ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 59

Document Title:

Idaho National Laboratory – Occupational 

External Dosimetry

Document Number: ORAUT-TKBS-0007-6 

Revision: 02

Effective Date: 06/18/2007

Type of Document TBD

Supersedes: Revision 01

Subject Expert(s): Steven L. Bump

Site Expert(s): Norman D. Rohrig

Approval: Signature on File Approval Date: 04/27/2007

Jo Ann M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 05/01/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 04/26/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 05/31/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 06/18/2007

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-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 2 of 59

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

04/06/2004 00 New technical basis document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational External 

Dosimetry. First approved issue. Initiated by Norman D. Rohrig.

11/07/2006 01 Document revised as a result of biennial review. Approved Revision 

01. Incorporates formal internal review comments. Revision 

constitutes a total rewrite of document. Adds Section 6.5.1 to 

incorporate comments from the Worker Outreach meeting with the 

PACE Local 8-0652 on April 28, 2004. Incorporates formal 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 Norman D. Rohrig.

06/18/2007 02 Approved Revision 02 initiated to incorporate Attribution and 

Annotation section. A minor change was made to Sections 6.3.4.5 

and 6.5.4.2, which will reduce the dose for a very few individuals. 

Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. 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-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 3 of 59

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

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

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

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 ...................................................................................18

6.3.2.1 Initial Film Badge.......................................................................................... 18

6.3.2.2 Multiple-Filter NRTS Film Badge.................................................................. 20

6.3.2.3 Original Lithium Fluoride Teflon Thermoluminescent Dosimeter System...... 20

6.3.2.4 Automatic Thermoluminescent Analyzer System.......................................... 21

6.3.2.5 Harshaw Two-Chip Thermoluminescent Dosimeter System......................... 21

6.3.2.6 Panasonic Four-Chip Thermoluminescent Dosimeter System...................... 21

6.3.2.7 Nuclear Track Emulsion, Type A for Neutrons.............................................. 22

6.3.2.8 Neutron Albedo Dosimetry ........................................................................... 23

6.3.3 Calibration ..................................................................................................................23

6.3.3.1 Beta-Gamma Radiation................................................................................ 23

6.3.3.2 Neutron Calibration ...................................................................................... 25

6.3.4 Workplace Radiation Fields ........................................................................................26

6.3.4.1 Gamma Radiation ........................................................................................ 28

6.3.4.2 Beta Radiation ............................................................................................. 29

6.3.4.3 Neutron Radiation ........................................................................................ 32

6.3.4.4 Materials Test Reactor Neutron Radiation.................................................... 34

6.3.4.5 Test Area North Fuel Storage Casks............................................................ 36

6.3.4.6 Typical Workplace Neutron Dosimeter Hp(10) Performance......................... 37

6.4 Adjustments to Recorded Dose.............................................................................................. 37

6.4.1 Neutron Weighting Factor...........................................................................................37

6.5 Missed Dose ..........................................................................................................................39

6.5.1 Dosimeter Not Worn ...................................................................................................39

6.5.2 Photon Missed Dose ..................................................................................................39

6.5.3 Missed Beta Dose ......................................................................................................39

6.5.4 Missed Neutron Dose ................................................................................................. 40

6.5.4.1 Before October 1976.................................................................................... 41

6.5.4.2 October 1976 to Present .............................................................................. 42

6.6 Organ Dose............................................................................................................................42

6.7 Uncertainty.............................................................................................................................42

6.8 Attributions and Annotations .................................................................................................. 43

References ........................................................................................................................................49

Glossary ............................................................................................................................................58
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LIST OF TABLES

TABLE TITLE

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

PAGE

14

6-2 Reasons Codes .....................................................................................................................15

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

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

6-5 1981 FNCFs...........................................................................................................................23

6-6 Laboratory sources of uncertainty for beta/photon dosimeter calibration parameters ............. 24

6-7 Common sources of laboratory bias in the calibration parameters for neutron dosimeters ..... 26

6-8 Penetrating photon dose (rem) percentiles by year for monitored workers ............................. 27

6-9 Selection of IREP beta and photon energies for INL facilities................................................. 30

6-10 Beta dosimeter thicknesses and associated underreporting................................................... 31

6-11 Facilities and periods for neutron exposure............................................................................ 32

6-12 Typical workplace neutron dosimeter Hp(10) performance ..................................................... 37

6-13 Calculated and recommended dose equivalent fractions and quality factor corrections ......... 38

6-14 Recommended IREP neutron energy fractions and correction factors.................................... 39

6-15 Beta/photon dosimeter period of use, type, MRL, exchange frequency, and potential 

annual missed dose ...............................................................................................................40

6-16 Neutron dosimeter type, period of use, exchange frequency, laboratory minimum 

detectable limit, and maximum annual missed dose............................................................... 41

LIST OF FIGURES

FIGURE TITLE

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

PAGE

13

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

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

6-4 Badge 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 in Figure 6-4 ................................. 17

6-6 Personnel exposure questionnaire partially completed for a hypothetical case ...................... 18

6-7 Response of DuPont 508 film with various filters to 140 mrem uranium beta and 100 mR 

of different energy photon irradiation...................................................................................... 19

6-8 Probability density of neutron spectrum from a 241Am-Be (α,n) source ................................... 26

6-9 Collective dose for INL personnel, 1951 to 1990.................................................................... 28

6-10 Gamma and beta radiation field characterization ................................................................... 29

6-11 Distribution of beta ranges ..................................................................................................... 31

6-12 Distribution of reportable neutron dose for the first 9 months of 1995..................................... 33

6-13 Neutron radiation field characterization .................................................................................. 33

6-14 Neutron spectra that simulate INL facilities ............................................................................ 34

6-15 Sample MTR spectra from Hankins Bonner measurements ................................................... 35

6-16 MTR neutron field components .............................................................................................. 35

6-17 Correlation of fast neutron dose equivalent to gamma dose at MTR ...................................... 36
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANL-W Argonne National Laboratory–West

ANSI American National Standards Institute

AP anterior-posterior

ARA Army (later Auxiliary) Reactor Area

ATLAS Automatic Thermoluminescent Analyzer System

CFA Central Facilities Area

CFR Code of Federal Regulations

Ci curie

cm centimeter

CPP Chemical Processing Plant

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

EBR Experimental Breeder Reactor

ECF Expended Core Facility (at NRF)

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EPRI Electric Power Research Institute

ERDA U.S. Energy Research and Development Administration 

FNCF facility neutron correction factor

g gram

GSD geometric standard deviation

H*

(10) ambient dose equivalent 

Hp(10) personal dose equivalent at 10 mm depth in tissue

Hp(d) personal dose equivalent

Hp,slab(d) personal dose equivalent (slab phantom)

hr hour

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

ILTSF Intermediate-Level Transuranic Storage Facility

in. inch

INEEL Idaho National Engineering and Environmental Laboratory

INEL Idaho National Engineering Laboratory

INTEC Idaho Nuclear Technology and Engineering Center

IREP Interactive RadioEpidemiological Program 

keV kiloelectron-volt, 1,000 electron-volts

LET linear energy transfer

LOFT Loss-of-Fluid Test

LPTF Low Power Test Facility
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MeV megaelectron-volt, 1 million electron-volts

mg milligram

mm millimeter

mR milliroentgen

mrad millirad

mrem millirem

mrep millirep

MRL minimum reporting level

mSv millisievert

MTR Materials Test Reactor

n neutron

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurement

NIOSH National Institute for Occupational Safety and Health

NRF Naval Reactors Facility

NRTS National Reactor Testing Station

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

OMRE Organic-Moderated Reactor Experiment

ORAU Oak Ridge Associated Universities

OW open window

PBF Power Burst Facility

PEQ Personnel Exposure Questionnaire

PER Program Evaluation Report

POC probability of causation

R roentgen

RBE relative biological effectiveness

RESL Radiological and Environmental Sciences Laboratory

RWMC Radioactive Waste Management Complex

SL-1 Stationary Low-Power Reactor No. 1

SMC Specific Manufacturing Capability

SPERT Special Power Excursion Reactor Test

SRDB Ref ID Site Research Database Reference Identification (number)

