-E for potentially occupied areas was 

located. The LANL External Dosimetry TBD (ORAUT 2005a) Figure 6-8 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. The SPUNIT code result was chosen to simulate neutron spectra at the 

ZGS absent any better information. Using EXCEL a spectrum was generated to visually match that in 

Figure 6-8 of the LANL TBD. The Ing and Makra (1978) parameterization for the NCRP 38 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 which is shown in Table 6-11 was determined from this spectrum.

The 500-MeV Intense Pulsed Neutron Source (IPNS) is capable of generating neutrons in the energy 

range from 0.025 eV to greater than 20 MeV. Outside of 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, but additional shielding was added above the enriched 235U 

target (Coulson 1989). To establish the dose equivalent fractions in the IREP neutron energy 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 28 of 69

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.

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

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

for the LAMPF ER-1 spectrum. 

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

Energy Group Percent of Dose equivalent

Zero 

Power

Reactors

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 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 29 of 69

restricted when accelerating certain light ions. No skyshine has been observed (Coulson 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 1989). No 

spectral information was located.

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 being measured (Dyer and Mundis 1966).

The Advanced Photon Source (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 of shielded areas (DOE 2003, 

ANL 1994, Moe et al. undated).

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 and other isotopes and (α,n) reactions on light elements leads to significant doses near the 

reactors even when shutdown. 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 working between the halves 

experienced significant dose equivalent rates, especially when working near a gap in the shielding, 

which was necessary to 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 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 38 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 

obtained 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 INEEL (ANL 1956b, ANL 

1961). Therefore, the recommendations in the TBD for INEEL are adopted for use here (ORAUT 

2004c). Similarly the recommendations in the Los Alamos TBD are adopted for plutonium chemistry 

and metallurgy (ORAUT 2005a). 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 >0.10–2.0 MeV energy interval. Table 6-11 also presents these results.

6.7.1.2 Reported Dose to Energy Groups

NIOSH (2002) requires that the neutrons be apportioned into the energy groups used in ICRP 

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

assumed that the neutron doses reported 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 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 30 of 69

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

Report 38 dose equivalents to ICRP 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 60 dose multipliers from Table 6-12 were used with the estimated neutron spectra in Table 

6-11 to develop multipliers which can be used to determine the neutron equivalent dose for each 

IREP energy interval. The corrected neutron dose reported in the claimant's dose record is 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. 

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

intervals. 

Energy Group Multipliers used to assign to IREP neutron energy intervals

Zero 

Power

Reactors

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 2

>0.10–2.0 MeV (wR=20) 1.3 1.0 0.18 0.38 1.0 0 

>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

6.7.2.1 Units for Dose Conversion Factors for Dose to Organ Dose

Calibration Factor

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 at 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. 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 31 of 69

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

correction discussed in Section 6.7.3.2.

The doses reported for the SSTRs and interim albedo dosimeters 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 ANL-E Technical Basis Document for External Dosimetry states that, “Calibration of 

the ANL-E albedo dosimeter has been conducted at various locations both on-site and off-site” (ANL

1996) and goes on to state the LOD for exposure to moderated 252Cf. From these statements the 

albedo TLDs at ANL-E were apparently 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

1945–1988 Ambient dose, H*(10)

1989–2005 Deep dose equivalent, Hp,slab(10)

6.7.3

6.7.3.1 Limit of Detection

Missed Dose

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 

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

contractor reported all doses read 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 minimum reported 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 et al. 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 claimant-favorable 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 given are approximate.
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 32 of 69

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

1988–2005 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 author stated that chronic low-level exposures in areas where personnel are sometimes 

located are not recorded by the film (Steele 1965).

A 1966 investigation pointed out that the majority of neutrons in areas 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 work force reduction in the Personal Monitoring Group led to a proposal to eliminate NTA 

film with the exception of the ZGS facility, which could indicate that NTA film 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 1973 b).

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.

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.

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-3 ́, CP-5 reactors 1.5

ZPR Pu-fueled reactors 4

Pu-handling facilities 4
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 33 of 69

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

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 is 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. 

Geometry-Angular Dependence 

No information is available for the albedo TLDs used at ANL-E.

