ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

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Document Title:

Argonne National Laboratory – West –

Occupational External Dose

Document Number: ORAUT-TKBS-0026-6

Revision: 01

Effective Date: 05/15/2007

Type of Document: TBD

Supersedes: Revision 00

Subject Expert(s): Steven L. Bump

Site Expert(s): Norman D. Rohrig

Approval: Signature on File Approval Date: 04/24/2007

Jo Ann M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 04/24/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 04/24/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 04/24/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 05/15/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 2 of 57

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

09/30/2004 00 New technical basis document for the Argonne National Laboratory -

West - Occupational External Dosimetry. First approved issue. 

Initiated by Norman D. Rohrig.

05/15/2007 01 Approved revision for biennial review. Incorporates Attributions and 

Annotations. Constitutes a total rewrite of the document. 

Incorporates formal internal and NIOSH review comments. This 

revision results in no change to the assigned dose and no PER is 

required. Training required: As determined by the Task Manager. 

Initiated by Jo Ann M. Jenkins.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 3 of 57

TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

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

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

6.2 Basis of Comparison...................................................................................................................9

6.3 Dose Reconstruction Parameters .............................................................................................12

6.3.1 Site Administrative Practices .........................................................................................12

6.3.2 Personnel Monitoring Systems Used at INEEL/ANL-W .................................................17

6.3.2.1 Initial Film Badge ............................................................................................17

6.3.2.2 Multiple-Filter NRTS Film Badge.....................................................................18

6.3.2.3 Original Lithium Fluoride Teflon TLD System ..................................................19

6.3.2.4 Automatic Thermo Luminescent Analyzer System ..........................................20

6.3.2.5 Harshaw Two-Chip TLD System .....................................................................20

6.3.2.6 Panasonic Four-Chip System..........................................................................20

6.3.2.7 Nuclear Track Emulsion, Type A for Neutrons ................................................21

6.3.2.8 Neutron Albedo Dosimetry ..............................................................................22

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

6.3.3.1 Beta-Gamma Radiation...................................................................................22

6.3.3.2 Neutron Calibration .........................................................................................24

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

6.3.4.1 Gamma Radiation ...........................................................................................25

6.3.4.2 Beta Radiation ................................................................................................26

6.3.4.3 Neutron Radiation...........................................................................................28

6.3.4.3.1 Materials Test Reactor (MTR), Zero Power Physics Reactor 

(ZPPR), and Transient Reactor Test (TREAT) Neutron 

Radiation.........................................................................................30

6.3.4.3.2 Typical Workplace Neutron Dosimeter Hp(10) Performance ...........34

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

6.4.1 Neutron Weighting Factor..............................................................................................35

6.5 Missed Dose .............................................................................................................................37

6.5.1 Photon Missed Dose .....................................................................................................37

6.5.2 Missed Beta Dose .........................................................................................................38

6.5.3 Neutron Unmonitored and Missed Dose........................................................................38

6.5.3.1 Before October 1976.......................................................................................39

6.5.3.2 After October 1976..........................................................................................39

6.6 Organ Dose ..............................................................................................................................40

6.7 Uncertainty................................................................................................................................40

6.8 Attributions and Annotations .....................................................................................................41
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References ........................................................................................................................................47

Glossary ............................................................................................................................................55
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LIST OF TABLES

TABLE TITLE PAGE

6-1 Area codes that could be in claimant dose files...................................................................... 15

6-2 Reasons codes ......................................................................................................................15

6-3 Irregularity codes ...................................................................................................................15

6-4 Column 20 codes ...................................................................................................................15

6-5 INEEL facility neutron correction factors from 1981................................................................ 22

6-6 Laboratory sources of uncertainty for beta/photon dosimeter calibration parameters ............. 23

6-7 Common sources of laboratory bias in the calibration parameters for neutron dosimeters ..... 24

6-8 Selection of IREP beta and photon energies for ANL-W facilities........................................... 26

6-9 Beta dosimeter thicknesses and associated under-reporting.................................................. 28

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

6-11 Facilities and time periods for neutron exposure .................................................................... 35

6-12 Dose equivalent fractions and quality factor corrections, estimated and recommended ......... 36

6-13 Recommended IREP neutron energy fractions and correction factors ................................... 36

6-14 Beta/photon dosimeter period of use, type, MRL, exchange frequency, and potential 

annual missed dose ...............................................................................................................37

6-15 Neutron dosimeter type, period of use, exchange frequency, laboratory minimum 

detectable limit, and maximum annual missed dose............................................................... 39

6-16 Spectrum correction for measured and missed neutron dose................................................. 39

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Individual dose reporting form in use until 1958 ..................................................................... 13

6-2 Film report form used in 1958 ................................................................................................ 14

6-3 Monthly badge reporting form from May 1959........................................................................ 14

6-4 Badge pull results from January 1961 for work in recovery from the SL-1 accident................ 16

6-5 Special badge report associated with a high beta reading listed in Figure 6-4........................ 16

6-6 Personnel exposure questionnaire partially completed for a hypothetical case ...................... 17

6-7 Response of DuPont Type 508 film with various filters to beta and photon 

irradiation ...............................................................................................................................19

6-8 Probability density of neutron spectrum from a 241AmBe (α,n) source .................................... 25

6-9 Gamma and beta radiation field characterization ................................................................... 26

6-10 Distribution of beta ranges ..................................................................................................... 27

6-11 Distribution of reportable neutron dose at INEEL for the first 9 months of 1995...................... 29

6-12 Neutron radiation field characterization .................................................................................. 29

6-13 Neutron spectra simulating INEEL facilities............................................................................ 30

6-14 Sample MTR spectra from Hankins Bonner measurements ................................................... 31

6-15 MTR neutron field components .............................................................................................. 32

6-16 Correlation of fast neutron dose equivalent to gamma dose at MTR ...................................... 32

6-17 ZPPR Assembly-3 Phase-1A Fast Neutron Spectrum............................................................ 33

6-18 ZPPR Assembly-3 Phase-1A Fast Neutron Dose Spectrum................................................... 34
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANSI American National Standards Institute

AP anterior-posterior

ARA Auxiliary Reactor Area

ATLAS Automatic Thermoluminescent Analyzer System

ANL Argonne National Laboratory

ANL-W Argonne National Laboratory - West

CFA Central Facilities Area

Ci curie

cm centimeter

DOE U.S. Department of Energy

DOE-ID DOE Idaho Operations Office

DOELAP DOE Laboratory Accreditation Program

EDF Engineering Design File

EG&G Edgerton, Germeshausen, and Greer Corporation

ERDA Energy Research and Development Administration

EBR Experimental Breeder Reactor

EEOICPA Energy Employees Occupational Illness Compensation Program Act

FNCF facility neutron correction factor

g gram

hr hour

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

ID (DOE) Idaho Operations Office

in inch

INEEL Idaho National Engineering and Environmental Laboratory

INEL Idaho National Engineering Laboratory

INL Idaho National Laboratory

IREP Interactive RadioEpidemiological Program 

kerma kinetic energy released to matter

keV kilovolt-electron, 1,000 electron volts

LET Linear Energy Transfer

LiF lithium fluoride

MeV megavolt-electron, 1 million electron volts

mg milligram

mm millimeter

mR milliroentgen

mrad millirad
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mrem millirem

mrep millirep

MRL Minimum Reporting Level

MTR Materials Test Reactor

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurement

NIOSH National Institute for Occupational Safety and Health

NRTS National Reactor Testing Station

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

PoBe polonium beryllium

POC probability of causation

R roentgen

RBE relative biological effectiveness

rem roentgen equivalent in man

rep roentgen equivalent physical

RESL Radiological and Environmental Services Laboratory

RWMC Radioactive Waste Management Complex

SL-1 Stationary Low-Power Reactor No. 1

TBD technical basis document

TLD thermoluminescent dosimeter

TRA Test Reactor Area

TREAT Transient Reactor Test

U.S.C. United States Code

wk week

yr year

Z atomic number

ZPPR Zero Power Physics Reactor

§ Section
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6.1 INTRODUCTION

6.1.1

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

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

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

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

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

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

Purpose

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

derived:

] guidelines established under 

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

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

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

 1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
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• Radiation from naturally occurring radon present in conventional structures

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

6.1.2

From the start of operations in 1951 until early 2005, Argonne National Laboratory – West (ANL-W) 

was operated by the University of Chicago under supervision of the Chicago Operations Office of the 

U.S. Atomic Energy Commission (AEC), the Energy Research and Development Agency (ERDA), and 

DOE. In 2005, the functions of ANL-W were combined with research activities at the former Idaho 

National Engineering and Environmental Laboratory (INEEL) to form the Idaho National Laboratory 

(INL), and ANL-W ceased to exist. A branch of the Idaho Operations Office (DOE-ID) provided 

external dosimetry resources and services at the INEEL, including ANL-W from the start of operations 

in 1951 [when it was called the National Reactor Testing Station (NRTS)] until 1989, when DOE 

transferred that responsibility to the prime operating contractor. Despite the fact that INEEL had 

several contractors at a time and that contractors changed often, the external dosimetry process has 

remained under technical management of a single organization with responsibilities for dosimetry 

development, operational dosimetry, and radiological records, which has provided a stable external 

dosimetry system. The former INEEL was originally called the National Reactor Testing Station 

(NRTS) and then the INEL (Idaho National Engineering Laboratory), but for simplicity, references will 

be to the INEEL. 

Scope

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 6.8.

6.2 BASIS OF COMPARISON

The Interactive RadioEpidemiological Program (IREP) calculates the probability of cancer induction in 

an organ from the external equivalent dose and internal dose received by that organ. Appendix B of 

the External Dose Reconstruction Implementation Guideline (NIOSH 2006) provides conversions from 

four photon dose quantities [deep dose equivalent, Hp(10); ambient dose equivalent, H*

(10);

exposure, X; and air kerma, Ka] and three neutron quantities [fluence, φ; ambient dose equivalent, 

H*

(10); and deep dose equivalent, Hp,slab(10)] to the organ doses. Over the years, as the National 

Council on Radiation Protection and Measurements (NCRP), International Commission on 

Radiological Protection (ICRP), and their predecessor organizations have developed the definitions of 

dosimetry parameters, dose parameters measured by the INEEL dosimetry system have received 

further definition. INEEL has reported doses as penetrating and nonpenetrating. The penetrating 

dose corresponds to the deep dose equivalent, and the nonpenetrating dose plus the penetrating 

dose corresponds to the shallow dose equivalent.

Horan and Braun (1993), Attix and Roesch (1968), and Meinhold (1975) discuss the history of 

radiation protection requirements from the 1930s. On September 29-30, 1949, representatives of the 

United Kingdom, Canada, and the United States met in Chalk River, Canada, at a Permissible Doses 

Conference (Warren et al. 1949) and picked the blood forming organs as critical for hard X-rays or 

gamma rays at a depth of 5 cm. Skin is the critical organ for soft X-rays or beta radiation. Dosage 

was to be reported in roentgen equivalent physical (rep) units equivalent to 93 ergs absorbed per gm

of tissue. The maximum permissible dose to the bone marrow was 0.3 rep/wk.

In 1949, the newly formed NCRP issued NCRP Report 7 as National Bureau of Standards Handbook 

42, which recommended a permissible dose of 0.3 R/wk (15 R/yr) for occupational workers (NBS 

1949, p 6). The term dose was undefined. A roentgen (R) was defined as the quantity or dose of Xrays such that the associated ionization per 0.001293 g of air (1 cm3 at standard temperature and 
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pressure) produces 1 electrostatic unit of charge of either sign. The unit roentgen only applies to Xrays or gamma rays, whereas the rep was used for all ionizing radiation and is the energy absorbed 

by tissue from the radiation (93 ergs/g, the same as for 1 R of gamma radiation).

A site manual in April 1952 stated the limit as “0.3 rep wk-1 at an effective depth in soft tissue of 5 cm, 

assumed to be the depth of the blood forming organs” (ACC 1952, p. IV: 1-1). It does not mention a 

quarterly or an annual limit. 

In 1953, the International Commission on Radiation Units and Measurements (ICRU) established a 

new unit, absorbed dose, which is the energy deposited in material per unit mass by radiation, using 

the rad (from radiation absorbed dose), a unit equal to 100 erg/g (ICRU 1954). ICRU specified the 

term exposure dose, later to become exposure, for the ionization capability in air for X- and gamma 

rays. In 1956, ICRU defined the term relative biological effectiveness (RBE) dose and the unit rem

(from roentgen equivalent in man), and introduced the concept of adding all types of external doses 

together (ICRU 1956). 

In 1957, the NCRP introduced an age prorating formula for the maximum allowable dose of 5 rem/yr

above age 18 [5 rem × (age [yr] -18)] (NBS 1958). This introduced 5 rem as an average annual dose,

but deemphasized that value as a limit. The AEC issued AEC Manual Chapter 0524, “Permissible 

Levels of Radiation Exposure,” on January 9, 1958, which adopted the prorating formula. It retained 

15 rem as the maximum annual dose and superseded the 13-week whole-body limit of 3 rem with “the 

provision that not more than one-fourth of the 15 rem maximum permissible yearly dose shall be 

taken in one-fourth of a year” (AEC 1958). 