SWEPP Stored Waste Examination Pilot Plant

TAN Test Area North

TBD technical basis document

TLD thermoluminescent dosimeter

TRA Test Reactor Area

TREAT Transient Reactor and Experiment Test

TSF Tower Shielding Facility

U.S.C. United States Code

WERF Waste Experimental Reduction Facility

wk week
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yr year

Z atomic number

ZPPR Zero Power Plutonium (later Physics) Reactor

α alpha particle

§ section
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6.1 INTRODUCTION

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

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

documents that provide 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. 

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.

• Radiation from naturally occurring radon present in conventional structures

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

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 9 of 59

6.1.1 

This document describes the processes and results for measuring external doses at the Idaho 

National Laboratory (INL), formerly known as the National Reactor Testing Station (NRTS), the Idaho 

National Engineering Laboratory (INEL) and the Idaho National Engineering and Environmental 

Laboratory (INEEL). For consistency, this document uses INL except in special cases. 

Purpose

6.1.2 

From the start of operations in 1951 (when it was NRTS), a branch of the U.S. Atomic Energy 

Commission (AEC) Idaho Operations Office provided external dosimetry resources and services at 

INL until 1989, when DOE transferred that responsibility to the prime operating contractor. Despite 

the fact that INL had several contractors at a time and that DOE changed contractors often, the 

external dosimetry process has remained under technical management of a single organization with 

responsibilities for dosimetry development, operational dosimetry, and radiological records; thereby

providing a stable external dosimetry system. 

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, the National Council on 

Radiation Protection and Measurements (NCRP), the International Commission on Radiological 

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

parameters. Dose parameters measured by the INL dosimetry system have received further 

definition. INL has reported doses as penetrating and nonpenetrating. The penetrating dose 

corresponds to the deep dose equivalent, and the sum of the nonpenetrating and penetrating doses 

corresponds to the shallow dose equivalent. 

Horan and Braun (1993), Attix and Roesch (1968, p. 6-18), and Meinhold (1975) discuss the history of 

radiation protection requirements from the 1930s. On September 29 and 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- 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 units of rep (rep is from roentgen-equivalent-physical; 1 rep is equivalent to 93 

ergs absorbed per gram of tissue). The maximum permissible dose to the bone marrow was 

0.3 rep/wk.

In 1949, the newly formed National Committee on Radiation Protection (now the National Council on 

Radiation Protection and Measurement) issued NCRP Report 7 through the National Bureau of 

Standards as Handbook 42 (NBS 1949, p. 6), which recommended a permissible dose of 0.3 R/wk

(15 R/yr) for occupational workers (NBS is now the National Institute of Standards and Technology). 

The term dose was undefined. A roentgen was defined as the quantity or dose of X-rays such that 

the associated ionization per 0.001293 g of air (1 cm3 at standard temperature and pressure) 

produces 1 electrostatic unit of charge of either sign. The unit roentgen only applies to X-rays or 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 10 of 59

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).

An April 1952 site manual stated the limit as “0.3 rep/wk 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). The manual does not 

mention a quarterly or an annual limit.

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

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

using the rad (radiation absorbed dose), which was defined as 100 ergs/g (ICRU 1954). It specified 

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

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

quantity rem (roentgen equivalent in man) and 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 that was 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 twice 

later as U.S. Energy Research and Development Administration (ERDA) Manual Chapter 0524

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

In 1957, NBS Handbook 63 specified a dependence of the relative biological effectiveness (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 quantity kerma (kinetic energy released per unit ma

In 1971, NCRP Report 39, Basic Radiation Protection Criteria (NCRP 1971a), recommended an

annual dose limit of 5 rem and the elimination of the quarterly limit. In April 1975, ERDA reissued 

Manual Chapter 0524 (ERDA 1975), which invoked the 5-rem annual dose limits in NCRP Report 39 

and required the addition of internal and external dose equivalents if both were known. 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 [3 rem].” Personnel monitoring equipment for each individual 

was required for external radiation. ERDA (1975, App. p. 10) stated that “to achieve optimum 

accuracy, personnel dosimeters should comply with the performance parameters in American 

National Standards Institute (ANSI) standards N13.5 (ANSI 1972), N13.7, and N13/42 WG1 Final 

draft 1974”. 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 guidance in NCRP 

(1971b) for interpolating in energy cannot be accomplished with an instrument. The dose equivalent 

ss) in 1962 (ICRU 

1962). 
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conversion factors in ICRP Publication 21 and the associated interpolation with energy do not present 

that problem (ICRP 1973). 

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. 

Order 5480.11 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, which led to the 

development of ANSI N13.11 (HPS 1983) 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 in 1993 to the personal dose equivalent Hp(d) where d

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

measured (ICRU 1993). 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). In 1986, 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, which are 

invoked by 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).
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 12 of 59

The quantities to be measured and reported by the dosimetry systems have evolved over the last 

50 years, but the changing definitions have had little effect on dosimetry measurements because, for 

gamma radiation, the differences are small [1]. 

6.3 DOSE RECONSTRUCTION PARAMETERS

6.3.1 

INL policy was that personnel who were expected to receive any radiation dose or whose work was 

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

at the operational area entrance security gates for INL facilities. Control badges, which are used to 

subtract background radiation, have also been located at the gates (Ruhter 2003). This practice 

might have led to the subtraction of environmental radiation from site activities, which would have 

reduced the reported doses (ORAUT 2003). Environmental radiation levels have been monitored for 

most of the life of INL using film badges first and thermoluminescent dosimeters (TLDs) later. Table 

4-13 in the INL environmental dose TBD (ORAUT 2004a) lists the results of this monitoring at facility 

fenceline locations near the security gates interpreted as net annual dose (i.e., in addition to natural 

background). 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 might have occasionally visited site facilities but did little work with radiation 

had badges at several different facilities [2]. It is not appropriate to base missed doses on the multiple 

badges issued. During the early years at INL, the badge exchange frequency was not the same for all 

workers. Workers with low probability of exposure were placed on a longer exchange cycle than 

those with more chance of exposure (Cusimano 1972). Therefore, missed doses should be based on 

the actual exchange frequency for a person, and the frequency can be determined from the worker’s 

data package.

The INL 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 the dosimetry information for former INL workers, which should include a dose 

summary for the employment period and a copy of each weekly, monthly, quarterly, etc. form with the 

work location, so the individual file could be several inches thick in hard copy. Each sheet is redacted 

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

recorded information about the exchange period for the person for that period. The following five

figures are a partial example set of redacted dose reporting forms.

From 1951 to 1958, the INL dosimetry staff recorded dose daily on a dose card (Figure 6-1), rezeroed 

the 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 ionization 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 zero 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 (OW, “O. W.” in the figure) result was zero, so the shielded film was not read. 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 13 of 59

On the other two, the OW 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). 
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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). 

Table 6-1. Area codes that could be in worker dose files (Hill-Ellis 2003). 