6.7.5

6.7.5.1 Film

Uncertainty

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 exposed to 300 mrem from a Pu-Be source were processed. Of those, 199 films 

(47.6%) were not within ±25% of the expected value (Strom 1982). This same contractor’s 

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 (Hankins 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 for the NTA film dosimeter conducted by the 

University of Michigan (Neal 1980). This data showed an average standard deviation of 10% for high 

doses (301-5000 mrem) of 252Cf. However, for lower doses (100-300 mrem) the average standard 

deviation was 64%. The standard deviation of both ranges was 48%. Since this performance may 

not have been typical for earlier periods, a standard deviation of 50% was selected for this analysis. 

Again the simplified uncertainty analysis of NIOSH (2002) was used. The results are shown in Table 

6-17. 

6.7.5.2 Thermoluminescent Dosimeter

The LOD of 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 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 34 of 69

found for the other interim albedo dosimeters. Until such data is found the values for the Hosger 

albedo TLD should be used.

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

The current albedo system did not undergo DOELAP testing in 1988 or 1991. The lower limit of 

detection for this badge is 10 mrem in normal personnel dosimetry practice (ANL 1996). Since this 

dosimeter involves reading 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

6.7.6.1 The sequence of application of the neutron correction factors for NTA film is:

Application of the Neutron Correction Factors

• Calculate the missed dose (Section 6.7.3.1), if applicable

• Apply the correction factor for unmonitored energy range (Section 6.7.3.2)

• Apply the correction for the neutron quality factor for each IREP energy range (Table 6-13, 

Section 6.7.1.2)

6.7.6.2 Example

• An ANL-E worker in Bldg. 200 (plutonium chemistry facility) is monitored by NTA film for 6 

months in 1960 and receives doses of 0 mrem, 0 mrem, 100 mrem, 300 mrem, 0 mrem, and 0 

mrem.

• The missed dose is LOD/2 (25 mrem) X 4 = 100 mrem. The total for the period is 100 mrem 

(missed dose) + 100 mrem + 300 mrem = 500 mrem.

• Since the neutron film did not respond to energies below about 800 keV, the correction for the 

unmonitored energy range is 500 mrem X 4 (Table 6-16) = 2000 mrem.

• The doses for each IREP energy group from Table 6-13 for a plutonium chemistry and 

metallurgy facility are:

o < 0.01 MeV, 2000 mrem X 0 = 0 mrem
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 35 of 69

o 0.01 to 0.10 MeV, 2000 mrem X 0.2 = 400 mrem

o > 0.10 to 2 MeV, 2000 mrem X 1 = 2000 mrem

o > 2 to 20 MeV, 2000 mrem X 0.52 = 1040 mrem

o > 20 MeV, 2000 mrem X 0 = 0 mrem

6.8 ELECTRON DOSE 

The earliest maximum permissible exposure limits published at ANL-E recognized the potential 

hazards of beta exposures of the skin (Nickson 1946). The pioneering work done there 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 (more than 1 MeV) 

betas, cannot be ruled out. The dose reconstructor should be aware that early results from ring and

wrist dosimeters may have been reported on separate forms and may need to be considered for skin 

dose to the hands and forearms.

6.8.1

Although documentation is lacking, it appears that for the two-element badge (1946 to approximately 

1960) the open window doses are nearly always equal to or greater than the shielded window doses, 

indicating that the open window reading includes both penetrating and non-penetrating 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. Thus penetrating and non-penetrating 

(beta only) radiations are reported separately after approximately 1960. No detail on the reporting of 

early film or TLD ring dosimeters was located. The current TLD system reports the shallow or skin 

dose as the sum of the penetrating and non-penetrating dose. ORAUT (2005b) suggests that a 

correction factor of 0.6 should be applied to the non-penetrating dose determined from early film 

dosimeters and that non-penetrating 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.

Energy Groups

All measured and missed non-penetrating doses that are considered electrons should be assigned to 

the > 15 keV IREP energy group.