The quarterly limit of 3 rem or 12 rem/yr replaced the 15 R/yr associated with the weekly limit (NBS 

1958, p. 3, footnote 2). President Eisenhower approved these values in 1960 for Federal agencies. 

AEC Manual Chapter 0524 was reissued in 1963 and 1968 (AEC 1963, 1968) and later still as ERDA 

Manual Chapter 0524 (ERDA 1975, 1977), which provided requirements for radiation safety.

In 1957, National Bureau of Standards (NBS, now the National Institute of Standards and Technology) 

Handbook 63 specified a dependence of the RBE on the linear energy transfer (LET) of the charged 

particles that actually deliver the dose (NBS 1957). NBS used this in the Snyder calculations of 

maximum permissible neutron flux (NBS 1961), and it is still used in the radiation control regulations 

for DOE.

At an April 1962 ICRU meeting, the use of the terms RBE and RBE dose in radiation protection was 

criticized, and the terms quality factor (QF, now Q) and dose equivalent (DE, now H) were introduced. 

The ICRU recommended the unit kerma (kinetic energy released in material) in 1962 (ICRU 1962).

In 1971, NCRP Report 39, Basic Radiation Protection Criteria, recommended an annual dose limit of 

5 rem, eliminating the quarterly limit (NCRP 1971a). In April 1975, ERDA reissued Manual Chapter 

0524, which invoked the 5-rem annual dose limits in NCRP Report 39 and required adding internal 

and external dose equivalents if both are known (ERDA 1975). Monitoring was required “where the 

potential exists for the individual to receive a dose or dose commitment ... in excess of 10% of the 

quarterly standard” of 3 rem. Personnel monitoring equipment for each individual was required for 

external radiation: “To achieve optimum accuracy, personnel dosimeters should comply with the 

performance parameters contained in American National Standards Institute (ANSI) standards N13.5

(ANSI 1972), N13.7 (HPS 1983a), and N13/42 WG1 Final draft 1974” (ERDA 1975, Appendix, p. 10). 

Quality factors from NCRP Report 38 are specified along with the neutron flux density for 100 mrem in 

40 hours as a function of neutron energy (NCRP 1971b). The NCRP Report 38 guidance for 

interpolating in energy cannot be accomplished with an instrument. The dose equivalent conversion 
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factors and the associated interpolation with energy reported in ICRP Publication 21 (ICRP 1973) do 

not present that problem.

In 1971, ICRU defined the quantity dose equivalent index, the maximum value in a 30-cm-diameter 

sphere, for describing ambient radiation fields for radiation protection purposes (ICRU 1971). ICRU 

extended this discussion in Conceptual Basis for the Determination of Dose Equivalent (ICRU 1976),

which defined the concept of deep and shallow dose equivalent indexes as those inside a 1-cm depth 

in the sphere and at a depth between 0.07 mm and 10 mm, respectively. A remaining issue was that 

the quantity was measured near the surface of the sphere, but applied to the center of the sphere, a 

distance of 14 or 15 cm. In 1980, ICRU identified the deep and shallow dose equivalent indexes as 

restricted indexes (ICRU 1980). In 1985, ICRU Report 39, Determination of Dose Equivalents 

Resulting from External Radiation Sources, introduced the concepts of aligned and expanded fields to 

eliminate issues of field direction and nonuniform fields; the document also introduced several dose 

equivalents: ambient dose equivalent, directional dose equivalent, individual dose equivalent 

penetrating, and individual dose equivalent superficial (ICRU 1985).

ICRP Publications 26 and 30 introduced new dose limits and the associated quantity effective dose 

equivalent as a weighted average over the radiation-sensitive organs of the body (ICRP 1977, 1979). 

In 1981, DOE Order 5480.1A, Chapter XI, “Requirements for Radiation Protection” (DOE 1981a),

superseded ERDA Manual Chapter 0524 (ERDA 1977). In 1988, DOE Order 5480.11, “Radiation 

Protection for Occupational Workers” (DOE 1988) superseded DOE Order 5480.1A, Chapter XI. This 

Order adopted much of the language of ICRP Publications 26 and 30 (ICRP 1977, 1979), and the 

monitoring threshold became 100 mrem effective dose equivalent. The order imposed slight changes 

in quality factor value for neutrons in one table, but did not capture those changes in the table of 

permitted neutron flux density.

Because of questions of quality control for dosimetry, the Conference of Radiation Control Program 

Directors encouraged development of a dosimetry accreditation process, leading to the development 

of ANSI N13.11 (HPS 1983b) and the National Voluntary Laboratory Accreditation Program (NVLAP). 

DOE Guidelines for the Calibration of Personnel Dosimeters (Roberson and Holbrook 1984) revised 

the ANSI (HPS 1983b) NVLAP processes. Calibration was to the quantities shallow and deep dose 

equivalent (Hs and Hd) and shallow absorbed dose (Ds), which are similar to the individual dose 

equivalent superficial and individual dose equivalent penetrating dose defined in ICRU (1985). These 

quantities were renamed to the personal dose equivalent Hp(d) (ICRU 1993) where d is the depth in

millimeters (0.07 mm for surface and 10 mm for deep) from the surface for which the dose is 

measured. In 1987, DOE Order 5480.15, “Department of Energy Laboratory Accreditation Program 

for Personnel Dosimetry,” established the DOE Laboratory Accreditation Program (DOELAP) system 

for dosimetry accreditation (DOE 1987). Standard for the Performance Testing of Personnel 

Dosimetry Systems specified the measurement of deep and shallow dose equivalents at depths of 

10 mm and 0.07 mm, respectively (DOE 1986a).

In 1990, the ICRP redefined the concept of dose equivalent to equivalent dose, redefined quality 

factor to radiation weighting factor, and generated new factors (ICRP 1991). These factors, invoked 

in NIOSH (2006), depend on neutron energy at the entrance to the body rather than on secondary 

particle LET where the dose is received. Dose conversion factors for organs and for ambient dose 

equivalent and personal dose equivalent were generated in ICRP Publication 74 (ICRP 1996) and are 

referenced in the external dose implementation guide (NIOSH 2006).
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Thus, the quantities to be measured and reported by the dosimetry systems have evolved over the 

last 50 years. Although the standards organizations were changing definitions, this had little impact 

on dosimetry measurements because, for gamma radiation, the numerical differences are small.

6.3 DOSE RECONSTRUCTION PARAMETERS

6.3.1

It was ANL-W policy that personnel expected to receive any radiation dose or personnel whose work 

was centered at the site were assigned a radiation monitoring badge. These badges were usually 

stored at the respective operational area entrance security gates for ANL-W facilities. Control

badges, which are used to subtract background radiation, have also been and are currently located 

where the badges are stored (Ruhter 2003). This practice could lead to subtracting environmental 

radiation from site activities, which would reduce the reported doses (ORAUT 2003). Environmental 

radiation levels have been monitored for most of the life of ANL-W, originally with film badges and 

later with thermoluminescent dosimeters (TLDs) (ORAUT 2004). Table 4-6 in the ANL-W 

environmental dose TBD (ORAUT 2004) lists results of this monitoring at facility fenceline locations 

near the security gates. A fraction (2000/8766) of these values can be added to an individual dose 

history or used for unmonitored workers working at the site.

Site Administrative Practices

Some individuals who could occasionally have visited site facilities but did little work with radiation had 

badges at several different facilities [1]. It is not appropriate to base missed doses on the multiple 

badges issued. Early on, the badge change frequency was not the same for everyone. Workers with 

low probability of exposure were placed on a longer change cycle than those with more chance of 

exposure (Cusimano 1972). Therefore, missed doses should be based on the actual change 

frequency for a person, which can be determined from the individual’s data package.

The INEEL dosimetry organization developed a set of basic administrative practices in 1951, which 

have changed somewhat as the technologies of ionizing radiation dosimetry and recordkeeping have 

changed (Cipperley 1958). 

DOE-provided dosimetry information for a former ANL-W worker will include copies of available 

dosimetry forms. These forms include a dose summary for each monitored employment period and a 

copy of each weekly, monthly, quarterly, etc., dosimeter result, which will also show the worker’s work 

location. The information can easily be several inches thick. Each sheet is redacted so only the 

person of interest’s name and applicable information are visible. This file provides the recorded

information on the exchange period for the person for that period. Figures 6-1 through 6-5 show a

partial example set of redacted dose reporting forms.

From 1951 to 1958, the INEEL dosimetry staff recorded each worker’s dose each day on a dose card 

(Figure 6-1), rezeroed pencil ionization chambers worn by workers, and entered the weekly badge

result on the same card (Cipperley 1958). On this sample, on October 28, November 16, and 

December 9, 1954, the badges were pulled and read in response to high pencil chamber readings. 

The personnel monitoring badges have always been considered more reliable than pencil dosimeters; 

so after the film badge results became available, the daily pencil readings were no longer considered 

doses of record. However, these values can be recovered from the earliest forms for a worst-case 

estimate of dose. In Figure 6-1, the pencil readings totaled 815 mR [255+280 (Oct 16)+280 (Nov 9)]

and the badges reported 0 mR for 18 badges.

Figure 6-2 is a report from reading the films in the same period. On three of the five badges, the more 

sensitive open-window result was zero, so the shielded film was not read. On the other two, the 
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open-window and shielded values were at the minimum recorded density of 0.02, which corresponded 

to a 30-mR penetrating dose.

Figure 6-1. Individual dose reporting form in use until 1958 (Vivian and Rockhold 2003, p 17).

After the pencil ionization chambers were replaced with self-reading pencil dosimeters (also ionization 

chambers), the ANL-W operational health physics staff would rezero the dosimeters (Horan 1959). 

The film reading was automated and results were stored in a computer. The form shown in Figure 6-3 

reports badge readings for May 1959 when badges were exchanged every 2 weeks. The column 

under the P of Personnel is an area designator with the code listed under Location at the bottom of 
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the page (Table 6-1 lists the codes for the areas). The column to the right was unused and dropped 

somewhat later. The next column to the right contained the reason codes listed in Table 6-2. 

Computer input card codes are listed in Tables 6-3 and 6-4.

Figure 6-2. Film report form used in 1958 (Vivian and Rockhold 2003, p 43).

Figure 6-3. Monthly badge reporting form from May 1959 (Vivian and Rockhold 2003, p 46). 

With the advent of computers, the reports are all computer-generated with the effect that even though 

many workers were not exposed to neutrons and did not receive neutron dosimeters, zeros are 

entered by the computer in the dose reports. A zero is not an indication that a dosimeter was 

assigned in a computer-generated report.
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Table 6-1. Area codes that could be in claimant dose files (Hill-Ellis 2003). 

Area code Description Area code Description

1 AEC Headquarters Bldg 20, 261, 264 TREAT

2 EBR – I 25 No information available

12 EX (X is construction) at EBR 26, 263,265 EBR -II

16 SL-1

For the early computer records (~1975-1984), the penetrating and nonpenetrating labels are reversed 

from the data that is printed. The nonpenetrating result is reported first, then the penetrating, and 

then the neutron result.

Table 6-2. Reasons codes (Column 68-69) (Hill-Ellis 2003).

Old Later Years Old Later Years

01 Regular Pull 11 Lost Pencil (or 

damaged)

Visitor HP 

Request

02 H.P. Request Misc. Pull 12 H.P. Check

03 High Dosimeter Reading Withdrawn 13 Late Pull

04 Recover Lost Badge Termination 14 Withdrawn Badge

05 Ring Reading 15 Termination

06 Wrist Badge Reading H.P. Request 16 Correction

07 Recovered Lost Badge & 

Withdrawn

17 Records Withdrawn

08 Late Pull for Not 

Available

18 Lost Film Reading

09 Miscellaneous Pull 19 X-Ray Exposure

10 Temporary Film Late Pull resolved 

by PEQ

20 Experiment Exposure

Table 6-3. Irregularity codes (columns 70-71) (Hill-Ellis 2003).

01 Defective Film 12 Dropped in Canal or Reactor

02 Impossible to Read 13

03 Light Leak 14 Not in Area

04 Water Soaked 15

05 16

06 O.W. Shot with X-Ray 17 Old Lot Film

07 Lost in Processing 18 Stuck Film

08 Heat Exposure 19 Not Available

09 Recovered Lost Badge 20 Lost Badge

10 Contaminated Badge 21 No Film

11 Wore Two Badges at one Time

Table 6-4. Column 20 codes (Hill-Ellis 2003).

“X” Master Card 6 Fast Neutron

1 Summary Card 7 Urinalysis

3 Sens. Beta-Gamma 8 Summary Card

4 Insen. Gamma 9 Summary Card

5 Slow Neutron 0 Total Body Results Card

Figure 6-4 is a listing of some doses received during recovery from the Stationary Low-Power Reactor 

No. 1 (SL-1) accident. Workers from several areas were pulled into the accident recovery process, 

and it is notable that one result exceeds the dose limits and that there are few zeros. Figure 6-5 is a 

follow-up badge report for one result on Figure 6-4.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 16 of 57

Figure 6-4. Badge pull results from January 1961 for work 

in recovery from the SL-1 accident (Vivian and Rockhold 

2003, p 50).