Area code Description Area code Description

1 AEC Headquarters Bldg 20, 261, 264 TREAT

2 EBR – I 21 LX

3, 034, 035 CFA 22 Gas-Cooled Reactor Experiment

4, 042,045 MTR, TRA 23 OX at OMRE

5, 053, 055 ICPP 24 ARHG

6 Naval Reactors Facility (NRF) 25 No information available

7 TAN (GE) 26, 263,265 EBR-II

8 Services 27 ML-1

9 NX (X is construction) at NRF 28 Onsite Survey

10 AX at TAN 29 Offsite Survey

11, 113 CX at CPP 30 Advanced Nuclear Propulsion

Program at Stationary Low-Power 

Reactor No. 1 (SL-1)

12 EX at EBR 31 Shield Test Pool Facility

13, 133,135 SPERT, PBF 65 ECF

14 Organic-Moderated Reactor 

Experiment (OMRE)

66 Nonsecurity

15 SX at SPERT 67 Division of Compliance
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 15 of 59

Area code Description Area code Description

16 SL-1 68 STEP

17, 333 MX at MTR 69 LPTF (Phillips & AEC)

18, 814,815 WP, RWMC 71 CADRE (guard force)

19, 772, 775 TAN (Phillips & AEC) 774,776 SMC

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

chambers), the INL 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 reading 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 

the page. Table 6-1 lists the codes for the areas. The next column was unused and dropped later. 

The next column was for the reason codes listed in Table 6-2. Computer input card codes are listed 

in Tables 6-3 and 6-4, respectively.

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 

and 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 OW 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 Sensitive Beta-Gamma 8 Summary Card

4 Insensitive Gamma 9 Summary Card

5 Slow Neutron 0 Total Body Results Card

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

though many workers were not exposed to neutrons and did not receive neutron dosimeters, zeros 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 16 of 59

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

assigned in a computer-generated report.

For the early computer records (about 1975 to 1984), the penetrating and nonpenetrating labels are 

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

and then the neutron result.

Figure 6-4 is a listing of some doses received during recovery from the 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. Badge results from January 1961 for work in recovery 

from the SL-1 accident (Vivian and Rockhold 2003, p. 50). 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 17 of 59

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

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

Figure 6-4. When there was a question about an assigned dose value, a Personnel Exposure 

Questionnaire (PEQ) was normally initiated as shown in 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-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 18 of 59

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

The badging system in place when operations began at the NRTS was called 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, 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 OW. Sensitive and insensitive DuPont 552 film was used for 

beta-gamma dosimetry for most locations; DuPont 558 film (a combination of types 508 sensitive and 

1290 insensitive films) was used at two reactor areas (Cipperley 1958). 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 19 of 59

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 OW density that corresponded to the gamma exposure was subtracted from the measured OW

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

DuPont 552 film has a threshold level of about 30 mR, and DuPont 558 film has a threshold level of 

about 10 mR (Cipperley 1958). The OW 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 OW overresponds 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). Because there was no isolated plutonium at the 

NRTS, the nonpenetrating radiation is considered to be beta radiation. Using a cadmium filter with its 

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

filters to 140 mrem from uranium beta and 100 mR of 

different energy photon irradiation. The original badge

used the OW and cadmium-shielded films. The multiplefilter badge used all three filters plus the OW (Cipperley 

and Gammill 1959).

high Z severely attenuates the photons that get to the film, so the over-response 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 the atomic number Z (it depends only on the density) so the 1-mm cadmium filter 

(~900 mg/cm2

) simulates a tissue depth of 9 mm for beta radiation. 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 20 of 59

Wrist badges used the same package attached to a wristband. 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 these chambers and 

recorded the results 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 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 that contained filters 

of 1 mm of cadmium, 0.013 mm of silver, and 0.5 mm of 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, which resulted in an absorber 

thickness of 100 mg/cm2 in the OW that filtered out beta radiation below 360 keV.

With the four absorbers, it was possible to separate beta radiation from photon radiation and to 

somewhat determine photon energy. 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 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 (MRL) of 10 mrem was used for both beta and gamma radiation (Horan 

1963).

The Instrument and Development Branch developed an automatic film reader and densitometer 

(Purcell and McGary 1963). An algorithm based on probit-corrected 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 OW, 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 other than uranium, the analysis had greater uncertainty.

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 at which 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 after the 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 be, and

in many cases was, fastened on a belt around the worker to determine a beta:gamma ratio for each 

particular entry (Cipperley, Henry, and Cusimano 1965). 

6.3.2.3 Original Lithium Fluoride Teflon Thermoluminescent Dosimeter System

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

than 0.5 rem/yr were given 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

) and impregnated with 28 mg LiF], were mounted in a badge behind an OW and a 1-mm 

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

with the Teledyne Model 7300 TLA reader. LiF was chosen because the average Z is close to that of 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 21 of 59

air and tissue, which results 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 radium-gamma response (within 

10% to 70o) is superior to film because the material acts like an ionization chamber. For normal 

monitoring, only the OW TLD was read, and that was considered the 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 (i.e., with longer exchange cycles, typically 

3 or 6 months). 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 quarterly 

processing was used for 960 TLDs (Cusimano 1972). Employees on a monthly badge exchange 

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 Thermoluminescent 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. This system 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. An MRL of 30 mR was used for gamma and beta (Walker 1974).

6.3.2.5 Harshaw Two-Chip Thermoluminescent Dosimeter System

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

beginning in December 1974 for the Idaho Chemical Processing Plant (ICPP), in February 1975 for 

the prime contractor at TAN, TRA, etc., and in May 1975 for Argonne National Laboratory–West 

(ANL-W). This commercial Harshaw system used two LiF TLDs that were 240 mg/cm2 thick. In 1976, 

holes were punched in the security badges to restore the OW. 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 of beta energy dependence. The initial thin aluminum 

filter (density thickness of 350 mg/cm2

) allowed higher energy beta radiation to expose the chip that 

was used for measuring the penetrating dose (at 1,000 mg/cm2

), so it was changed to 2 mm in July 

1977 (INEL ca. 1978). The practice was to read only the OW chip to determine if the nonpenetrating 

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

be read and the penetrating and nonpenetrating doses would be computed (Kalbeitzer 1983).

6.3.2.6 Panasonic Four-Chip Thermoluminescent Dosimeter System

In 1986, with the advent of DOELAP, INL changed to the four-element Panasonic 814 AS4 System 

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

and aluminum filtration provided 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 provided a strong low-energy photon response. The elements were 15 mg/cm2 thick. Element 1 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 22 of 59

had filtration of 16 mg/cm2, element 2 had filtration of 58 mg/cm2 of plastic, and elements 3 (CaSO4) 

and 4 had filtration of 550 mg/cm2 of plastic and 50 mg/cm2 of aluminum, respectively. Although none 

of the elements was 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 DOELAP for the 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 MRL 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) film 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 three (four in some 

references) grains of emulsion.

The minimum assigned dose 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). 

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). Blanks are

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 other technicians, which provided 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 a 10- and 20-mrem neutron dose equivalent

(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), which resulted in 

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

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

The field health physics personnel were aware of the energy limitations of the NTA badge (Sommers 

1967, 1969), and they 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. Procedures were established using boron trifluoride (BF3) pencils to monitor 

neutrons in the Radioactive Waste Management Complex (RWMC) transuranic waste areas where 

NTA would not respond to low-energy neutrons (Sommers 1975).
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 23 of 59

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 results were 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 INL (Hankins 1973), TLDs with 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 (in millirem but not dose equivalent) with 

the detector in the 3-in.-diameter PNR-4 insert. A plot of FNCF versus 9- to 3-in. ratio is used to 

determine the FNCF from the measured ratio (Hankins 1976). Values of the FNCF as shown in Table 

6-5 (Cusimano 1981) were measured for different fields at INL, were tabulated for assigning the dose 

equivalent from the badge results, and were routinely updated. 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. 1981 FNCFs (Cusimano 1981).

a

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

a. ZPPR = Zero Power Plutonium (later Physics) Reactor.