6.8.2

During the use of the two-element film badge, the term arbitrary units or a.u. was used to report the 

open window dose. 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 caused by exposure to unknown beta-gamma and/or X-ray 

radiation to the equivalent density of a film in the badge window position caused by exposure to a 

known 60Co or 226Ra source in milliroentgens (Strom 1982). The arbitrary units were later correlated 

with an exposure to uranium. Although the open window in these dosimeters over responded to low 

energy photons, the records may underreport the doses from electrons (see 6.8.3.1).

Calibration Factor - Units for Dose Conversion Factors for Dose to Organ Dose
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 36 of 69

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 radiation 

weighting factor for electrons at all energies is 1 (ICRP 1991), thus, reported beta doses are 

equivalent to mrem.

ANL-E’s current beta-gamma dosimeter, 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 (ANL 1996).

6.8.3

In 1956, the two-element film badge was capable of measuring doses from 50 mR to 500 R (ANL 

1956a). No separate LOD was given for beta radiation, so it is assumed to be the same. The 

specifications for the multi-element 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 (ANL 1996). Table 6-18 lists the beta 

LODs for the ANL-E dosimeters. Specialized extremity dosimetry used 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 (ANL 1996).

Missed Dose - Limit of Detection

Table 6-18. Beta dosimeter LODs. 

Years LOD

1945–1959 50 mrem

1960–1988 25 mrem

1989–2005 30 mrem whole body

10 mrem TLD ring

6.8.3.1 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 2004c). Since the dosimeters used 

at the two sites were similar and few details are available for ANL-E, the INEEL analysis is used in 

part here. Table 6-19 provides the fraction of beta dose recorded for the various dosimeters. To 

determine the corrected beta dose, the dose reconstructor should divide the non-penetrating results 

by the values in the last column of Table 6-19. This result will probably be an overestimate since 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 underreporting. 

Dosimeter

System

Period Cover

(mg/cm2)

Detector thickness

(mg/cm2)

Estimated fraction 

of dose reported

Two-filter film

Early film ring

1945-1959 50 50 0.49

Multi-filter film 1960-1988 100 50 0.35

Old TLD ring 1970-1977 104 240 0.21

Current TLD ring 1978-2005 18 89 0.56

Panasonic TLD 1989-2005 16 15 0.78
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 37 of 69

6.8.4

6.8.4.1 Angular Dependence 

Geometry

The sensitive dosimeter elements are 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 (ANL 1996). Because the angle of exposure is more or less random with 

this arrangement, no correction is possible.

6.8.4.2 Exposure Geometry 

Due to exposure geometry, recorded beta doses may 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

The method in NIOSH (2002) 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).

Uncertainty 

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 1945 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 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.

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 ANL-E Technical Basis Document, 

the current beta-gamma TLD (Panasonic UD-814SA4) responds within 20% of the mixed radiation 

fields tested by DOELAP. The largest positive deviations were for mixtures of X-rays and beta 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 38 of 69

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 (ANL 1996).

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

The standard deviation of the null readings associated with the DOELAP-accredited Panasonic 

dosimeter was not documented. The analysis therefore uses the simplified dosimetry uncertainty 

calculation from NIOSH (2002, Section 2.1.1.3.3), where the critical level Lc is the LOD estimated, as 

described in Section 6.6.3.1. The percent standard error at high air kermas was estimated to be 10%. 

This is a claimant-favorable assumption for this dosimetry system. Table 6-20 lists the upper 95% 

confidence doses.

6.8.6

Skin contamination incidents have occurred throughout the history of the site. Skin decontamination 

methods and procedures were generally well documented, but the specific forms used 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 AEC, Energy Research and Development 

Agency (ERDA), and DOE sites where skin contaminations were documented in the health and/or 

dosimetry records. The Environment, Safety & Health [ES&H] Manual currently requires recording of 

the nonuniform shallow dose equivalent to the skin caused by contamination on the skin if the dose is 

equal to or greater than 2% of the limit specified for the skin in 10 C.F.R. 835.702(b) (ANL 2004).

Skin Contamination

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 used in addition to portable monitors (ANL 1956a).