Figure 6-5. Special badge report associated with a high beta reading 

listed in Figure 6-4 (Vivian and Rockhold 2003, p 57).

When there has been a question about a dose value being assigned to an ANL-W worker, a 

Personnel Exposure Questionnaire (PEQ) was normally initiated (Figure 6-6 shows a hypothetical 

case) (Cipperley 1958). Based on this form, a beta dose of 500 mrem and a gamma dose of 350 

mrem for a total dose of 850 mrem would override the pocket meter dose of 290 mrem total.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 17 of 57

Figure 6-6. Personnel exposure questionnaire partially completed for a hypothetical 

case (Cipperley 1958). 

6.3.2

6.3.2.1 Initial Film Badge

Personnel Monitoring Systems Used at INEEL/ANL-W

The badging system in place when operations began at the NRTS was the Self-Service System 

(Cipperley 1958). This film system, in use from August 1951 to March 1958, used the Oak Ridge 

National Laboratory stainless-steel holder design, which was 1.875-in. long, 1.375-in. wide, and 0.25-

in. thick. Badges were processed weekly. The upper portion of the badge was shielded with 1 mm of 

cadmium and the lower portion was an open window. Sensitive and insensitive DuPont Type 552 film 
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 18 of 57

was used for beta-gamma dosimetry for most locations; DuPont Type 558 film (Type 508 sensitive 

and Type 1290 insensitive) was used at two reactor areas (Cipperley 1958). 

Gamma calibration was to a radium source, and beta calibration was to a metallic uranium plate. To 

determine doses, the film densities were read to ±0.02 density unit (Cipperley 1958). A calibration 

curve was used to convert the cadmium-shielded portion to penetrating gamma exposure in roentgen. 

The open-window density corresponding to the gamma exposure was subtracted from the measured 

open-window density and the remainder was converted to beta dose in rep (Cipperley 1958).

Type 552 and 508 films have Minimum Reporting Levels (MRLs) of approximately 30 and 10 mR, 

respectively, to radium gamma radiation (Cipperley 1958). The open window responds to beta 

radiation as well as X-rays and low-energy gamma rays. Because of the high atomic number (Z) of 

film in relation to air or tissue, the open window over-responds per unit exposure to low-energy photon 

radiation, as shown in Figure 6-7, by about a factor of 30 at 40 keV (Cipperley and Gammill 1959). 

Using a cadmium filter with its high Z severely attenuates the photons that get to the film, so the overresponse is reduced to about a factor of 2 at 125 keV and becomes less than 1 at about 50 keV. The 

beta particle range is independent of Z; the range depends only on the density (BRH 1970, p 29), so 

the 1-mm cadmium filter (~900 mg/cm2

) acts like a tissue depth of 9 mm for beta radiation.

Wrist badges used the same package attached to a wrist band. A finger ring used a small piece of 

film with a silver or cadmium filter (Cipperley 1958). Pencil ionization chambers were used to monitor 

daily doses and control operational activities. The dosimetry group read and recorded these pencil 

readings on cards. Film badge readings were written on the same cards to indicate the dose of 

record (Cipperley 1958). In 1958, the Victoreen 352 pencil ionization chambers being read by the 

dosimetry group were replaced with self-reading dosimeters that were read and rezeroed by the field 

health physics technicians (Horan 1959, p. 11). Film readings remained the dose of record.

6.3.2.2 Multiple-Filter NRTS Film Badge

In March 1958, the security badge and film badge were combined in a film badge containing filters of 

1 mm cadmium, 0.013 mm silver, and 0.5 mm aluminum with thicknesses of 950 mg/cm2

, 

203 mg/cm2

, and 175 mg/cm2, respectively, including the plastic in which they were mounted 

(Cipperley 1968). This NRTS badge was also a security badge, resulting in an absorber thickness of 

100 mg/cm2 in the open window, which filtered out beta radiation below 360 keV.

With the four absorbers, it was possible to separate beta radiation from low-energy photon radiation 

and to determine photon energy to a degree. Figure 6-7 shows photon energy dependence of the 

darkening behind the four filters for a combination of uranium beta and X-ray irradiation provided by 

NBS (Cipperley and Gammill 1959). With DuPont Type 508 film, mixed exposures of radium gamma 

and uranium beta of 10, 20, and 30 mR or mrep were measurable within ±12 mR with 95% 

confidence. A minimum reporting level of 10 mrem was used for both beta and gamma radiation 

(Horan 1963).

The DOE Idaho Operations Office (IDO, later ID) Instrument and Development Branch developed an 

automatic film reader and densitometer (Purcell and McGary 1963). An algorithm based on probitcorrected densities was developed to determine the high-energy photon, beta, and low-energy photon 

contributions separately (Cipperley 1968, p. 94). The cadmium filter provided the hard gamma 

component. The uranium beta responses under the open window, aluminum, and silver filters were 1, 

0.2, and 0.1, respectively. By assuming a beta signal and subtracting it, the remaining signal could be 

attributed to low-energy photons and the energy could be estimated. For beta sources other than 

uranium, the analysis had greater uncertainty.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 19 of 57

Figure 6-7. Response of DuPont Type 508 film with 

various filters to 140 mrem uranium beta and 100 mR of 

different energy photon irradiation provided by NBS. The 

original badge used the open window and cadmiumshielded films. The multiple filter badge used all three 

filters plus the open window (Cipperley and Gammill 

1959).

Because about 95% of the weekly badge films had doses less than 30 mrem, in 1958, the badging 

interval was increased to biweekly or monthly with the exception of the high-dose areas where the 

weekly schedule continued (Horan 1959). The introduction of punch cards increased the efficiency of 

report and record generation. A 12-point calibration curve was generated for radium and for 137Cs 

gamma and uranium beta. Calibration did not use a phantom.

Experience following the Stationary Low-Power Reactor No. 1 (SL-1) accident showed a wide 

variation of beta-to-gamma ratios and necessitated controlling both radiations rather than just the 

gamma. A set of as many as 18 badges could, and in many cases was, fastened on a belt around the 

worker to determine a beta-to-gamma ratio for each particular entry into SL-1 during recovery 

operations (Cipperley, Henry, and Cusimano 1965). 

6.3.2.3 Original Lithium Fluoride Teflon TLD System

Beginning in November and December 1966, individuals projected to receive doses of less than 

0.5 rem/yr were placed on a lithium fluoride (LiF) disk TLD badge, which was exchanged quarterly 

(Cusimano and Cipperley 1968). Two 13-mm-diameter Teflon disks, 0.4 mm thick [100 mg 

(75 mg/cm2

) impregnated with 28 mg LiF], were mounted in a badge behind an open window and a 1-
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 20 of 57

mm cadmium filter (Watkins, date unknown). The disks, manufactured by Teledyne Isotopes, were 

read with the Teledyne Model 7300 TLD reader. LiF was chosen because the average Z is close to 

that of air and tissue, resulting in little energy correction for beta or gamma radiation. The badge 

could read 30 mR on a quarterly basis (Watkins, date unknown), so more small doses were reported

(Cusimano and Cipperley 1968). The angular dependence of the gamma response (within 10% to 

70o) is superior to film because the material acts like an ionization chamber. For normal monitoring, 

only the open-window TLD was read and considered penetrating dose unless it read more than 125 

mrem, in which case the shielded TLD was also read (Watkins, date unknown). 

The pilot tests were successful, and the LiF Teflon TLD system was phased into use in 1966, 

particularly for individuals who would receive low doses, with longer exchange cycles, typically 3 or 

6 months (Vallario, Hankins, and Unruh 1969). In July 1968, the monitoring period was increased 

from 3 to 6 months (Voelz 1969). In December 1972, annual processing was used for 1,190 low-dose 

individual TLDs and 960 were processed quarterly (Cusimano 1972). Employees on a monthly badge 

change were moved to this system as late as September 1973.

The system had an automatic badge calibrator that did not involve a phantom to provide backscatter 

(Cipperley 1966; Voelz 1970, p. 8). A later discussion introduced the use of a 137Cs source, so these 

earlier calibrations probably used radium, which was used for testing (Cusimano and Cipperley 1968).

6.3.2.4 Automatic Thermo Luminescent Analyzer System 

Development began in 1969 on a patented Automatic Thermoluminescent Analyzer System (ATLAS). 

It used LiF in a homogeneous mixture with Teflon and replaced the film in the multielement badge 

using the same filters. ATLAS became operational for monthly badge changes in February 1974. In 

June 1974, questions about this system were formalized (Black 1974; Walker 1974) and the system 

was soon replaced. A minimum reporting level (MRL) of 30 mR was used for gamma and beta 

(Walker 1974).

6.3.2.5 Harshaw Two-Chip TLD System

Several unstable characteristics with ATLAS led to rapid implementation of a two-chip TLD system in 

May 1975 for ANL. This commercial Harshaw system used two LiF TLDs 240 mg/cm2 thick. In 1976, 

holes were punched in the security badges to restore the open window. One chip was covered by 

540 mg/cm2 (initially 350 mg/cm2

) of aluminum and the other was under 4 mg/cm2 of Mylar. The 

aluminum-covered chip provided penetrating dose at a nominal tissue depth of 1 cm. The beta dose 

was calculated from the difference between the two chips. Because of the thickness of the 

Mylar-covered chip, the beta dose was accurate only for the beta energy used in calibration. Field 

calibrations were used to reduce the problem with beta energy dependence. The thin aluminum filter 

(density thickness 350 mg/cm2

) allowed higher energy beta radiation to expose the chip used for 

measuring the penetrating (1,000 mg/cm2

) dose, so it was changed to 2 mm in July 1977 (ERDA ca. 

1978). The practice was to read only the open-window chip to determine if the nonpenetrating dose 

was above 15 mrem and thus reportable. If the threshold dose was exceeded, both chips were read 

and the penetrating and nonpenetrating doses were computed (Kalbeitzer 1983).

6.3.2.6 Panasonic Four-Chip System

In 1986, with the advent of DOELAP, INEEL went to a four-element system, the Panasonic 814 AS4 

(Gesell, Hall, and Andersen 1992; DOE 2001). Lithium borate (Li2B4O7) TLD elements with plastic 

and aluminum filtration provide an improved measurement of deep dose equivalent and, with a thinner 

filter, an improved measurement of the shallow dose equivalent. A calcium sulfate (CaSO4) TLD 
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 21 of 57

provides a strong low-energy photon response. The elements are 15 mg/cm2 thick. Element one has 

filtration of 16 mg/cm2

, element two has filtration of 58 mg/cm2 plastic, and elements three (CaSO4) 

and four have filtration of 550 mg/cm2 of plastic and 50 mg/cm2 of aluminum. Although none of the 

elements are at a depth of 7 mg/cm2

, the specified depth for the shallow dose equivalent, an 

acceptable response can be obtained by using elements at 16 and 58 mg/cm2

. This system is 

qualified in the DOELAP beta, photon, and mixture performance categories. The angular dependence 

of this system has been measured for 137Cs gamma rays and four X-ray energies from 16 to 70 keV 

(INEEL 2001). For the lower energies, attenuation in the absorbers reduces the measured dose. 

(The attenuation also reduces the delivered deep dose, but this effect is not incorporated in the 

results.)

The minimum reporting level was 15 mrem beta and gamma from January to July 1986 (Gesell 1986), 

10 mrem gamma and 30 mrem beta from July 1986 to about September 1989, and 15 mrem for 

gamma and 30 mrem for beta until 1993 (Perry, Andersen, and Ruhter 1993), when it returned to 

10 mrem gamma.

6.3.2.7 Nuclear Track Emulsion, Type A for Neutrons

Kodak nuclear track emulsion, Type A (NTA) was used for neutron monitoring when the field radiation 

protection staff requested it. NTA responds to neutrons with energies above 500 to 800 keV, for 

which the proton recoil tracks leave enough energy to expose at least four (three in some references) 

grains of emulsion (Cusimano 1963, Cipperley 1968).

The minimum dose assigned was 14 mrem (Cipperley 1958, p 6). Before 1958, if a proton recoil track 

was counted in 40 microscope fields, it was read twice more for a total of 120 fields (Cipperley 1958, p 

17). On one data sheet from March 1958 with 10 neutron readings, three persons received 14 mrem 

and one received 42 mrem. Two of the four had gamma readings (the MRL was 30 mrem). A blank 

is recorded for 17 people on the data sheet, probably because they were not monitored for neutrons 

or the film was not read. Only the two individuals received measurable gamma doses. A person on 

weekly exchange from January to March 1958 received neutron dose equivalents of 14, 28, 42, and 

14 mrem and gamma doses of 130, 70, 30, 30, 50, 30, and 20 mrem with both neutron and gamma 

doses received twice (Vivian and Rockhold 2004, p 5). These values indicate that the data sheets 

support the MRL of 14 mrem.