The date of the change from NTA to albedo neutron monitoring is somewhat in dispute. Different 

organizations would typically have changed to new monitoring systems at different times. The 

present record suggests the switch occurred near the end of 1974 or early 1975 (Ruhter and Perry 

2002; Gesell et al. 1996), but an informal list of “Dosimetry Branch Changes” from 1978 (INEL ca. 

1978) states, “initial testing of albedo neutron dosimeter and replacement of NTA neutron monitoring 

film with same,” in October 1976. Aoki (1979) said the albedo system replaced the NTA badge in 

1977. Dose reconstructions should make the assumption that this transition occurred on October 1, 

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 that were 5 mm thick 

were used for beta calibrations. Cesium-137 was considered for a calibration source in 1959 (AEC 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 24 of 59

1960, p. 83) and was installed in the instrument calibration facility in 1961 (Horan 1962). An 

automatic badge irradiator that was developed in the 1960s (Cipperley 1966) did not use a phantom 

to provide backscatter.

Table 6-6. Laboratory sources of uncertainty for beta/photon dosimeter calibration parameters [3]. 

Parameter Historical Description Uncertaintya Comment

In-air 

calibration

In 1981, INL 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, INL dosimeter 

systems were typically calibrated 

to a photon beam measured as 

exposure.

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

gamma radiation used to calibrate 

dosimeters, this caused a slight (about 3%) 

under-response in recorded dose. 

Tissue depth 

of dose

Historically, INL used an 

unspecified depth to estimate the 

deep dose.

±5% The numerical effect of this for photon 

radiation is comparatively low. INL

dosimeter designs had filtration density 

thickness of about 1,000 mg/cm2 that would 

relate closely to the 1-cm depth in tissue.

Angular 

response

INL dosimeter system is 

calibrated using anterior-posterior 

(AP) laboratory irradiations.

>300 keV, 

~20%

Recorded dose of record likely too low 

because the dosimeter response is usually 

lower at non-AP angles. Effect is highly 

dependent on dosimeter type, radiation 

type, energy, and angle. 

Environmental 

stability

INL 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 an area of 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 after the Three Mile Island-2 reactor accident in 1978. TLD badges were calibrated to 

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

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

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

Separation of penetrating from nonpenetrating dose was an issue in 1957 (Bennett 1957) and 1976 

(Jenson 1976), particularly for ICPP where strong high-energy beta fields were not unusual. 

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 developed in terms of 

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

NVLAP (a precursor for DOELAP) standard, dosimeters for calibration were irradiated with 137Cs using 

a phantom backing. To convert from exposure in roentgen to dose equivalent index in rem, a 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 25 of 59

conversion factor Cx value of 1.08 was used (DOE 1981b). The current recommended Cx value of 

1.03 for 137Cs (DOE 1986a, Table 2) was used beginning in June 1981 (Gesell 1982b; Kalbeitzer 

1984).

In 1989, the INL dosimetry service transferred from the DOE Radiological and Environmental 

Sciences Laboratory (RESL) to EG&G Idaho, the prime contractor. Calibrations continued to use 

DOE RESL sources and no changes were made to the dosimetry system. The 1991 Tiger Team 

Review (DOE 1991) of the INL site indicated that use by the INL contractor and the Idaho Operations 

Office of the same sources for calibration led to a conflict of interest or an advantage in DOELAP 

tests. As a result, EG&G purchased a Shepherd panoramic irradiator with a 137Cs source for badge 

irradiations. This irradiator did not use a phantom; it was cross-referenced using many TLD 

irradiations to the DOE source using a phantom (Andersen 1995). In addition, the contractor 

developed and characterized a uranium slab for beta irradiations (Bean 1995).

6.3.3.2 Neutron Calibration

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

dosimeters based on comparison of the recorded dose with 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 Weiss 1982). The yield of 

the AmBe source should be only about 3% larger than that for the PoBe source (Anderson and Hertz 

1971). Kluge and Weiss (1982) calculate conversion factors of 3.51 to 3.76 × 10-8 rem-cm/n

depending on the particular measure of dose equivalent chosen. IAEA (1988) provided 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), which corresponds to a dose 

conversion factor of 4.17 × 10-8 rem-cm2

/n, so the recorded dose is 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). Use of the 50-Ci AmBe 

source continued until 1993.

In 1993, a 40- by 40- by 15-cm polymethyl methacrylate 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).
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 26 of 59

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

[4].

a

Parameter Historical Description Anticipated laboratory biasb

Source energy 

spectrum

In 1976, INL began using dosimeters that 

were calibrated on a phantom to simulate a 

worker’s body. The previous calibrations 

did not include response from 

backscattered radiation. 

NTA film tends to be insensitive to albedo 

neutrons, so this probably had minimal effect.

Radiation 

quantity

Neutron dose quantities that were used to 

calibrate INL neutron dosimeters have 

varied historically. The first collision dose 

for fast neutrons and a quality factor of 10 

were used for many years. 

As noted above, NTA calibration would result 

in the reported dose being about 11% high. 

The effects of the respective neutron dose 

quantities used to calibrate INL dosimeters is 

uncertain and could be evaluated in 

comparison to the Hp(10) dose used in 

DOELAP performance testing. 

Angular response INL dosimeters are calibrated using AP 

laboratory irradiation.

Recorded dose of record is likely too low 

because the dosimeter response is lower at 

non-AP angles. The effect is highly 

dependent on neutron energy.

Environmental 

stability

INL NTA film dosimeters and TLDs are 

subject of signal fade with time, heat, 

humidity, light, etc.

Recorded dose of record is likely too low; 

however, this depends strongly upon when 

the calibration dosimeters are irradiated 

during the dosimeter exchange cycle. 

Midcycle calibration minimizes the effects. 

a. Judgment based on INL dosimeter response characteristics.

b. Recorded dose compared to Hp(10).

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

241Am-Be(α,n) source (Kluge and Weiss 1982). 

6.3.4 

Radiation fields varied a great deal at INL as discussed in the Site Description TBD (ORAUT 2005a). 

At times, very high radiation fields existed, particularly at ICPP, and work was sometimes done in 

those fields under tight controls. 

Workplace Radiation Fields
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 27 of 59

The epidemiological study of INL (Schubauer-Berigan et al. 2005) provides the excellent summary of 

radiation exposure history at INL in Table 6-8. The table shows the number of people monitored in 

Table 6-8. Photon dose (rem) percentiles for monitored workers (Schubauer-Berigan et al. 2005). 