Natural and depleted uranium was used and processed at ANL-E, and their progeny potentially 

contained in the material would result in beta exposures to the skin. Other high-energy beta emitters 

such as 90Sr/90Y were present at the 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, and dose reconstructors should estimate it in 

accordance with Section 2.3.3 of NIOSH (2002). The contamination report will probably not indicate 

the length of time that the contamination was present on the skin. A reasonable claimant-favorable 

assumption is that the contamination was present for 4 hr 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, which normally resulted

in removal of most of the contamination. If these initial efforts failed to remove enough of the 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 39 of 69

contamination, the individual was taken to the occupational medicine group where further 

decontamination efforts were made under medical supervision (ANL 1956a; ANL 1973).

Dose reconstructors could use software such as VARSKIN (recommended in NIOSH 2002) or other 

appropriate means for this calculation. 

6.9 UNMONITORED INDIVIDUALS

6.9.1

The available information indicates all workers at ANL-E were badged in the early years. By 1965, 

nearly every employee was still badged. By the early 1970s, ANL-E apparently had started to badge 

workers based on their exposure potentials (Section 6.4.1). The dosimeters used by workers 

assigned to a facility were stored at the entrance to that facility. Rover dosimeters were assigned to 

visiting radiological workers. Under this system, it seems unlikely that anyone would have routinely

entered radiological facilities without a monitoring device. However, the following paragraph provides 

a claimant-favorable method of assigning dose to unmonitored individuals.

In Radiologically Controlled Areas

From 1973 to 1992, the radiation protection guideline was 3 rem/qtr (ANL 1973). Areas where 

personnel were required to wear monitoring devices were established at 10% of this value. 

Therefore, the missed dose estimate for unbadged individuals in radiologically controlled areas would 

be one-half of 10% of 3 rem/qtr, or 600 mrem/yr. A lognormal distribution with geometric mean of 600 

mrem should be assumed, with the upper 95% dose estimate for these individuals of 10% of the 

guideline (1.2 rem/yr) (NIOSH 2002). After 1992, a monitoring threshold of 100 mrem/yr should be 

used.

6.9.2

After the early 1970s, those who did not work in radiological areas would not have been badged. 

Visitor dosimeters were stored at the entrances to the various facilities for use by personnel who were 

not considered radiological workers.

Outside of Radiologically Controlled Areas

For individuals working outside the radiologically controlled areas, environmental exposure would be a 

better estimate of their exposure (see ORAUT 2006).

6.10 EXTREMITY DOSIMETRY

Monitoring exposure to the skin was important very early in ANL-E’s history and maximum permissible 

levels were established (Nickson 1946), but it is not clear what monitoring methods were in use in 

what periods. In the early years of the site, body badges were apparently fitted with alternate 

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-wk 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 (ANL 1956a). The 

badge was likely similar to the whole-body film badge (two-element design with an open window and 
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 40 of 69

a 1-mm cadmium filter). Wrist badges containing neutron film were available and were required at 

CP-5 (ANL 1956a).

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 in. by 0.125 in. 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 (ANL 1996). Curves of net counts versus 

exposure show that the difference in response of the four elements is greatest at the lower energies, 

as could be expected (ANL 1996). From the ANL-E technical basis document it is not clear 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.
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 41 of 69

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A for 14 MeV Neutrons,” memorandum to J. R. Novak, September 22, Argonne National 

Laboratory, Argonne, Illinois.

Till, H. A., 1966, “Evaluation of Bendix Tissue Equivalent Dosimeters,” memorandum to J. R. Novak, 

August 26, Argonne National Laboratory, Argonne, Illinois.
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 45 of 69

Watson, 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.

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

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

Northwest Laboratory, Richland, Washington. 

Yule, T. J., 1971, “Spectrum Measurements Between the Halves of ZPR-6 Assembly 7,” Applied 

Physics Division Annual Report, July1, 1970 to June 30 1971, Argonne National Laboratory, 

Argonne, Illinois.

Yule, T. J., R. J. Armani, and R. Gold, 1972, “The Adequacy of Solid-State-Track-Recorder FastNeutron Personnel Dosimeters for Use in ZPR Environments,” Applied Physics Division 

Annual Report, July 1, 1971 to June 30, 1972, Argonne National Laboratory, Argonne, Illinois.