After 1959, if more than three proton recoil tracks were counted in 40 microscope fields, a different 

location on the film was counted by two more technicians, providing three independent results

(Cipperley 1968). Two tracks or fewer were attributed to background. This resulted in a somewhat 

higher MRL. In November 1959, a data sheet shows 10- and 20-mrem neutron dose equivalents

(Vivian and Rockhold 2004, p 15). In January 1962, a data sheet shows a 20-mrem dose (Vivian and 

Rockhold 2004, p 6). A data sheet from April 1959 shows neutron dose equivalents of 20, 20, and 40 

mrem (Vivian and Rockhold 2004, p 3). These values suggest an MRL of 20 mrem.

Calibration was with a polonium-beryllium (PoBe) source (approximately 30 Ci), resulting in 

5.87 × 10-4 tracks/neutron (Cusimano 1963). Uncertainties were assigned at the 90% confidence 

level. Cipperley (1968, pp. 102-115) discusses this process.

The field health physics staff was aware of the energy limitations of the NTA badge (Sommers 1967, 

1969) and compensated with neutron-detecting pencil dosimeters and field measurements. A request 

to read NTA film occurred if the hard spectra neutron exposure was likely to exceed 10 mrem. 
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 22 of 57

6.3.2.8 Neutron Albedo Dosimetry

Because of the missed dose from neutrons below the NTA energy threshold of 0.5 to 0.8 MeV, 

particularly at plutonium facilities, and because of the advent of TLD techniques, several development 

efforts in neutron dosimetry occurred in the early 1970s. The result was several designs using the 

albedo technique in which scattered neutrons from the wearer’s body were absorbed by 6

Li in a TLD

(Gesell et al. 1996). 

In the Hankins dosimeter used at the INEEL and ANL-W (Hankins 1973), TLDs (6

Li to capture thermal 

neutrons) are inside a polyethylene case (to lower the neutron energy) inside a cadmium shell (to 

eliminate thermal neutrons from outside). Lithium-7 TLDs are used to subtract the gamma dose

(Gesell et al. 1996). Because the 6

Li(n,α)3He reaction has a strong energy dependence, the response 

does not follow the flux-to-dose-equivalent conversion, so the neutron signal is divided by a facility 

neutron correction factor (FNCF) (Gesell et al. 1996). An FNCF that converts the TLD gamma 

equivalent signal to neutron dose equivalent can be generated from the ratio of the dose equivalent 

measured with a 9-in.-diameter Eberline PNR-4 and the corresponding signal (reads in mrem, but not 

dose equivalent) with the detector in the 3-in.-diameter PNR-4 insert. A plot of FNCF versus 9” to 3” 

ratio is used to determine the FNCF from the measured ratio (Hankins 1976). The FNCF values listed

in Table 6-5 were measured for different fields at INEEL, were tabulated for assigning the dose 

equivalent from the badge results, and were routinely updated (Cusimano 1981). This correction was 

applied to generate the reported neutron dose. An MRL of 15 mrem was used (Gesell et al. 1996). 

The angular dependence of this system has been measured for moderated 252Cf neutrons (INEEL 

2001).

Table 6-5. INEEL facility neutron correction factors from 1981

(Cusimano 1981).

Organization FNCF Organization FNCF

DOE-CFA 0.092 EG&G-TRA (bare PuBe) 0.06

EG&G-CFA 0.092 EG&G-TRA (PuBe in poly) 0.23

ANL-TREAT 1.05 EG&G-LOFT 3.5

ANL-ZPPR 0.94 EG&G-ARA III 2.0

EG&G-TRA (L & S) 1.2 EG&G-RWMC 0.33

EG&G-TRA (SA) 2.7

The date of changeover from NTA to albedo neutron monitoring is in dispute. Typically, different 

organizations would change to new monitoring systems at different times. The present record 

suggests it occurred in late 1974 or early 1975 (Ruhter and Perry 2002; Gesell et al. 1996), but an 

informal list of “Dosimetry Branch Changes” from 1978 (ERDA ca. 1978) states “initial testing of 

albedo neutron dosimeter and replacement of NTA neutron monitoring film with same” in October 

1976. Aoki (1979) stated the Albedo system replaced the NTA badge in 1977. Dose reconstructions 

should make the assumption that this transition occurred in October 1976.

6.3.3

6.3.3.1 Beta-Gamma Radiation

Calibration

Table 6-6 lists common sources of laboratory bias for personnel beta/photon dosimeter calibration 

based on comparison of the recorded dose with Hp(10).

Gamma calibration initially used a radium source. Victoreen R meters standardized by NBS were 

used to measure radiation fields (Horan 1959, p. 132). Uranium metal bars 5 mm thick were used for 
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 23 of 57

beta calibrations. Cesium-137 was considered for a calibration source in 1959 (AEC 1960, p. 83) and 

was installed in the instrument calibration facility in 1961 (Horan 1962). An automatic badge irradiator 

developed in the 1960s did not use a phantom to provide backscatter (Cipperley 1966).

Table 6-6. Laboratory sources of uncertainty for beta/photon dosimeter calibration parameters [2].

Parameter Historical description Uncertaintya Comment

In-air 

calibration

In 1981, INEEL began exposing 

calibration dosimeters on 

phantoms (used to simulate 

worker body). Previous 

calibrations do not include 

response from radiation 

backscatter response.

+10% Recorded dose of record too high. 

Backscatter radiation from worker body is 

highly dependent on dosimeter design. 

Radiation 

quantity

Before 1981, INEEL dosimeter 

systems were typically calibrated 

to photon beam measured as 

exposure.

-5% For higher energy 226Ra and 137Cs gamma 

radiation used to calibrate dosimeters, this 

caused slight (about 3%) under-response

in recorded dose. 

Tissue depth 

of dose

Historically, INEEL used 

unspecified depth to estimate 

deep dose.

±5% Numerical effect of this for photon radiation 

is comparatively low. INEEL dosimeter 

designs had filtration density thickness of 

about 1,000 mg cm-2 that would relate 

closely to 1-cm depth in tissue.

Angular 

response

INEEL dosimeter system is 

calibrated using anteriorposterior (AP) laboratory 

irradiations.

>300 keV, 

~20%

Recorded dose of record likely too low 

because dosimeter response is usually 

lower at non-AP angles. Effect is highly 

dependent on dosimeter type, radiation 

type, energy, and angle. 

Environmental 

stability

INEEL film dosimeter and TLD 

systems are subject to signal 

fade with time, heat, humidity, 

light, etc.

±10% Recorded dose of record depends strongly 

on dosimetry parameters such as when 

calibration dosimeters were irradiated and 

processed. Midcycle calibration minimizes 

effects. 

a. Uncertainty estimate in recorded dose compared to Hp(10) based on judgment.

As reported in 1981, an extrapolation chamber was built for the measurement of beta doses (Gupta 

1981). The chamber window was polycarbonate, the gas was air, and the thick collecting electrode 

was Shonka tissue-equivalent plastic. The chamber was used to calibrate a 2.5 Ci 90Sr/Y source to 

tissue rad. The source, with area 2.5 cm2

, was constructed by the Amersham Searle Corporation in 

February 1975. This source was used to measure beta correction factors for several instruments 

following the Three Mile Island TMI-2 reactor accident in 1978. TLD badges were calibrated to 

500 mrad tissue using a 1.78-cm-thick phantom 50 cm (300 rad/hr) from the source (Gupta 1981).

In January 1983, the natural uranium slab again became the primary calibration source for 

nonpenetrating radiation to better approximate field beta spectra (Gesell 1982a).

Use of a phantom in calibration apparently started about 1981, with the NVLAP certification process 

developed for non-DOE dosimetry processors. About this time, calibration techniques developed in 

terms of absorbed dose to tissue rather than exposure. Beginning in January 1981, in response to a 

draft NVLAP, standard dosimeters for calibration were irradiated with 137Cs using a phantom backing. 

To convert from exposure in roentgen to dose equivalent index in rem, the Radiological and 

Environmental Services Laboratory (RESL) used a conversion factor (Cx) of 1.08 (DOE 1981b). The 

current recommended Cx value of 1.03 for 137Cs (DOE 1986b, Table 2) was used beginning in June 

1981 (Gesell 1982b; Kalbeitzer 1984).
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 24 of 57

In 1989, the INEEL dosimetry service transferred from RESL to Edgerton, Germeshausen, and Greer 

Corporation (EG&G), Inc., the prime contractor. Calibrations continued to use DOE RESL sources,

and no changes were made to the dosimetry system. The 1991 Tiger Team Review of the INEEL site 

indicated that the joint use by the INEEL contractor and DOE-ID of the same sources for calibration 

led to a conflict of interest or an advantage in DOELAP tests (DOE 1991, p 4-178). As a result, 

EG&G purchased a Shepherd panoramic irradiator with a 137Cs source for badge irradiations. This 

irradiator does not use a phantom, but was cross-referenced using many TLD irradiations to the DOE 

source using a phantom (Andersen 1994). In addition, the contractor developed and characterized a 

uranium slab for beta irradiations (Bean 1995).

6.3.3.2 Neutron Calibration

Table 6-7 lists several common sources of expected laboratory bias for personnel neutron dosimeters 

based on comparison of the recorded dose with Hp(10).

Table 6-7. Common sources of laboratory bias in the calibration parameters for neutron dosimeters

[3].

a

Parameter Historical description Anticipated laboratory biasb

Source energy 

spectrum

In 1976, INEEL began using dosimeters 

calibrated on phantom to simulate worker’s 

body. Previous calibrations did not include 

response from backscattered radiation. 

NTA film tends to be insensitive to albedo 

neutrons, so probably had minimal effect.

Radiation 

quantity

Neutron dose quantities used to calibrate 

INEEL neutron dosimeters have varied 

historically. First collision dose for fast 

neutrons and quality factor of 10 were used 

for many years. 

As noted above, NTA calibration would 

result in reported dose being about 11% 

high. Effects of respective neutron dose 

quantities used to calibrate INEEL 

dosimeters is uncertain and could be 

evaluated in comparison to Hp(10) dose 

used in DOELAP performance testing. 

Angular 

response

INEEL dosimeters calibrated using AP 

laboratory irradiation.

Recorded dose of record is likely too low 

because dosimeter response is lower at 

non-AP angles. Effect is highly dependent 

on neutron energy.

Environmental 

stability

INEEL NTA film and TLD dosimeters are 

subject to signal fade with time, heat, 

humidity, light, etc.

Recorded dose of record is likely too low; 

however, this depends strongly on when 

calibration dosimeters were irradiated 

during dosimeter exchange cycle. 

Midcycle calibration minimizes the effects. 

a. Judgment based on INEEL dosimeter response characteristics.

b. Recorded dose compared to Hp(10).

The initial NTA neutron badges were calibrated using a PoBe neutron source (30 Ci in 1958) (Horan

1959). In 1982, an AmBe source was used (Cusimano 1982). Alpha particles from the americium or 

polonium interact in the 9

Be(α,n)12C reaction and generate a broad spectrum of neutrons up to about 

11 MeV (mean energy about 5 MeV) as shown in Figure 6-8 (Kluge and Weise 1982). The yield of 

the AmBe source should be only about 3% larger than that for the PoBe source (Anderson 1971). 

Kluge and Weise (1982) calculate conversion factors of 3.51 to 3.76 × 10-8 rem cm2

/n depending on 

the particular measure of dose equivalent chosen. IAEA (1988) provides a dose conversion factor for 

AmBe of 3.8 × 10-8 rem cm2

/n for the maximum average dose equivalent. A dose equivalent of

1.5 rem required 3.6 × 107 n/cm2 (Cusimano 1963), corresponding to a dose conversion factor of 

4.17 × 10-8 rem cm2/

n, so the recorded dose will be about 11% high. Monte Carlo calculations for 5-

MeV neutrons show a dose equivalent of about 4.2 × 10-8 rem cm2

/n averaged over the 0- to 2-cm 

shell on a 30-cm-diameter cylindrical phantom (NCRP 1971b). 
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 25 of 57

In 1993, a 40- by 40- by 15-cm polymethyl methacrolate phantom was placed near the unmoderated 

252Cf source used for instrument calibration, and the system was characterized for TLD calibration 

(Gesell et al. 1996, Appendix A). This system has since been used for neutron dosimeter quality 

assurance measurements. Calibration factors from the DOELAP manual are used (DOE 1986b).

Figure 6-8. Probability density of neutron spectrum from a

241AmBe (α,n) source (Kluge and Weise 1982).

6.3.4

6.3.4.1 Gamma Radiation

Workplace Radiation Fields

In response to a Tiger Team finding (DOE 1991), radiation fields at the INEEL have been 

characterized by making field measurements with a NaI(Tl) gamma spectrometer and TLDs mounted 

on a phantom, then comparing the results (Reilly 1998). Figure 6-9 shows the percentage bias for the 

beta and gamma measurements. Most results lie between +27% and -43%.

The high gamma bias results were for locations at the Radioactive Waste Management Center 

(RWMC) looking at skyshine (backscattered and thus, low-energy photons) from low-level waste in 

the Subsurface Disposal Area (Reilly 1998). The doses measured with NaI(Tl) were low (6 and 11 

mrem) and the threshold energy on the NaI(Tl) detector was about 100 keV, so some low-energy 

photons are likely to have been missed (Reilly 1998). 