Year 0% 25% 50% 75% 90% 97.5% 99.5% 100% Mean St. Dev. Number

1951 0 0 0 0 0 0 0.4 0.4 0.0022 0.0292 188

1952 0 0 0 0 0.03 0.222 0.727 1.275 0.0216 0.1002 913

1953 0 0 0.06 0.25 0.56 1.09 2.097 7.57 0.2002 0.4034 1,408

1954 0 0 0 0.21 0.9 2.498 4.077 5.76 0.2838 0.6567 2,449

1955 0 0 0.1 0.59 1.6565 3.237 5.162 8.47 0.5157 0.9314 2,946

1956 0 0 0 0.2 1.015 2.555 4.267 22.06 0.3142 0.8423 3,209

1957 0 0 0 0.085 0.54 1.513 3.045 5.12 0.1706 0.4517 4,695

1958 0 0 0.04 0.325 1.26 2.89 4.378 10.51 0.3766 0.8015 5,079

1959 0 0 0.04 0.26 0.96 2.347 4.127 21.85 0.3099 0.7402 5,344

1960 0 0 0.05 0.255 1.075 2.647 3.734 5.01 0.3268 0.6755 5,827

1961 0 0 0.04 0.42 1.704 3.566 4.972 27.26 0.5063 1.2806 5,192

1962 0 0 0.03 0.175 1.115 3.225 5.048 9.885 0.3571 0.8942 5,339

1963 0 0 0.025 0.18 1.0795 2.954 4.022 5.1 0.318 0.7292 5,520

1964 0 0 0.01 0.215 1.2765 3.111 3.98 4.815 0.3538 0.7886 5,446

1965 0 0 0 0.43 2.2045 4.39 6.018 9.815 0.579 1.2107 5,520

1966 0 0 0.02 0.375 1.5595 3.467 4.466 6.045 0.4383 0.8926 5,180

1967 0 0 0 0.17 1.0725 3.084 4.377 4.805 0.3194 0.7702 6,304

1968 0 0 0 0.215 1.14 3.104 4.194 5.295 0.3364 0.7781 4,922

1969 0 0 0 0.24 1.195 2.73 3.982 4.45 0.3279 0.7151 4,758

1970 0 0 0 0.175 1.028 2.624 4.207 4.68 0.2952 0.702 5,051

1971 0 0 0 0.17 0.82 1.899 3.252 4.71 0.2357 0.5427 4,764

1972 0 0 0 0.161 0.855 2.375 3.835 4.665 0.2606 0.634 4,762

1973 0 0 0 0.115 0.6525 2.126 3.909 5.2 0.2185 0.5878 4,494

1974 0 0 0 0.1 0.515 1.696 2.993 4.065 0.1734 0.4574 4,878

1975 0 0 0.006 0.09 0.405 1.412 2.789 3.945 0.1531 0.4071 5,025

1976 0 0 0 0.11 0.506 1.646 2.643 4.145 0.1712 0.4294 5,489

1977 0 0 0 0.094 0.485 1.947 3.385 10.77 0.1869 0.526 5,677

1978 0 0 0.009 0.087 0.394 1.535 2.872 4.386 0.1563 0.4295 6,551

1979 0 0 0 0.074 0.369 1.46 2.732 4.18 0.1419 0.4064 6,863

1980 0 0 0 0.047 0.277 1.062 2.112 16.93 0.1073 0.3579 7,380

1981 0 0 0 0.046 0.252 0.824 1.674 3.289 0.0876 0.2358 6,722

1982 0 0 0 0.038 0.195 0.638 1.5 2.904 0.0715 0.204 6,556

1983 0 0 0 0.035 0.187 0.516 0.976 1.577 0.0582 0.1473 6,610

1984 0 0 0 0.034 0.179 0.577 1.198 2.285 0.0619 0.1766 7,476

1985 0 0 0 0.039 0.219 0.872 1.24 2.415 0.082 0.226 7,917

1986 0 0 0 0.024 0.192 0.773 1.749 9.338 0.077 0.2623 8,568

1987 0 0 0 0.025 0.155 0.733 1.417 3.158 0.0659 0.2098 8,575

1988 0 0 0 0.022 0.145 0.537 1.086 3.086 0.0545 0.1641 8,667

1989 0 0 0 0.016 0.101 0.559 1.315 7.811 0.0516 0.2169 8,848

1990 0 0 0 0.012 0.0954 0.566 1.266 2.728 0.049 0.1852 10,165

1991 0 0 0 0 0.037 0.249 0.694 4.577 0.0232 0.1094 10,742

1992 0 0 0 0 0.037 0.169 0.376 1.276 0.0157 0.0563 9,571

1993 0 0 0 0 0.065 0.342 1.06 1.535 0.0311 0.1233 9,048

1994 0 0 0 0.01 0.0786 0.335 0.729 1.394 0.0322 0.1121 8,473

1995 0 0 0 0.011 0.1031 0.449 1.153 1.844 0.0428 0.1533 7,818

1996 0 0 0 0.016 0.108 0.421 0.79 1.368 0.0393 0.1215 6,459

1997 0 0 0 0.01 0.081 0.259 0.574 1.108 0.0261 0.082 6,280

1998 0 0 0 0.002 0.062 0.208 0.462 0.844 0.0205 0.065 5,875
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 28 of 59

the last column, the mean dose received (1 mSv = 100 mrem), and doses at several percentages on 

the distribution of doses received. For example, in 1952 90% of the 913 monitored workers had 

doses less than 30 mrem and 99.5% of the workers had doses less than 727 mrem. The mean dose 

was 2.2 mrem, and the standard deviation of dose was 100.2 mrem. 

Figure 6-9 shows the collective dose received by all workers through 1990 from Horan and Braun 

(1993). The highest collective dose was 3,448 rem in 1965.

Figure 6-9. Collective dose for INL personnel, 1951 to 1990 (Horan 

and Braun 1993). 

6.3.4.1 Gamma Radiation

In response to a Tiger Team finding (DOE 1991), radiation fields at INL have been characterized by 

comparing field measurements with a NaI(Tl) gamma spectrometer and TLDs mounted on a phantom 

(Reilly 1998). Figure 6-10 shows the percentage bias for the beta and gamma measurements. Most 

results lie between +27% and –43%. The high gamma-bias results are for locations at RWMC looking

at skyshine (back-scattered, low-energy photons) from low-level waste in the Subsurface Disposal 

Area. 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 were likely to have been missed. 

The radiation fields at INL, with a few exceptions, have been generated primarily by mixed fission and 

activation products. Therefore, most of the photon dose has been from photons with energy greater 

than 250 keV [5]. The INL dosimeters are judged to measure these fields well [6]. 

The few exceptions are usually characterized by low dose rates. Much of the waste at the RWMC is 

transuranic waste from the Rocky Flats Plant. This contains predominantly plutonium and americium 

with 17- and 59-keV photons, which are highly adsorbed by the waste and the shipping container. 

The dominant fields at the RWMC come from mixed fission and activation products dominated by 

137Cs and 60Co [7]. 

Analytical X-ray generators operating below 100 keV are used in several laboratories. These are 

easily shielded so the fields are usually low [8]. 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 29 of 59

There are a few 250-keV X-ray generators used for radiography or radiography development studies. 

Wall shielding is generally adequate, and any transmitted photons have energy near the operating 

voltage because of the hardening caused by the shielding [9]. The radiography facility at TRA has no 

roof shielding, so nearby dose rates of less than 10 mrem/hr are possible, usually for short periods. 

The scattered photons have lower energy than the primary beam. The people likely to be exposed 

are radiographers who receive fairly high exposures, mostly from the ambient radiation fields at their 

worksites and some from 300-keV 192Ir or 1.25-MeV 60Co [10]. 

 

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 6-10. Gamma and beta radiation field characterization (Reilly 1998). 

All INL radiological work areas essentially involved beta and photon radiation covering a wide range of 

energies. These fields can be generally classified according to the IREP codes in Table 6-9. 