Yule, T. J., 1973, “Report on the Fourth Workshop on Personnel Neutron Dosimetry Held at Miami 

Beach, June 13-15, 1973,” memorandum to Distribution, June 19, Argonne National 

Laboratory, Argonne, Illinois.
Document No. ORAUT-TKBS-0036-6 Revision No. 00 Effective Date: 02/09/2006 Page 46 of 69

GLOSSARY

Albedo dosimeter

A personnel dosimeter which measures the dose equivalent from neutrons incident on the 

body by detecting the moderated neutrons reflected from the body.

Atomic Energy Commission (AEC)

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

U.S. Department of Energy.

arbitrary unit

A unit used to compare the density of a film in the badge window position caused by exposure 

to unknown beta-gamma and/or x-radiation to the equivalent density of a film in the badge 

window position caused by exposure to a known 60Co or 226Ra source expressed in mR units.

beta dose

A designation (i.e., beta) on some external dose records referring to the dose from less 

energetic beta, X-ray, and/or gamma radiation (see open window, or shallow dose).

beta particle

An electron or positron emitted spontaneously at high velocity from the nuclei of certain 

radioactive elements. Most of the direct fission products are (negative) beta emitters.

deep absorbed dose

The absorbed dose at the depth of 1.0 cm in a material of specified geometry and

composition.

deep dose equivalent (Hd) 

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

dose equivalent (H)

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

The special unit is the rem, the International System (SI) unit is the sievert (1 sievert = 100 

rem).

DOELAP

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

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

dosimeter

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

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

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

dosimeter, neutron film dosimeter, thermoluminescent dosimeter.)

dosimetry

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

from external or internal sources of radiation.
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dosimetry system

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

extremities. This includes the fabrication, assignment, and processing of dosimeters as well as 

interpretation and documentation of the results.

exchange period (frequency)

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

exposure

As used in the technical sense, exposure refers to a measure expressed in roentgens (R) of

the ionization produced by photons (i.e., gamma and X-rays) in air. In the generic sense, 

ionizing radiation applied to matter.

extremity

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

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

toes.

field-specific calibration

Dosimeter calibration based on radiation types, intensities, and energies in the work

environment.

film

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

When developed, the film has an image caused by radiation that can be measured using an

optical densitometer.

film density

See optical density.

film dosimeter

A small packet of film within a holder that attaches to a wearer. Also referred to in early site 

documents as a “badge” or “badge meter.”

fission

The splitting of a heavy atomic nucleus, accompanied by the release of energy.

gamma rays

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

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

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

not originate in the nucleus.

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through

interactions with matter.

isotope

Elements having the same atomic number but different atomic weights; identical chemically

but having different physical and nuclear properties.
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minimum reporting level

Based on a policy decision, the minimum dose level that is routinely recorded.

neutron

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

neutron film dosimeter

A film dosimeter that contains a Kodak Nuclear Track Emulsion, Type A, film packet.

Nuclear Track Emulsion, Type A (NTA)

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

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

1000X-power microscope or a projection capability.

open window

Designation on film dosimeter reports that implies the use of little (i.e., only film wrapper) 

shielding. Commonly used to label the film response corresponding to the open-window area.

optical density

The quantitative measurement of photographic blackening; density defined as D = Log10 (Io/I).

pencil dosimeter

A type of ionization chamber used by personnel to measure radiation dose. Often referred to 

in early site records as a “pocket meter” or simply “dosimeter” as opposed to the term “badge” 

used for film badges. Also pocket dosimeter, Shonka-type dosimeter, and pocket ionization 

chamber (PIC).

personal dose equivalent Hp(d)

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

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

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

personal dose equivalent, slab Hp,slab(d)

Represents the dose equivalent in a tissue-equivalent slab phantom at an appropriate depth 

(d). This quantity is used for the calibration of personnel dosimeters.

photon

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

photon - X-ray

Electromagnetic radiation of energies between 10 keV and 100 keV whose source can be an

X-ray machine or radioisotope.

quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose. Also appears as “QF” 

in some site documents.

radiation

Alpha, beta, neutron, and photon radiation.