The radiation fields at ANL-W, with a few exceptions, have been generated primarily by mixed fission 

and activation products. Therefore, most of the photon doses have come from photons with energy 

greater than 250 keV [4]. The INEEL dosimeters are judged to measure these fields well.

The few exceptions have usually been characterized by low dose rates. Analytical X-ray generators 

operating below 100 keV have been used in several laboratories. These are easily shielded so the 

fields have usually been low [5].

A 320-keV and a 160-keV X-ray generator were used for radiography or radiography development 

studies. Wall shielding has generally been adequate and any transmitted photons would have had 

energy near the operating voltage because of the hardening caused by the shielding [6].
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 26 of 57

 

0

1

2

3

4

5

6

7

-85 to -75%-75 to -65%-65 to -55%-55 to -45%

-45 to-35%

-35 to -25%-25 to -15%

-15 to -5%

-5 to 5%

5 to 15%

1'5 to 25%

25 to 35%35 to 45%45 to 55%55 to 65%65 to 75%75 to 85%85 to 95%

95% to 105%

105% to 115%115% to 125%125% to 135%

Relative Bias [(Dosimeter-Spectrometer)/Spectrometer]

Number

Gamma

Beta

Figure 6-9. Gamma and beta radiation field characterization (Reilly 1998).

Essentially all ANL-W radiological work areas involved beta/photon radiation covering a wide range of 

energies. These fields can be generally classified according to the IREP codes in Table 6-8.

Table 6-8. Selection of IREP beta and photon energies for ANL-W facilities.

Process/

buildings Description

Operations

Radiation 

type

Energy 

group 

Begin End (keV) Percentage

Reactors Highly dispersed fields of higher energy photon radiation fields 

from fission process, activation, and fission product nuclides. 

Potential for significant airborne nuclides, and there could be 

significant higher energy beta radiation.

Beta >15 100

Photon 30–250 25

>250 75

EBR, ANL

1952 2003

Processing 

plants 

Highly dispersed fields of higher energy photon radiation fields 

from activation and fission product nuclides dominant to most 

exposure profiles. Potential for higher energy beta radiation 

during sampling and maintenance work resulting from fission 

products. 

Beta >15 100

Photon 30–250 25

>250 75

ICPP, ANL 1952 2003

Calibrations Calibration of instruments and dosimeters Beta >15 100

1952 2003 Photon 30 – 250 25

>250 75

6.3.4.2 Beta Radiation

Beta radiation fields are usually associated with activation or fission product radioactivity that is lightly 

shielded, handled by workers such as in hot cells, or outside a container such as a spill [7]. The high 

bias results in Figure 6-9 from comparison of TLD to a phoswich beta spectrometer are for beta 

sources at contact or at 1 cm, which results in a geometry that is difficult to reproduce. The low-bias 

results are for large area beta sources for which even the spectrometry results have large variations. 

The beta occupational radiation fields in Figure 6-9 (only 3) have a bias less than 15%.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 27 of 57

Beta field dosimetry became fairly accurate with the definition of DOELAP requirements in the early 

1980s. Before then, beta monitoring systems had various flaws, primarily in a detector too thick to 

give a good surface result or one that was covered with extra material. Calibration was to high-energy 

betas from either uranium or strontium (see § 6.3.3.1). The dose from low-energy betas can be 

missed altogether if the beta energy is not sufficient to penetrate the detector cover, and the dose can

be under-reported if the beta energy is not sufficient to penetrate the entire detector. The mean beta 

energy for the spectrum from a particular nuclide is about one-third of the maximum beta energy for 

that nuclide.

Based on the range energy curve for beta particles and the beta energy distribution of beta emitters 

(HEW 1970, pp. 90, 91, and 123), the fraction of beta radionuclides with ranges greater than the 

abscissa is plotted on the ordinate in Figure 6-10. The beta nuclides varied by location and time, so a 

correction factor common for all facilities was estimated [8]. This analysis used the entire mixture of 

radionuclides to avoid questions of whether the choice is correct and to reflect the wide variety of 

radionuclides used. To reflect that the beta spectrum is not monoenergetic because of the energy 

carried off by the neutrino, a curve is presented for the mean energy (one-third of the maximum 

energy). To reflect that some beta particles enter the detector at an angle, a curve is provided for 45o 

incidence at the maximum energy and the mean energy. These curves of the fraction of nuclides with 

a larger range essentially show the depth dependence of beta dose because the energy loss of 

electrons does not have much energy dependence. These curves also demonstrate why early 

dosimeters with thick, sensitive elements failed to correctly report the beta dose at a depth of 

7 mg/cm2

, which was chosen in the early 1980s. In addition, these curves demonstrate why the beta 

dose assigned for skin is inappropriate to use for the breast and testes, in which much of the organ is 

at a depth greater than 1 mm or 100 mg/cm2

, and for most of persons at depths greater than 1 cm.

 

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 50 100 150 200 250 300

Range (mg/cm^2)

Fraction of Nuclides

Max energy

Max 45 Incid

Mean Energy

Mean 45 Incid

Figure 6-10. Distribution of beta ranges [9].
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 28 of 57

To calculate the fraction of dose missed by a dosimeter, compare the average of the appropriate 

curve of this nature over the depth of the active detector to the value at a depth of 7 mg/cm2

. The 

appropriate curve should be the curve of the mean range for the beta spectrum and the angular 

distribution of the radiation exposure. To estimate this value, this analysis added the mean energy 

curve for perpendicular incidence and 1.4 (relative path length) times that for 45o incidence for the 

mean energy and added one-half that value for 45o incidence for the maximum energy [10]. The 

curves are the result of a polynomial trend line to the data, so averaging the fraction of radionuclides 

is relatively easy.

Table 6-9 provides the cover and detector thicknesses for the INEEL beta badges. To determine the 

corrected beta dose, the measured and missed non-penetrating result should be multiplied by the 

values in the last column of Table 6-9. The reported dose will likely be somewhat higher than this 

because the calibration probably did not consider such a correction and reported the dose for the 

calibration exposure [11]. For the Panasonic system, such a correction has already been made 

(Gesell 1986), so the recommended correction is 1.

Table 6-9. Beta dosimeter thicknesses and associated under-reporting [12].

Dosimeter 

system Period

Covers

(mg/cm2)

Detector thickness

(mg/cm2)

Beta correction 

factor

Two-filter film 1951–1958 50a 50a 2.0

Multifilter film 1958–1974 100 20a 2.8

Low-dose TLD 1969–1974 100 75 3.0

ATLAS 1974–1975 100 100a 3.3

Harshaw TLD 1975–1976 104 344 4.8

Harshaw TLD 1976–1985 4 240 2.4

Panasonic TLD 1986–2004 16 15 1.0

a. Estimated.

6.3.4.3 Neutron Radiation

Most ANL-W workers have not been exposed to neutrons and so, have not been badged to measure 

neutrons. Neutron fields are specific to a few facilities. 

In 1969, 150 out of 2,900 film-badged and more than 3,000 TLD-badged employees at the INEEL,

including ANL-W, were involved in radiation work that required their NTA neutron dosimeters to be 

evaluated (Vallario, Hankins, and Unruh 1969).

For calendar year 1979, five people for all of the INEEL received neutron doses between 0.5 and 1 

rem, and 79 received measurable neutron doses below 0.5 rem (Jones 1980). 

Individuals who have the potential to receive neutron dose currently wear albedo badges, and 

experience has shown that most do not receive significant doses. In the first 9 months of 1995 for all 

of the INEEL, only 1,461 neutron dosimeters were issued (both monthly and quarterly badges) 

compared to about 50,000 beta/gamma badges. Only 54 badges had reportable doses (≥15 mrem) 

as shown in Figure 6-11 (Gesell et al. 1996). Only six were above 35 mrem. The Hankins albedo 

dosimeter badges in use since 1975 see all neutron fields. An FNCF determined from the 9-in. to 3-

in. ratio in the worker location is used to adjust the measurement result to dose equivalent.

In 1997, several workplace neutron fields were measured with TLDs mounted on a phantom and at 

nearly the same time with a ROSPEC neutron spectrometer (Reilly 1998). Figure 6-12 shows the 

relative biases (Dosimeter-Spectrometer/Spectrometer) for the neutron fields. These results show a 

greater dispersion than the gamma results.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 29 of 57

Figure 6-11. Distribution of reportable neutron dose at 

INEEL for the first 9 months of 1995 (Gesell et al. 1996).

 

0

1

2

3

4

-55 to -45%

-45 to-35%

-35 to -25%-25 to -15%

-15 to -5%

-5 to 5%

5 to 15%

15 to 25%25 to 35%35 to 45%45 to 55%55 to 65%65 to 75%75 to 85%85 to 95%

Relative Bias [(Dosimeter-Spectrometer)/Spectrometer]

Number

Neutron

Figure 6-12. Neutron radiation field characterization (Reilly 1998).

The two lowest values (-0.52 and -0.51) are for TLD measurements on opposite sides of a phantom 

where the field is from 252Cf on an overhead filter bank. The phantom attenuates the radiation from 

each side so the TLDs see only half the radiation field. The next lowest value (-0.38) is for the 252Cf 

instrument calibration source at a distance of 3.5 m where the operator stands. The two highest 

values (0.94 and 0.71) are for a waste drum that was reanalyzed and a new 9-in. to 3-in. ratio 

determined because of the unsatisfactory initial result. Reilly (1998) suggests that other waste barrels 

may have had neutron sources causing interference. The remaining bias values lie between -0.16 

and 0.44 (Reilly 1998).

Total number of dosimeters with reportable dose = 54 Total number of dosimeters with reportable dose = 54
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 30 of 57

Sources of neutron exposure include neutron sources at the instrument calibration laboratories. For 

the spectra from these sources, the NTA works reasonably well. Use of small 252Cf sources for 

research began after use of albedo badges. Figure 6-13 provides spectra for the AmBe (Kluge and 

Weise 1982) and 252Cf (fission) neutron sources and the 14-MeV neutron generator as seen through 

10 cm of polyethylene shielding (Ing and Makra 1978), which is typical of the INEEL/ANL-W facilities. 

 

0.0001

0.001

0.01

0.1

1

0.000001 0.00001 0.0001 0.001 0.01 0.1 1 10 100

Neutron Energy (MeV)

Number neutrons

Fission

14 MeV

AmBe

Figure 6-13. Neutron spectra simulating INEEL facilities [13].

Most of the reactors at the INEEL/ANL-W have not had beam ports, and the neutrons have therefore 

been generally well contained away from the workplace. A reactor core environment is characterized 

not only by high neutron levels, but also by very high gamma levels. The gamma shielding is often 

water and concrete, which are also very good neutron shields. The neutron fields in the energy 

spectrum for reactors and lower are attenuated much more quickly in concrete or water than are 

gamma fields. This is not true for lead or iron, but these have not usually been used as gamma 

shields where neutrons exist. Therefore, neutron fields are generally not a large concern at an 

enclosed reactor.

6.3.4.3.1 Materials Test Reactor (MTR), Zero Power Physics Reactor (ZPPR), and Transient 

Reactor Test (TREAT) Neutron Radiation

The exceptions to the above discussion are:

• The MTR at NRTS/INEL, which operated from 1952 to 1970 and had beam ports and 

neutron beams extending onto a research floor.

• The ZPPR at ANL-W, which operated from 1969 to1992. When fueled with plutonium, 

fairly large fields existed between the halves of the reactor due to (α,n) reactions.

• The TREAT at ANL-W.

Some neutron surveys of the MTR experimental floor have been recovered (Sommers 1959, 1962; 

Hankins 1961), but these individually do not provide all components of the radiation field. Hankins 

(1961) used 2-, 3-, and 8-in. polyethylene Bonner balls in a cadmium shield to characterize the 

intermediate and fast neutrons at 21 locations around the MTR floor. He made these measurements 

and also measured the thermal neutron component at six other locations. The Hankins data have 
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 31 of 57

been reanalyzed (ORAUT 2006) using more recent Bonner response curves (Hertel and Davidson 

1985). Figure 6-14 shows the resultant neutron spectra for locations 3 and 23, which had higher 

doses and nearly the maximum low-energy intermediate and fission components, respectively. 

10

100

1000

10000

100000

1.E-01 1.E+01 1.E+03 1.E+05 1.E+07

Energy (eV)

Number Neutrons

Position 23

Position 3

Figure 6-14. Sample MTR spectra from Hankins 

Bonner measurements (ORAUT 2006).

Figure 6-15 shows the correlations of the thermal and intermediate neutron dose equivalents to the 

fast neutron dose equivalent for the ORAUT reanalysis of the Hankins data. The trend line for the 

reanalyzed intermediate-energy neutron dose equivalent has a R2 value of 0.85 and 0.91 as 

compared to a R2 value of 0.50 for the original analysis, which fits the data better. The average ratio 

of thermal to fast neutron is 0.071 ±0.025; for the low-energy intermediate to the fast, the ratio is 

0.177 ±0.057; and for the higher energy intermediate to the fast, the ratio is 0.149 ±0.046, where fast 

neutrons are taken as those above 0.2 MeV (ORAUT 2006).