6.3.4.2 Beta Radiation

Beta radiation fields are usually associated with activation or fission products outside a container such 

as in spills or hot cells or when only lightly shielded [12]. High beta fields were not unusual at the 

Idaho Nuclear Technology and Engineering Center (INTEC; previously ICPP) where there are large 

quantities of fission products. Pure high-energy beta fields in some locations, particularly in the 

exhaust stream, have caused dosimetry problems because the badge shielding or instrument 

packages did not provide a full 10-mm tissue-equivalent coverage and the beta fields would therefore 

be measured as gamma fields (Black 1974; Jenson 1976; INEL ca. 1978). 

The high bias beta results in Figure 6-10 from comparison of TLDs to a phoswich beta spectrometer 

are for sources at contact or at 1 cm, which results in hard-to-reproduce geometry. The low-bias beta 

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

beta occupational radiation fields (only three) have a bias of more than 15% (Reilly 1998). 

Beta field dosimetry became reasonably 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 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 30 of 59

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

Table 6-9. Selection of IREP beta and photon energies for INL facilities [11].a

Process/

buildings

Description Radiation 

type

Energy 

Operations Begin End group (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 might be 

significant higher energy beta radiation

Beta >15 100

Photon 30–250 25

>250 75

PBF, TRA, ARA, TAN, EBR, ANL-W, SPERT 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 1952 2003

Calibrations Calibration of instruments and dosimeters. Beta >15 100

CF 633, 636 1952 2003 Photon 30–250 25

>250 75

Waste 

handling

Radiation characteristics are highly dependent on source of waste, 

but typically fission product nuclides (Sr/Y-90, Cs-137) are 

dominant. Transuranic waste from Rocky Flats Plant contains 

Am-241 with 59-keV photon.

Beta >15 100

Photon 30–250 25

RWMC, WERF 1953 2003 >250 75

Uranium 

handling

Produced special armor from depleted uranium. Primarily beta 

radiation from U-238 daughters. Some gamma from contaminants 

and Cs-137 sources used in process.

Beta >15 100

Photon 30–250 90

>250 10

SMC 1985 2003

a. PBF = Power Burst Facility; SMC = Special Manufacturing Capability.

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

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

abscissa is plotted on the ordinate in Figure 6-11. Beta-emitting nuclides varied from location to 

location and time to time at INL, so a correction factor common for all facilities was estimated [13]. 

This analysis used the entire mixture of radionuclides to obviate whether the choice is correct and to 

reflect the wide variety of radionuclides used at INL. 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 or 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 thicker, more sensitive elements failed to report the beta 

dose correctly at a depth of 7 mg/cm2

, which was chosen in the early 1980s. These curves 

demonstrate why the beta dose that was assigned for skin is inappropriate to use for the breast and 

testes, where much of the organ is at a depth greater than 1 mm or 100 mg/cm2

, and for most persons

at depths greater than 1 cm.

To calculate the fraction of dose missed by a dosimeter, the dose reconstructor needs only to average 

the appropriate curve of this nature over the depth of the active detector and compare it to the value 

at a depth of 7 mg/cm2

. The appropriate curve should be the curve of the mean range for the beta 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 31 of 59

spectrum and the angular distribution of the radiation exposure. To estimate this, 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 of that value for 45o incidence for the maximum 

energy [15]. The curves are the result of a polynomial trend line to the data, so averaging the fraction 

of radionuclides is relatively easy.

 

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 6-11. Distribution of beta ranges [14]. 

Table 6-10 provides the cover and detector thicknesses for the beta badges used at INL. 

Thicknesses with a “~” are estimated. The fraction of measured beta dose shown in Table 6-10 is the 

average as described above. To determine the corrected beta dose, the measured and missed 

nonpenetrating results from the dosimetry system should be multiplied by the values in the last 

column of Table 6-10. 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 [16]. For the Panasonic system, such a correction has already been made (Gesell 1986), 

so the recommended correction is 1.

Table 6-10. Beta dosimeter thicknesses and associated underreporting [17]. 

Dosimeter 

system Period

Covers

(mg/cm2)

Detector thickness

(mg/cm2)

Beta Correction 

Factor

Two-filter film 1951–1958 ~50 ~50 2.0

Multifilter film 1958–1974 100 ~50 2.8

Low-dose TLD 1969–1974 100 75 3.0

ATLAS 1974–1975 100 ~100 3.3

Harshaw TLD 1975–1976 104 240 4.8

Harshaw TLD 1976–1985 4 240 2.4

Panasonic TLD 1986–2006 16 15 1.0
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 32 of 59

6.3.4.3 Neutron Radiation

Most INL workers have not been exposed to neutrons and so have not been badged to measure 

neutrons. Neutron fields at INL have been specific to a few facilities. The high-dose locations where 

most of the gamma and beta dose has been received, such as the ICPP and SL-1, have not had 

associated neutron dose. Table 6-11 lists the facilities and periods for which neutron exposure is 

considered likely.

Table 6-11. Facilities and periods for neutron exposure [18].a

Location Period Comments

Programmatic

TREAT 1958–1994

ZPPR 1969–1992 Between reactor halves

MTR 1952–1970 Research floor

TAN Warm Shop 1986–1988 See EPRI reports

TAN 1986–2006 Spent fuel storage pad

ANL Neutron 

Radiography Facility 

1977–2006

TRA Hot Cell Cave Cf-252 on filters

RWMC ILTSF Cargo container

RWMC WM 632 Transuranic waste drum

14 MeV

RWMC SWEPP 1990–2004 14 MeV for waste characterization

TAN Warm Shop 1991–1994 Refurbish 14 MeV for waste characterization

TRA 635 1990–2006 PINS Cf-252 6 and 14 MeV

Sealed sources

CPP-1649 1985–2006 PuBe Calibration Facility

CF-633 ~1970–2002 Cf-252 Calibration Facility

CF-636 1952–1994 AmBe Calibration Facility

IRC AmBe Cf-252

a. ILTSF = Intermediate-Level Transuranic Storage Facility; SWEPP = Stored Waste Examination Pilot 

Plant.

In 1969, 150 workers were involved in radiation work that required their NTA neutron dosimeters to be 

evaluated. At the time, there were 2,900 film-badged employees and more than 3,000 TLD-monitored 

personnel (Vallario, Hankins, and Unruh 1969).

In 1979, 5 people 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, only 

1,461 neutron dosimeters were issued (both monthly and quarterly badges) in comparison with about 

50,000 beta/gamma badges. Only 54 badges had reportable doses (≥15 mrem) as shown in 

Figure 6-12 (Gesell et al. 1996). Only six were above 35 mrem. The Hankins albedo dosimeter 

badges in use since 1975 are sensitive to all neutron fields. An FNCF determined from the 9- 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 that were mounted on a phantom 

and, at nearly the same time, with a ROSPEC neutron spectrometer (Reilly 1998). The relative 

biases [(Dosimeter – Spectrometer)/Spectrometer] for the neutron fields are shown in Figure 6-13. 

These results show a greater dispersion than the gamma results. The two lowest values (–52% and 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 33 of 59

–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 

of the radiation field. The next lowest value (–38%) is for the 252Cf instrument calibration source at a 

distance of 3.5 m, where the operator stands. The two highest values (94% and 71%) are for a waste 

drum that was reanalyzed and a new 9- to 3-in. ratio that was determined because of the 

unsatisfactory initial result. The report suggests that other waste barrels might have had neutron 

sources that caused interference. The remaining bias values lie between –16% and 44% (Reilly 

1998). 