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

neutrons from unstable nuclei.

radionuclide

A radioactive isotope of an element, distinguished by atomic number, atomic weight, and

energy state.

rem

A unit of dose equivalent equal to the product of the number of rad absorbed and the quality

factor.

Roentgen (R or r)

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

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

An exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for

higher (~>100 keV) energy photons.

rover dosimeter or rover

A pencil dosimeter that was not permanently assigned. Worn by visitors or personnel not 

normally assigned to an area.

shallow dose equivalent (Hs) 

Dose equivalent at a depth of 0.007 cm in tissue.

shielding

Any material or obstruction that absorbs (or attenuates) radiation and thus tends to protect

personnel or materials from radiation.

skin dose

Absorbed dose at a tissue depth of 7 mg/cm2

. 

thermoluminescence

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

thermoluminescent dosimeter (TLD)

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

The measurement of this light provides a measurement of absorbed dose.

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1000 mg/cm2

); however,

also used to refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin. Photon generated from 

acceleration of electron or positron in x-ray generator or synchrotron radiation storage ring.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 1 of 17

Figure A-1. External Radiation Exposure Record (page 1 of 2). 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 2 of 17

Figure A-1. External Radiation Exposure Record (page 2 of 2). 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 3 of 17

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

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 4 of 17

Figure A-3. Individual Meter Record. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 5 of 17

Figure A-4. Cumulative Exposure Record. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 6 of 17

Figure A-5. Accumulated Exposure Report. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 7 of 17

Figure A-6. Semi-Annual Accumulative External Exposure Report. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 8 of 17

Figure A-7. Whole Body Radiation Exposure Summary. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 9 of 17

Figure A-8. Master Radiation Exposure File
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 10 of 17

Figure A-9. Master External Radiation Exposure Report. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 11 of 17

Figure A-10. Neutron Exposure Report. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 12 of 17

Figure A-11. Special Exposure Records. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 13 of 17

Figure A-12. Special Film Request. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 14 of 17

Figure A-13. Special Meter Assignment and Radiation Exposure Report. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 15 of 17

Figure A-14. Special TLD Finger Exposure Report. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 16 of 17

Figure A-15. Personnel Dose Equivalent Evaluation – Level 1. 
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 17 of 17

Figure A-16. Termination Occupational Exposure Report. 
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ATTACHMENT B

FACILITIES BY BUILDING NUMBER AND SECTION CODE

Page 1 of 3 

Table B-1. Facilities by building number and section codea

. 

Building

Section codes

Facility name

Years of operation

1980 2005 (approximate) Additional information Facility type

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-3 ́ 1944–1954 CP-3 was heavy-water moderated, natural uranium 

fuel. Natural U fuel replaced with enriched and 

renamed CP-3 ́ 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 (M Wing 

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

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 Advanced Materials Fabrication 

Facility (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
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ATTACHMENT B

FACILITIES BY BUILDING NUMBER AND SECTION CODE

Page 2 of 3

Table B-1 (Continued). Facilities by building number and section code. 

Building

Section codes

1980 2005 Facility name Years of operation Additional information Facility type

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 Cs-137.

Radioactive 

Materials

303 Mixed Waste Storage Facility 

(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

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 2005 

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 Reactor 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 decomissioning of 

reactor completed 2000.

Reactor

331 QW QW EBWR 1956–1967 Reactor removed from shell in 1995 during 

decontamination and decomissioning. 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
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ATTACHMENT B

FACILITIES BY BUILDING NUMBER AND SECTION CODE

Page 3 of 3

Table B-1 (Continued). Facilities by building number and section code. 

Building

Section codes

1980 2005 Facility name Years of operation Additional information Facility type

340 MP Experimental Animal 

Quarters

1950– Housed animals irradiated in Bldg. 202. Bone cancer 

research 1963 to 1980.

Radioactive 

Materials

350 SH SR, VN Fuel Fabrication Facility

(now New Brunswick Lab, 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

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 pps, at 20 MeV.

Accelerator

361, 391, 375 SA, SB, SC, SD, SE, 

SF, SG, SK, SQ, SS, 

SU, SV, SX, SZ, WC

IPNS 1981– Use s 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. blank = none.