The ANL personnel likely to receive neutron dose were assigned NTA film in their dosimetry packets, 

but the packets would have missed doses below 0.5 to 0.8 MeV. For the MTR spectra, the fraction of 

neutron dose equivalent above 0.8 MeV has an average value of 0.52 ±0.08 and varies from 35% to 

66% depending on the location (ORAUT 2006). To correct for missed dose on ANL reactor

experiment floors, the NTA results should be multiplied by 2 ± 0.3 (1/0.52, 0.08/0.522

) for a Monte 

Carlo dose reconstruction (ORAUT 2006).

Sommers (1962) reported thermal and fast neutron dose equivalent rates and gamma dose rates 

around the MTR beam lines. The thermal measurements near beams are believed to not be 

representative of the general workplace. Figure 6-16 shows the correlation of fast neutron dose

equivalent to the gamma dose for these measurements. The fast neutron component was less than 

the measurement capability of 1 mrem/hr for several measurements. These values are shown with 

the triangles at one-half of the smallest measured value. Using the Shapiro-Wilks Normality Test 

(Gilbert 1987) and including the insignificant fast neutron values at one-half of the minimum reported 

value suggests that the normal distribution is a slightly better description of the data than a lognormal 

distribution. The fast neutron dose equivalent is 0.42 ±0.35 of the gamma dose rate for this data set
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 32 of 57

(ORAUT 2006). Combining these results, the total neutron dose equivalent is 0.58 ±0.48 of the

gamma dose equivalent on the MTR experimental floor (ORAUT 2006). Equation 6-1 elaborates this 

calculation. The variation in the different components of neutron dose rate are so much smaller than 

the variation between the fast neutron and gamma dose equivalent rate and as such, are unimportant.

 

y = 4.633x

R2

 = 0.501

y = 0.148x

R2

 = 0.914

y = 0.049x

R2

 = 0.701

y = 0.115x

R2

 = 0.852

0.01

0.1

1

10

100

0.01 0.1 1 10

Fast Neutron Dose Equivalent Rate (mrem/hr)

Dose Equivalent Rate (mrem/hr)

InterHankins

Inter .5 eV-10 keV

Inter 10-200 keV

ThermalNeut

Figure 6-15. MTR neutron field components

(ORAUT 2006).

 

 

0.1

1

10

100

0.1 1 10 100 1000

Gamma DE Rate(mrem/hr)

Fast Neutron DE Rate (mrem/hr)

Fast Neutron

Fast Insig

Fit 0.419 *Gamma

Figure 6-16. Correlation of fast neutron dose 

equivalent to gamma dose at MTR (ORAUT 2006).
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 33 of 57

(0.42 0.35)(1 0.071 0.177 0.149) 0.58 0.48

(1 )

= ± + + + = ±

= + + +

FastN

HiInter

FastN

LoInter

FastN

Thermal

GammaDose

FastN

GammaDose

TotalNeutronDose

(Eq. 6-1) [14]

Although this limited data for a small portion of the INEEL with weak correlation might suggest

multiplying the gamma dose by 1.6 (i.e., 1+0.58) or 2.1 (i.e., 1+1.06), doing so is probably 

inappropriate because many of the people wearing NTA film would receive gamma dose at locations 

other than on the MTR experimental floor while the reactor was operating [15]. For example, health 

physics technicians would often be covering jobs with only beta-gamma fields. A craftsperson could

service pumps carrying radioactive water and not receive any neutron dose. Further extrapolating 

experience at the MTR at TRA to TREAT or ZPPR at ANL-W is a stretch [16]. While similarities of the 

neutron spectra and associated corrections are likely valid, the relative amount of gamma and neutron 

radiation might not be. In addition, dosimetry data sheets demonstrate that small neutron doses were 

as likely or more likely to be determined as small photon doses, so this sort of correction is not 

necessary [17].

A fairly large neutron field existed between the halves of the ZPPR when it was off due to 240Pu 

spontaneous fission decay and subsequent multiplication in the subcritical reactor. Rather than trust 

the NTA film in use at the time, the practice was to conduct neutron and gamma surveys using a 9” 

ball detector. Simons, Young, and Thalgott (1972) report spectral measurements (shown in Fig. 6-17)

and neutron dose assignments based on time and neutron dose rate. Neutron doses were estimated 

based on the measured neutron:gamma ratio and gamma doses measured with pencil dosimeters. 

These results were entered into the dosimetry records (Burke 2006). Shields were also used to 

minimize the reactor surface open to workers. Figure 6-18 shows the integral distribution of neutron 

dose with energy for this spectrum. The fraction of dose below the NTA cutoff is quite consistent with 

the MTR result.

Figure 6-17. ZPPR Assembly-3 Phase-1A Fast Neutron Spectrum Measured Midway between 

Halves with Both Shields Withdrawn (Simons, Young, and Thalgott 1972).
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 34 of 57

Figure 6-18. ZPPR Assembly-3 Phase-1A Fast Neutron Dose Spectrum 

Midway between Halves with Both Shields Withdrawn (Simons, Young, 

and Thalgott 1972).

6.3.4.3.2 Typical Workplace Neutron Dosimeter Hp(10) Performance

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

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

neutron dosimeters is the difference between calibration and workplace neutron energy spectra [18]. 

Table 6-11 summarizes the locations at ANL-W where neutron dose is credible.

Table 6-10. Typical workplace neutron dosimeter Hp(10) performance [19].

a

Parameter Description Potential workplace biasb

Workplace neutron 

energy spectra

NTA dosimeter response decreases and TLD 

response increases with decreasing neutron 

energy.

Depends on workplace neutron spectra. NTA 

recorded dose of record likely too low 

because of high 500-keV threshold for 

detection of neutrons. 

Exposure geometry NTA dosimeter response increases with 

increasing exposure angle and TLD response 

decreases with increasing exposure angle.

NTA recorded dose likely too high because 

dosimeter response is higher at angles other 

than AP. TLD recorded dose is lower at 

angles other than AP. Effect is highly 

dependent on neutron energy. 

Missed dose Doses less than MRL recorded as zero. Recorded dose of record is likely too low. 

The impact of missed dose is greatest in early

years because of higher MRLs and shorter 

exchange cycles of neutron dosimeters. 

Environmental effects Workplace environment (heat, humidity, etc.) 

fades the dosimeter signal.

Recorded dose of record is likely too low.

a. Judgment based on INEEL dosimeter response characteristics.

b. Recorded dose compared to Hp(10).
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 35 of 57

Table 6-11. Facilities and time periods for neutron exposure [20].

Location Time period Comments

Programmatic

TREAT 1958-1994

ZPPR 1969-1992 Between reactor halves during maintenance

ANL Neutron 

Radiography Facility 

1977-2006

Sealed sources

ZPPR 1969-1992 For instrument calibration

6.4 ADJUSTMENTS TO RECORDED DOSE

6.4.1

All dose equivalents measured at ANL-W and reported in this document used the quality factors 

based on the LET of the ionizing secondary particles established in the 1950s and used since by U.S. 

regulatory agencies [21]. In 1990, the ICRP developed new dose concepts that have been used by 

NIOSH. The quality factor, Q, as a function of LET, was replaced with a radiation weighting factor, wR,

which is a function of the neutron energy (ICRP 1990, Table 1).

Neutron Weighting Factor

A correction needs to be made to the ANL-W reported data to change from dose equivalent under 

NCRP 38 (NCRP 1971b) to a newer dose quantity (NIOSH 2006). ICRP Publication 74 (ICRP 1996) 

tabulates the ambient dose equivalent (dose equivalent at 10 mm depth in a 30-cm diameter sphere) 

for neutrons. NIOSH (2006) tabulates the ratios of organ dose equivalents to ambient dose 

equivalent, so this quantity is to be used for the conversion. Ambient dose equivalent is an ICRU 

quantity, so it uses a revised Q(L) rather than a wR, so the correction factors are not as large as in 

other TBDs. The dose equivalent for a spectrum of particle energies is the result of an integral of the 

fluence spectrum, Φ(E), times a dose equivalent conversion factor, DECF(E), which also depends on 

energy over the range of energies considered:

H DECF E E dE

E

E

∫ =

2

1

( ) φ( ) (Eq. 6-2)

These factors are incorporated into statements of dose equivalent values and calibrations following 

generally accepted principles. The conventional dose conversion factors are most clearly and 

correctly stated in ICRP Publication 21 (ICRP 1973). NCRP Report 38 tabulates a neutron flux 

density associated with the annual dose limit that is proportional to the reciprocal of the dose 

conversion factor (NCRP 1971b). Conventionally, the primary geometry considered is from one 

direction with the maximum dose in the body tabulated. More recent references (ICRU 1985; ICRP 

1987, 1996) consider the dose to individual organs for different irradiation geometries, so the more 

recent tabulations give results lower by factors up to about 10 from attenuation in the human body. 

Dosimeters are designed to respond to radiation entering the body on the side where they are located 

and work best for an AP irradiation geometry with the dosimeter on the front of the body [22].

Equation 6-2 also applies to ambient dose equivalent, except a tabulation of the ambient dose 

equivalent dose conversion factor is used (ICRP 1996). The correction factor for an energy interval is 

then the ratio of the two integrals. Because IREP uses different radiation effectiveness factors for 

different radiation types and energies, it is appropriate to use the IREP energy intervals for calculating 

the correction factors [23].
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 36 of 57

Table 6-12 lists the calculated fractions of dose equivalent in the IREP energy groups and the 

conversion factors from dose equivalent to equivalent dose for ANL-W spectra. The ratios of average

radiation weighting factor to average quality factor for the IREP energy groups have some variation, 

particularly for the 10-to-100-keV group where the energy dependence of the fluence is radically 

different for the fission and 14-MeV source than for the reactor spectrum [24]. The lower part of the 

table lists the recommended default values for the dose equivalent fractions and quality factor 

corrections. These values are combined in Table 6-13. This correction should be applied to the 

measured, the missed, and the unmonitored neutron doses either separately or after they are added 

together.

Table 6-12. Dose equivalent fractions and quality factor corrections, estimated and 

recommended [25].

IREP energy interval <10 keV 10-100 keV 100 keV-2 MeV 2-20 MeV

Spectrum calculated values

Dose equivalent fractions

Bare fission 4.4E-05 0.20 0.80

AmBe 0.15 0.85

14 MeV 10 cm poly 2.4E-08 3.1E-06 1.5E-03 1.00

MTR exp floor avg 0.18 0.06 0.49 0.28

MTR exp floor max 0.24 0.08 0.52 0.35

MTR exp floor min 0.13 0.03 0.46 0.19

ICRP 74 H*10 /NCRP 38 H 

Bare fission 1.46 1.32 1.09

AmBe 1.41 1.05

14 MeV 10 cm poly 0.69 1.47 1.36 0.93

MTR exp floor avg 0.86 1.08 1.33 1.12

MTR exp floor max 0.80 1.08 1.37 1.12

MTR exp floor min 0.92 1.08 1.30 1.12

Recommended defaults

Dose equivalent fractions

14 MeV 10 cm poly 0.05 0.95

Source calibrations 0.20 0.80

MTR exp floor 0.2 0.05 0.50 0.25

H*(10)/H 1 1.1 1.4 1.1

Table 6-13. Recommended IREP neutron energy fractions and correction factors [26].

Process Description Operations

Neutron 

energy

Default 

dose 

(%)

Ambient 

dose equiv/ 

dose equiv

Net 

correction 

factor

Instrument 

calibration

Alpha Be source 

calibrations

Cf-252 source 

calibrations

1951 1993 0.1-2 MeV 20 1.4 0.28

1993 2003 2-20 MeV 80 1.1 0.88

Waste 

characterization

RWMC SWEPP 14 MeV 

neutron generator

~1980 2003 0.1-2 MeV 5 1.4 0.07

2-20 MeV 95 1.1 1.05

Neutron source 

based research

2003 0.1-2 MeV 20 1.4 0.28

2-20 MeV 80 1.1 0.88

MTR, ZPPR, and 

TREAT reactors

Experiment floor and 

adjacent rooms during 

operation

1953 1970 < 10 keV 20 1 0.20

10 -100 keV 5 1.1 0.06

0.1-2 MeV 50 1.4 0.7

2-20 MeV 25 1.1 0.28
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 37 of 57

6.5 MISSED DOSE

6.5.1

Missed photon dose for ANL-W workers would occur where (1) there was no recorded dose because 

workers were not monitored or the dose was otherwise unavailable, or (2) a zero dose was recorded 

for the dosimeter systems for any response less than the site dose recording threshold (the MRL). 

The missed dose for dosimeter results less than the MRL is particularly important for earlier years 

when MRLs were higher and dosimeter exchange was more frequent [27]. The missed dose is 

calculated, as described in NIOSH (2002), using MRL/2 multiplied by the number of zero dose results. 

Table 6-14 lists the potential missed photon dose by year, dosimeter type, and badge exchange 

frequency. The MRLs shown are based on Cipperley (1958, 1968) and Cusimano (1963) for film; 

Kalbeitzer (1983), Gesell (1986), Gesell, Hall, and Anderson (1992), and Perry, Anderson, and Ruhter

(1993) for TLDs; and Ruhter and Perry (2002) for film and TLD. The exchange frequency must be 

determined from the individual worker’s dose submittal package for each year because it was shorter 

for highly exposed individuals and longer for those with lower doses.