Figure 6-12. Distribution of reportable neutron dose for the 

first 9 months of 1995. Of 1,461 dosimeters, 1,407 were 

below the 15-mrem reporting level (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 6-13. Neutron radiation field characterization (Reilly 1998). 

Total number of dosimeters with reportable dose = 54 Total number of dosimeters with reportable dose = 54
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 34 of 59

Figure 6-14 provides spectra for the 14-MeV neutron generator as seen through 10 cm of

polyethylene shielding typical of the INL facilities using the 252Cf (fission) neutron source (Ing and 

Makra 1978) and the AmBe neutron source (Kluge and Weiss 1982). Sources of neutron exposure 

include neutron sources at the instrument calibration laboratories and 14-MeV neutron generators 

used to characterize waste. For these spectra, the NTA film works reasonably well. Use of small 

252Cf sources for research began after albedo badge use. 

 

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 6-14. Neutron spectra that simulate INL facilities [19]. 

Most of the reactors built at INL had no beam ports. The neutrons were thus generally well contained 

away from the workplace. The 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

energy) are attenuated much more quickly in concrete or water than the gamma fields. This is not 

true for lead or iron, but these are usually not used as gamma shields where neutrons also exist. 

Neutron fields are thus generally not a problem at an enclosed reactor.

6.3.4.4 Materials Test Reactor Neutron Radiation

The exception to the above discussion is the Materials Test Reactor (MTR), which operated from 

1952 to 1970 and had beam ports and neutron beams extending onto a research floor. ZPPR and 

TREAT, both at ANL-W, are also in this category. Some neutron surveys of the MTR experimental

floor have been recovered (Sommers 1959, 1962; Hankins 1961), but these do not individually 

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 and measured the thermal neutron component at six other locations. The 

Hankins data have been reanalyzed (ORAUT 2006a) using more recent Bonner response curves 

(Hertel and Davidson 1985). Figure 6-15 shows the resultant neutron spectra for locations 3 and 23, 

which have higher doses and nearly the maximum low-energy intermediate and fission components, 

respectively. Figure 6-16 shows the correlations of the thermal and intermediate neutron dose 

equivalents to the fast neutron dose equivalent for the ORAUT (2006a) reanalysis of the Hankins 

data.

The trend line for the reanalyzed intermediate-energy neutron dose equivalent has R2 values of 0.86 

and 0.92 in comparison with an R2 value of 0.5 for the original analysis, which demonstrates a better 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 35 of 59

fit to the data. The average ratio of thermal to fast neutrons is 0.071 ±0.025, that for low-energy 

intermediate to fast neutrons is 0.177 ±0.057, and that for the higher energy intermediate to fast 

neutrons is 0.149 ±0.046, where fast neutrons are those above 0.2 MeV. 

The MTR personnel who were likely to receive neutron dose were assigned NTA film in their 

dosimetry packets, but it would have missed the dose 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 

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 6-15. Sample MTR spectra from 

Hankins Bonner measurements (ORAUT 

2006a). 

 

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 6-16. MTR neutron field components

(ORAUT 2006a). 

from 35% to 66% depending on the location (ORAUT 2006a). The dosimetry record location code for 

the TRA was 4 (later 40 to 45). To correct for missed dose on the MTR experiment floor, the NTA 

results from MTR should be multiplied by 2 ±0.3 (1/0.52, 0.08/0.522

) for a Monte Carlo dose 

reconstruction (ORAUT 2006a).

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 be not 

representative of the general workplace. Figure 6-17 shows the correlation of fast neutron dose 

equivalent to the gamma dose for these measurements. 

The fast neutron component was insignificant for several of the measurements. The values in Figure 

6-17 are shown with the triangles at one-half of the smallest measured value. Using the ShapiroWilks Normality Test (Gilbert 1987) and including the insignificant fast neutron values at one-half of 

the minimum reported value suggests that a 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. Combining these results as shown in Equation 6-1 yields a total neutron dose 

equivalent of 0.58 ±0.48 of the gamma dose equivalent on the MTR experimental floor [20]: 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 36 of 59

48.058.0)149.0177.0071.01)(35.042.0(

1( )

±=+++±=

= + ++

FastN

HiInter

FastN

LoInter

FastN

Thermal

GammaDose

FastN

GammaDose

TotalNeutronDose

(6-1)

The variations within the different components of neutron dose rate are so small in comparison with

the variation between the fast neutron and gamma dose equivalent rate that they are unimportant

[21]. 

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. For example, health physics technicians would 

 

 

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 6-17. Correlation of fast neutron dose 

equivalent to gamma dose at MTR (ORAUT 2006a). 

often have been covering jobs with only beta/gamma fields. A craftsperson might have serviced 

pumps that carried radioactive water and not received any neutron dose. Therefore, simply 

multiplying the gamma dose by 1.6 (i.e. average = 1+0.58) or 2.1 (i.e. 84% confidence = 1 + 0.58 +

0.48), although favorable to claimants, is probably inappropriate [22]. 

6.3.4.5 Test Area North Fuel Storage Casks

As noted in the INL Site Description TBD (ORAUT 2005a, Section 2.2.1.3), fuel storage casks are on 

a storage pad at TAN. The dose rates are 25 to 30 mrem/hr gamma and 40 mrem/hr neutron. The 

metal cask attenuates the gamma radiation, but does not appreciably affect the neutron field. The 

loaded casks were temperature-tested in the TAN Warm Shop in 1985 and 1986. The shop has 

offices nearby at one end on the second floor. Neutron radiation levels were discovered in the offices, 

and the people (about six) in these offices were not wearing albedo neutron badges but were wearing 

beta/gamma dosimetry. Each of the three casks was in the area about 2 weeks while temperature 

measurements were made before they were moved outside onto the storage pad. The casks contain 

irradiated oxide fuel in which the 18O(α,n) reaction generates neutrons in the range of a few 

megaelectron-volts. This radiation was attenuated by distance and the building concrete between the 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 37 of 59

cask and the offices. A TLD area-monitoring albedo system identified the radiation field. Based on 

Electric Power Research Institute (EPRI) documents (EPRI 1986, 1987; PNL, VPC, and EG&G 1987), 

the temperatures were measured from September 11 to 23, 1985; January 14 to February 6, 1986; 

and June 2 to 27, 1986. The person most affected was in operations. Safety and radiological 

engineers were also affected. The estimated dose equivalent for full-time occupancy is less than 

50 mrem for each of the three cask evolutions. The records indicate that manual additions to the 

dosimetry records were made at the time for neutron dose based on area-monitoring TLDs [23].

6.3.4.6 Typical Workplace Neutron Dosimeter Hp(10) Performance

Table 6-12 summarizes typical neutron personnel dosimeter parameters important to Hp(10)

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

neutron dosimeters is the difference between calibration and workplace neutron energy spectra [24]. 

Table 6-12. Typical workplace neutron dosimeter Hp(10) performance [25].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 

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 were recorded 

as zero dose.

Recorded dose of record is likely too low. 

The impact of missed dose is greatest in 

earlier years because of the higher MRLs 

and shorter exchange cycles of the 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 INL dosimeter response characteristics.

b. Recorded dose compared to Hp(10). 

6.4 ADJUSTMENTS TO RECORDED DOSE 

6.4.1 

All measured and reported dose equivalents from INL in this document used the quality factors based 

on the LET of the ionizing secondary particles that was established in the 1950s and used since by 

U.S. regulatory agencies [26]. 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 1991, Table 1). 