Photon Missed Dose

Table 6-14. Beta/photon dosimeter period of use, type, MRL, exchange frequency, and potential 

annual missed dose [28].

Period of usea Dosimeter

Exchange 

frequency

MRLb (mrem)

Annual missed dose 

(mrem)c

Photon Beta Photon Beta

August 1951–

March 1958

INEEL Initial Film 552 DuPont 

Film

Weekly (n=52) 30 30 780 780

Monthly (n=12) 180 180

Reactor Areas DuPont 558 Film Weekly (n=52) 10 30 260 780

March 1958–

December 1966

INEEL Multi-Element DuPont 508 

Film

Weekly (n=52) 10 30 260 780

Biweekly (n=26) 130 390

Monthly (n=12) 60 180

December 

1966–February 

1974

INEEL Multi-Element DuPont 508 

Film

Weekly (n=52) 10 30 260 780

Biweekly (n=26) 130 390

Monthly (n=12) 60 180

INEEL LiF TLD Quarterly (n=4) 15 15 30 30

Semi-ann (n=2) 15 15

Annual (n=1) 7.5 7.5

February 1974–

May 1975d

INEEL Atlas TLD LiF in Teflon Monthly (n=12) 30 30 180 180

Quarterly (n=4) 60 60

Semi-ann (n=2) 30 30

Annual (n=1) 15 15

December 

1974–December 

1985d

INEEL Harshaw Two-chip TLD Monthly (n=12) 15 15 90 90

Quarterly (n=4) 30 30

Annual (n=1) 7.5 7.5

January 1986–

2006

INEEL Panasonic Four-chip TLD Monthly (n=12) 15e 15f 90 90

Quarterly (n=4) 30 30

Monthly (n=12) 10e 30f 60 180

Quarterly (n=4) 20 60

a. For many years, INEEL workers had a dosimeter assigned to each operating area where they worked, or were issued visitor 

dosimetry. All Area dosimetry was issued beginning in January 2000.

b. MRLs are based on Cipperley (1958), Cipperley (1968), Cusimano (1963), Kalbeitzer (1983), Gesell (1986), Gesell, Hall and 

Anderson (1992), Perry, Anderson, and Ruhter (1993), and Ruhter and Perry (2002).

c. Maximum annual missed dose calculated using N × MRL/2 from NIOSH (2002).

d. ANL-W began using the Harshaw dosimeter in May 1975.

e. The MRL was 15 mrem from January 1, 1986, to July 7, 1986; 10 mrem from July 7, 1986, to about September 1989; and 15 

mrem until 1993 when it returned to 10 mrem.

f. The MRL was 15 mrem from January 1, 1986, to July 7, 1986, and 30 mrem after that.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 38 of 57

6.5.2

Beta dose is important for certain cancers. Beta electrons are always above 15 keV if they are 

considered as external dose. Because the INEEL had essentially no separated plutonium, the 

electron dose is the difference between the shallow and deep doses. All nonpenetrating dose for the 

INEEL should be considered to be electron or beta dose. An alternative method is to use the 

nonpenetrating dose. The number of positive readings for beta exposure is likely to be much smaller 

than for photons, so the missed beta dose will generally be larger. The calculation of missed beta 

dose should (1) determine the measured beta dose and the number of nonzero readings; (2) 

determine the missed dose using MRL/2, the MRLs shown in Table 6-14, and the number of zero 

readings; then (3) correct the total dose by multiplying by the factor in the last column of Table 6-9 for 

the appropriate period.

Missed Beta Dose

6.5.3

Neutron radiation was present only at the ANL-W reactors, and for calibration of criticality alarms, etc. 

Most ANL-W workers were not exposed to neutrons and were not monitored for them [29]. For other 

locations, unmonitored neutron dose is very unlikely because of the very low potential for neutron 

exposure. The badge of a worker normally monitored for neutron dose might not have been read out 

and thus, the dose would be unmonitored. This would be indicated by a blank under neutrons on the 

badge report at a time when other workers have zeros [30]. A reasonable dose to assign for such a 

situation is the average dose for that worker for nearby monitoring periods [31].

Missed Neutron Dose

To calculate the missed dose, the reconstructor must first determine if the person worked near 

neutrons and which category of neutrons. This is best accomplished by review of the work location(s)

and whether a worker or others in the badge reporting group were assigned any neutron dose 

equivalent. For many years, individuals who were assigned NTA film, which was read with minimal 

dose, have zeros in the record [32]. A blank means that that individual did not have NTA film or it was 

not read because neutron exposure was believed to be unlikely [33]. It is not certain that this practice 

was followed for all time [34]. The work location code for TREAT is 20, but ZPPR did not have a 

unique code. If no neutron dose was assigned to the individual in question or to coworkers for several 

months, the dose reconstructor should assume that the worker was not exposed to neutrons [35]. 

Therefore, there would not be any unmonitored neutron dose. 

If a worker was likely to have been exposed to neutrons, the dose reconstructor should assign missed 

neutron dose equivalent using Table 6-15 for the times when workers did not have reported neutron 

dose [36]. For the period when NTA film was used, the recorded neutron dose should be multiplied 

using Table 6-16 by 1.25 for all facilities except the TREAT or ZPPR and by 2 for the TREAT or ZPPR 

when they were operating [37]. The dose equivalent is then apportioned into the IREP groups using 

Table 6-13.

For example, if in 1970 a person was an experimenter at the ZPPR, and seven of the weekly badges 

recorded a total of 300 mrem neutron dose equivalent and the other sheets all showed zeros, the 

missed dose would be 450 mrem [(52 – 7) × 20 ÷ 2] so the total dose by the badges would be 

750 mrem. Note that if some reports had blanks rather than zeros (and zeros were entered for other 

individuals), this would be an unmonitored dose that must be estimated by the first paragraph of this 

section. Although both ZPPR and TREAT did not operate on full-time schedules, so neutrons would 

only exist occasionally, it is favorable to claimants to assume that they did [38]. Because the badge 

only responds to about one-half the ZPPR or TREAT neutron dose equivalent (from Section 6.3.4.2 

and Table 6-16), the total dose equivalent is 1.5 rem (this multiplication should also be done for the 

unmonitored dose). To convert the 1.5 rem received from neutrons at the ZPPR to equivalent dose, 
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 39 of 57

multiply the total dose equivalent by the last column of Table 6-13 to get 300 mrem to the <10-keV 

group, 90 mrem to the 10-to-100-keV group, 1,050 mrem to the 0.1-to-2-MeV group, and 420 mrem to 

the above-2-MeV group, for a total equivalent dose of 1.86 rem.

Table 6-15. Neutron dosimeter type, period of use, exchange frequency, laboratory 

minimum detectable limit, and maximum annual missed dose [39].

Dosimeter Period

Exchange 

frequency

Laboratory 

MRL (mrem)

Maximum 

annual missed 

dose (mrem)

NTA film 1951-1958 Weekly 14 364

NTA film 1959–

September 1976

Weekly 20 520

Biweekly 20 260

Monthly 20 120

TLD October 1976–2006 Biweekly 15 195

Monthly 15 90

Quarterly 15 30

Table 6-16. Spectrum correction for measured and missed neutron dose [40].

Period Dosimeter Work location Correction

1951–Sept 1976 NTA TREAT or ZPPR 2

All other 1.25

October 1976–present Albedo TLD All 1

The neutron missed dose is divided into two historical periods in the following discussion. The first is 

before 1976 when only NTA film dosimeters were used with supplemental recording of thermal 

neutron doses from B-10 pencil dosimeters [41]. The second period is after 1976 when only Hankins 

albedo dosimeters were used [42]. The estimated MRLs for these neutron dosimeters are 

summarized in Table 6-15. It is possible to estimate the missed neutron dose using the MRLs 

because the neutron dosimeters were calibrated with neutron sources that had energies similar to 

those encountered in the workplace and because most of the neutrons of the calibration source had 

energies greater than the 500-to-800-keV threshold of the NTA dosimeters [43]. There was, of 

course, no threshold energy for the measurements using neutron albedo TLD badges.

6.5.3.1 Before October 1976

The use of NTA films for neutron dosimetry before 1976 is documented in various INEEL reports 

(Cusimano 1963; Cipperley 1958, 1968). As noted above, it is possible to estimate the missed dose 

using the MRLs. There are many recorded zeros in the neutron dose data for INEEL workers 

because an NTA film indicated a neutron dose equivalent that was less than the 14-mrem MRL [44]. 

When the MRL for NTA film is used in estimating the missed neutron dose, it should be multiplied by 

1.25 for most workers and by 2 for workers on the TREAT or ZPPR experimental floor [45].

6.5.3.2 After October 1976

Since October 1976, the neutron dose has been measured using the Hankins albedo-type TLD. The 

characteristics of this dosimeter are well documented (Gesell et al. 1996), and the MRL to be used in 

estimating missed dose is 15 mrem. A location-specific FNCF has been applied to convert the 

reading to dose equivalent (Gesell et al. 1996). The adjustment to equivalent dose provided in Table 

6-13 must then be made.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 40 of 57

6.6 ORGAN DOSE

The dose values should be used to calculate organ doses of interest using NIOSH (2002). It is 

recommended that the AP geometry should be assumed for the irradiation geometry and for 

conversion to organ dose [46]. The calculated neutron doses in each energy group should be 

multiplied by the conversion factors from ambient dose equivalent to organ dose for AP irradiation 

from Appendix B of NIOSH (2006). For photons before 1981, the conversion factor from exposure to 

organ dose should be used [47]. For 1981 and after, the conversion factor from deep dose equivalent 

to organ dose should be used [48].

6.7 UNCERTAINTY

Measurement uncertainties arise from many sources. For gamma rays, the standards for exposure 

have existed with only minor changes since the 1930s as required for medical uses of radiation. The 

INEEL used ionization chambers standardized by NBS for their calibrations. Use of a phantom for 

dosimeter irradiation began in the early 1980s, but backscatter only causes a minor change for highenergy photon dosimetry [49]. The over-response of the multielement film badge to deep dose in 

tissue arises from their calibration to exposure, which is somewhat more at low energies than the 

deep dose [50]. The INEEL environment did not have a significant low-energy photon field like a 

plutonium finishing plant, so the nonpenetrating component was attributed to beta radiation. A 

realistic estimate of total uncertainty for photon dosimetry is about 35% at 1 sigma [51]. This is 

roughly consistent with the results in Figure 6-9. For those measurements, the standardization 

instrument contributed some significant uncertainty.

To determine beta radiations in relation to skin cancer etc., the dose reconstructor should divide the 

reported nonpenetrating dose by the fraction of dose measured shown in Table 6-9. The uncertainty 

for beta radiation is somewhat larger than that for gamma rays at an estimated 50% at 1 sigma [52]. 

This large uncertainty is driven by the uncertainties in field geometry and because beta radiation is 

often stopped by thin materials such as clothing and air [53]. Algorithms are used to estimate the 

dose at a depth of 7 mg/cm2 from dosimeters at depths of 15 to 250 mg/cm2

, and such depth 

differences can change the signal significantly [54]. The difference between a point source irradiation 

and a planar source can confuse an algorithm [55]. Earlier techniques did not provide a thin detector 

with minimal covering, which is important for simulating the skin for beta dosimetry.

The determination of uncertainty for neutron radiations is more complex. The NTA films used before 

1975 did not see low-energy neutrons below 0.5 to 0.8 MeV (Cusimano 1963, Cipperley 1968). 

Corrections are described in Section 6.5.3 for handling this issue. The TLD albedo system provides a 

very indirect way of measuring dose equivalent to a person. Dose to people is primarily due to 

hydrogen recoils rather than 6

Li(n,α) reactions [56]. The response of the 9-in. PNR-4 detector used to 

standardize the TLD measurements arises from a different process than dose deposition in the 

human body. The total uncertainty for neutrons is probably a GSD of about 1.6. The uncertainty of 

the base curve for calculating FNCF is a GSD of about 1.35 (Gesell et al. 1996) although this 

uncertainty does not contribute in the DOELAP process, resulting in a total GSD of 1.7 for albedo 

neutron measurements [57].

The cause of the greatest uncertainty for neutrons is the variation of dose caused by the position of 

the organ within the body. For 1-MeV neutrons, the dose facing the source is about 1,000 higher than 

the dose on the back side of a 30-cm-diameter sphere of tissue-like material [58]. In a work 

environment, the direction of the neutrons might be unknown, but it is often from many directions,

which reduces the impact of this uncertainty driver. For simplicity and because it often is true, this 

analysis assumed that the geometry of exposure is AP [59]. The apparent discrepancy in Figure 6-
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 41 of 57

11, in which the dosimeters showed about one-half of the spectrometer readings, occurred because

the spectrometer does not simulate the attenuation of the body which, in the measurement, was 

simulated by the phantom, so the reading was high by a factor of 2.

6.8 ATTRIBUTIONS AND ANNOTATIONS 

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

information, conclusions, and recommendations provided to assist in the process of worker dose 

reconstruction. These callouts are listed here again in the Attributions section of the document, with 

information provided to identify the source and justification for each associated item. Conventional 

references are provided in the next section of this document, linking data, quotations, and other 

information to documents available for review on the ORAU Team servers.

Norman Rohrig served as the initial Document Owner for this document. Mr. Rohrig was previously 

employed at INL which shared boundaries with ANL-W and used the same dosimetry systems and his 

work involved management, direction or implementation of radiation protection and/or health physics 

program policies, procedures or practices related to atomic weapons activities at the site. This 

revision has been overseen by a new Document Owner, who is fully responsible for the content of this 

document, including all findings and conclusions. Mr. Rohrig continues to serve as a Site Expert for 

this document because he possesses or is aware of information relevant for reconstructing radiation 

doses experienced by claimants who worked at the site. In all cases where such information or prior 

studies or writings are included or relied upon by the Document Owner, those materials are fully 

attributed to the source. Mr. Rohrig’s Disclosure Statement is available at www.oraucoc.org.

[1] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This statement generally applied to high level bosses, such as DOE officials and company 

managers, who occasionally came to the site for meetings.

[2] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This table is a requirement from ORAUT-PROC-0031, “DOE Technical Basis Document

Development, Review, and Approval Process” and uses values based on experience and from 

the Hanford Site Profile.

[3] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This table is adopted to INL from the Hanford Site Profile and uses values based on 

experience at the INL.

[4] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

The raw photon spectrum of the expected nuclide fission product mixes was considered and 

the dose fraction in the 30-250 keV band was 1.7% to 5.1% of the total dose. Monte Carlo 

calculations were performed using a code provided by SC&A, and the dose contribution from 

0-250 keV photons through a concrete wall was 0.4% to 7.4%. Combining these effects 

shows that the 25%/75% split is appropriate.

[5] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

The contact radiation field on these devices is less than 0.5 mrem/hr (phone conversation

with Larry Burke, Radiological Engineer at ANL-W and INL,11/06/06, Rohrig 2006).
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 42 of 57

[6] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

Because of the energy dependence of photon cross sections, low energy photons are 

absorbed more strongly than high energy photons, so spectral hardening (reduction of low 

energy photons) results.

[7] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This statement follows from the fact that beta particles are directly ionizing and thus have a 

limited range. For there to be any dose, they must be lightly shielded. 

[8] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

Retrospectively, there is no simple way of knowing the beta emitters that caused employee X’s 

exposure in week Y, so this is a generic approach to the issue.

[9] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

As discussed in the paragraph above, the end-point energy distribution of beta emitting 

nuclides was converted to a distribution of ranges and then used to generate this EXCEL plot.

[10] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This choice of angles and energies was used to generate a simplified depth dose curve.

[11] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This correction factor is unlikely to have been applied during the calibration factor. The net 

correction is the ratio of corrections for the calibration spectrum and the field spectrum, so the 

net effect will be smaller than the corrections in Table 6-9.

[12] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

Dosimeter thicknesses have been stated in previous paragraphs. Based on the theoretical 

approach described, the correction factors are derived. 

[13] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This EXCEL plot was generated from spectra previously cited [Ing and Makra (1978) and 

Kluge and Weiss (1982)].

[14] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This equation is a direct result of using four energy groups for the neutrons and applying mean 

values as determined in OTIB-0009 for the different ratios.

[15] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

Based on the judgement of Mr. Rohrig, using the gamma dose to estimate the neutron dose is 

not recommended because the two fields may not be directly related to each other, they often 

have different sources.

[16] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

Since the gamma and neutron fields do not have a common source, there is no reason to 

believe that the ratios from one facility would apply at the other one.

[17] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

Using photon to neutron ratios to derive neutron doses assumes that valid photon data are 

considerably more likely to exist than valid neutron data. This does not seem to be the case 

on examining dose records.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 43 of 57

[18] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This subjective statement is made based on the experience of Mr. Rohrig and knowledge of 

personnel neutron dosimetry; the performance of personnel neutron dosimeters is significantly 

dependent upon the similarity between the calibration and workplace radiation field.

[19] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This standard TBD table is based on information in the Hanford TBD and, in the judgment of 

Mr. Rohrig, performance of INL dosimeters for similar parameters of performance.

[20] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This table is based on facility operating history cited in ANL-W 2 and the potential for worker 

exposure to neutron radiation based on the facility type and processes judgment of Mr. Rohrig.

[21] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This is based on guidance in OTIB-0055. 

[22] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

For neutrons, shielding by a human is significant (factor of 1000 for 1 MeV neutrons) and thus 

dosimeters need to be facing the radiation source to read reliably.

[23] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This is based on guidance in OTIB-0055.

[24] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

The values in Table 6-12 are different, namely 1.08 and 1.46 or 1.47.

[25] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This table was derived with spectral data in an EXCEL spreadsheet, as shown in Figure 6-13 

and 6-14, with the associated energy dependent quality factor and dose effectiveness factors. 

The last four rows are averages of the associated calculated values higher in the table.

[26] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This table is derived from the values in Table 6-12.

[27] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

The calculated missed dose, based on OCAS-IG-001 guidance and described in the 

succeeding text, is based on the number of dosimeter exchanges multiplied by the minimum 

detection level divided by 2 for all reported results less-than the minimum detection level 

divided by 2.

[28] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

The values in this table are required to calculate the missed dose per ORAUT-PROC-0031 

and are based on judgment of Mr. Rohrig.

[29] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This statement is made based on understanding of the type and variety of work performed at 

ANL-W.

[30] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This statement is made based on studying dose record sheets and conferring with INL 

dosimetry staff.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 44 of 57

[31] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This judgment is made by Mr. Rohrig based on reviews of historical INL dosimetry information

and the reference Watson, et al., 1994.

[32] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This judgment is made by Mr. Rohrig based on reviews of historical INL dosimetry information.

[33] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This judgment is made by Mr. Rohrig based on reviews of historical INL dosimetry information.

[34] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This judgment is made by Mr. Rohrig based on reviews of historical INL dosimetry information.

[35] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This judgment is made by Mr. Rohrig based on reviews of historical INL dosimetry information.

[36] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This judgment is made by Mr. Rohrig based on reviews of historical INL dosimetry information.

[37] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

These values are based on weighting neutron spectra with dose conversion factors to 

determine the fraction of the dose below 0.8 MeV.

[38] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

Assuming a non-zero value when the machine is off is favorable to claimants.

[39] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

The values in this Table 6-15 are required to calculate the missed dose per ORAUT-PROC0031 and are based on judgment of Mr. Rohrig.

[40] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

These values are based on weighting neutron spectra with dose conversion factors to 

determine the fraction of the dose below 0.8 MeV.

[41] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This reflects the time period discussed in § 6.3.2.7.

[42] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This reflects the time period discussed in § 6.3.2.8.

[43] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

A perfect calibration source would have all its radiation detected by the detector being 

calibrated. If most of the radiation is detected, the adjustment would be small and not 

contribute a major uncertainty. 

[44] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

By definition of the MRL (minimum reporting level), values less than the MRL are reported as 

zero (practice at INL) or a blank. 

[45] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This correction is listed in Table 6-16.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 45 of 57

[46] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This is based on the common practice of wearing the dosimeter on the front of the body and 

often facing the radiation source during work. Early program guidance recommended using 

the A-P geometry exclusively.

[47] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This recommendation is made because the measurement quantity of dose in that time period 

was exposure.

[48] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This recommendation is made because the measurement quantity of dose in that time period 

was deep dose equivalent.

[49] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

For high energy, the angular distribution is primarily forward. For example, see Figure 2-19 of 

Knoll (1989).

[50] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

At low energies, air, which has very little hydrogen compared to tissue, has a higher cross 

section than tissue because of the Z dependence of the photoelectric effect.

[51] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This is an estimate, partially based on Figure 6-9 and partially based on uncertainties of many 

error processes which contribute uncertainty.

[52] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This is an estimate, partially based on Figure 6-9 and partially based on uncertainties of many 

error processes which contribute uncertainty.

[53] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

The range of beta particles is shown in Figure 6-10.

[54] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

These estimated corrections are provided in Table 6-9, and they are not small.

[55] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

For this reason, DOELAP has had different categories for point and slab beta geometries. 

[56] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

Lithium is only a trace constituent of the human body. Dose from neutron radiation to people 

is primarily due to hydrogen atom recoils in tissue rather than the 6

Li(n,α) neutron absorption 

reactions of the neutron dosimeter.

[57] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

Most of this uncertainty is from the Facility Neutron Correction Factor (FCNF) for the field of 

the neutron exposure.

[58] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

See Figure 16 of NCRP 38 (1971b).
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 46 of 57

[59] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This is based on the common practice of wearing the dosimeter on the front of the body and 

often facing the radiation source during work. It is understood that early program guidance 

recommended using the A-P geometry exclusively.
Document No. ORAUT-TKBS-0026-6 Revision No. 01 Effective Date: 05/15/2007 Page 47 of 57

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AEC (U.S. Atomic Energy Commission), 1963, Standards for Radiation Protection, AEC Manual, 

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Attix, F. H., and W. C. Roesch, 1968, Radiation Dosimetry, Second Edition, Academic Press.

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BRH (Bureau of Radiological Health), 1970, Radiological Health Handbook, Rockville, MD, January.

Burke, L. L., 2006, Telephone conversation with N. Rohrig, ~September 15.

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Cipperley, F. V., R. W. Henry, and J. P. Cusimano, 1965, “Personnel Monitoring Practices at the 

National Reactor Testing Station,” Symposium on Personnel Dosimetry for Accidental 

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Vienna, Austria, March 8-12, 1965. [SRDB Ref ID: 8105]

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GLOSSARY

1/E spectrum

For neutrons, the number of neutrons in an energy interval scales as the width of the energy 

interval divided by the energy of the neutrons in that interval.

beta particle (β)

See beta radiation.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

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).

dose equivalent index

Maximum dose equivalent within the ICRU sphere centered at the point in space to which the 

quantity is assigned, HI. The outer 0.07-mm-thick shell is ignored. It is also called the 

unrestricted dose equivalent index.

deep dose equivalent index

Maximum dose equivalent in the ICRU sphere within a core radius of 14 cm. The sphere is 

centered at the point in space to which the quantity is assigned. This quantity is one of the two 

restricted dose equivalent indices.

DOE Laboratory Accreditation Program (DOELAP)

Program for accreditation by DOE of DOE site dosimetry programs based on performance 

testing with independent review every 2 years. 

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual.

effective dose equivalent, HE

The weighted average of the dose equivalents in certain organs or tissues of the body, HT, 

each weighted by an organ weighting factor, WT. The organ weighting factors were chosen by 

the ICRP to reflect the relative risk of death from cancer or occurrence of severe hereditary 

effects in the first generations after uniform whole body exposure.

exposure

(1) In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.
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exposure-to-dose-equivalent conversion factor for photons (Cx)

Ratio of exposure in air to the dose equivalent at a specified depth in a material of specified 

geometry and composition. Cx factors are a function of photon energy, material geometry 

(e.g., sphere, slab, or torso), and material composition (e.g., tissue-equivalent plastic, soft 

tissue without trace elements, or soft tissue with trace elements). 

linear energy transfer (LET)

The lineal rate of local energy deposition by a charged particle.

minimum reporting level (MRL)

Level below which an analytical dose is not recorded in the worker’s dose record, usually 

based on a site-specific policy decision. The recording level is not necessarily the same as 

the minimum detectable amount or activity for that measurement. Also called less-than value, 

minimum reportable dose, minimum recordable or recording dose, recording level, and 

reporting level.

nonpenetrating dose

Dose from beta and lower energy photon radiation, often determined from the open window 

dose minus the shielded window dose.

pencil dosimeters

See pocket ionization chamber. 

penetrating dose equivalent

Photon dose measured by shielded film or elements plus neutron dose equivalent. 

Essentially, personal dose equivalent HP(10).

personal dose equivalent, Hp(d)

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeters 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. These are 

noted as Hp(0.07) and Hp(10), respectively. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as dose quantity for radiological 

protection. 

pocket ionization chamber

Cylindrical monitoring device commonly clipped to the shirt or laboratory coat pocket to 

measure ionizing radiation. Also called pencil, pocket pencil, pencil dosimeter, and pocket 

dosimeter.

polymethyl methacrolate

Scientific name for plastic commonly known as Lucite or Plexiglas.

ROSPEC

Model name for a rotating neutron spectrometer marketed by Bubble Technology Industries. 

The system uses four pressurized proportional counters at different pressures to distinguish 

neutron reactions from gamma events and estimate a neutron energy distribution from 50 keV 

to 4.5 MeV.
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redact

To select item(s) to be visible for viewing or for publication by obscuring others.

shallow absorbed dose (Ds)

Absorbed dose at a depth of 0.007 centimeters (7 milligrams per square centimeter) in a 

material of specified geometry and composition.

shallow dose equivalent (Hs)

Dose equivalent in units of rem or sievert at a depth of 0.007 centimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

tissue rad

Absorbed dose in tissue.