Neutron Weighting Factor

The reported INL data require correction to change from dose equivalent (pre-ICRP Publication 60) to 

the newer dose quantity (ICRP 1991; NIOSH 2006). ICRP Publication 74 tabulates the ambient dose 

equivalent (dose equivalent at 10-mm depth in a 30-cm-diameter sphere) for neutrons (ICRP 1996). 

The ratios of organ to ambient dose equivalents are tabulated in Appendix B of NIOSH (2006), so this 

quantity is 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 those in other TBDs.
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 38 of 59

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 considered range of energies:

H DECF dEEE

E

E

∫ =

2

1

φ )()( (6-2)

Error! Bookmark not defined.

For ambient dose equivalent, the same equation applies except that 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 to calculate 

the correction factors [28]. 

These factors are incorporated into statements of dose equivalent 

values and calibrations that follow 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). The primary geometry is conventionally

considered 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 due to 

attenuation in the human body. Dosimeters are designed to respond to radiation entering the body on 

the side where they are located, and they work best for an AP irradiation geometry with the dosimeter 

on the front of the body [27]. 

Table 6-11 summarizes the locations at INL where neutron dose is credible. Table 6-13 lists the 

calculated fractions of dose equivalent in the IREP energy groups and the conversion factors from 

Table 6-13. Calculated and recommended dose equivalent fractions and quality factor corrections

[29]. 

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 polyethylene 2.4E-08 3.1E-06 1.5E-03 1.00

MTR experiment floor average 0.18 0.06 0.49 0.28

MTR experiment floor maximum 0.24 0.08 0.52 0.35

MTR experiment floor minimum 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 polyethylene 0.69 1.47 1.36 0.93

MTR experiment floor average 0.86 1.08 1.33 1.12

MTR experiment floor maximum 0.80 1.08 1.37 1.12

MTR experiment floor minimum 0.92 1.08 1.30 1.12

Recommended defaults

Dose equivalent fractions

14 MeV 10 cm polyethylene 0.05 0.95

Source calibrations 0.20 0.80

MTR experiment floor 0.2 0.05 0.50 0.25

H*(10)/H 1 1.1 1.4 1.1
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 39 of 59

dose equivalent to equivalent dose for INL 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 [30]. 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 6-14.

Table 6-14. Recommended IREP neutron energy fractions and correction factors [31]. 

Process Description Operations

Neutron 

energy

Default 

dose 

(%)

Ambient 

dose equiv/ 

dose equiv

Net 

correction 

factor

Instrument 

calibration

Alpha Be source 

calibrations

1951 1993 0.1-2 MeV 20 1.4 0.28

Cf-252 source 

calibrations

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 

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

6.5 MISSED DOSE

6.5.1 

Workers have reported that, on some occasions, they did not wear dosimeters while working in 

radiation areas (Wages et al. 1998; ORAUT 2004b). The latest revision of Use of Coworker 

Dosimetry Data for External Dose Assignment (ORAUT 2005b) provides guidance to assist dose 

reconstructors in the evaluation of the support for a claimant’s allegation that he or she did not wear a 

dosimetry badge at all times. 

Dosimeter Not Worn

6.5.2 

Missed photon dose for INL workers would occur when 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 [32]. The missed dose is calculated as described in 

NIOSH (2006) using MRL/2 multiplied by the number of zero dose results. Table 6-15 lists the 

potential missed photon doses 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

6.5.3 

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

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

Missed Beta Dose
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 40 of 59

is the difference between the shallow and deep doses. All nonpenetrating dose for INL 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 is generally 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 6-10 for the appropriate period.

Table 6-15. Beta/photon dosimeter period of use, type, MRL, exchange frequency, and potential 

annual missed dose [33]. 

Period of usea Dosimeter

Exchange 

frequency

MRLb (mrem)

Annual missed dose 

(mrem)c

Photon Beta Photon Beta

August 1951–

March 1958

INL 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

INL multielement DuPont 508 

film

Weekly (n=52) 10 30 260 780

Biweekly (n=26) 130 390

Monthly (n=12) 60 180

December 

1966–February 

1974

INL multielement DuPont 508 

film

Weekly (n=52) 10 30 260 780

Biweekly (n=26) 130 390

Monthly (n=12) 60 180

INL 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

INL 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 

INL 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

INL 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, INL workers had a dosimeter assigned to each operating area where they worked, or they were issued 

visitor dosimetry. All area dosimetry was issued beginning in January 2000.

b. MRLs are based on Cipperley (1958, 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 (2006). 

d. ICPP began using the Harshaw TLD in December 1974, the prime contractor began in February 1975, and ANL-W 

began 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.

6.5.4 

Neutron radiation was present at the INL reactors, the 14-MeV neutron generators at RWMC, for a 

short time at TRA and TAN, in small sources used for research at TRA and the INL Research Center, 

in calibration laboratories CF633 and CF636, and for calibration of criticality alarms at TRA and ICPP

[34]. For other locations, there is likely limited missed neutron dose because of the very low potential 

for neutron exposure. To calculate the missed dose, the dose reconstructor must first determine if the 

Missed Neutron Dose
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 41 of 59

person worked near neutrons and the category of neutrons. This can best be accomplished by 

looking for the work location and at whether a coworker or others in the badge reporting group were 

assigned a neutron dose equivalent. The work location code for TRA where the MTR operated is 4 

(later 40 to 45). If no neutron dose was assigned to the worker or coworkers for several months, the 

dose reconstructor should assume that the person was not exposed to neutrons [35]. 

If a worker was likely to have been exposed to neutrons, the dose reconstructor should assign missed 

neutron dose equivalent using Table 6-16 for the times when workers did not have reported neutron 

dose [36]. For the period when NTA film was used, the missed neutron dose should be multiplied by 

1.25 for all facilities except the MTR experimental floor and by 2 for the MTR experimental floor when 

the MTR was operating between 1953 and 1970 [37]. The dose equivalent is then apportioned into 

the IREP groups using Table 6-16. 

Table 6-16. Neutron dosimeter type, period of use, exchange frequency, laboratory minimum 

detectable limit, and maximum annual missed dose [38]. 

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

For example, if in 1955 a person was an experimenter at the MTR, and seven of the weekly badges 

recorded a total of 185 mrem neutron dose equivalent, then the missed dose would be 315 mrem 

[(52 – 7) × 14 ÷ 2] so the total dose by the badges would be 500 mrem. Because the badge only sees 

about one-half of the MTR neutron dose equivalent (see Section 6.3.4.4), the total dose equivalent is 

1 rem. To convert the 1 rem received from neutrons on the MTR experimental floor to equivalent 

dose, the total dose equivalent is multiplied by the last column of Table 6-14, which results in 

200 mrem to the <10-keV group, 60 mrem to the 10- to 100-keV group, 700 mrem to the 0.1- to 

2-MeV group, and 280 mrem to the above-2-MeV group for a total equivalent dose of 1.24 rem. 

The neutron missed dose is divided into two historical periods in the following discussion. The first is 

before October 1976 when only NTA film dosimeters were used with supplemental recording of 

thermal neutron doses from B-10 pencil dosimeters [39]. The second period is from and including 

October 1976 to the present when only Hankins albedo dosimeters were used [40]. Table 6-16 

summarizes the estimated MRLs for these neutron dosimeters. It is possible to estimate the missed 

neutron dose using the MRLs because the neutron dosimeters were calibrated with neutron sources 

of energies similar to those in the workplace and because most of the neutrons to which workers were 

normally exposed had energies greater than the 500- to 800-keV threshold of the NTA film dosimeters

[41]. There was, of course, no threshold energy for the measurements using neutron albedo TLD 

badges.

6.5.4.1 Before October 1976

The use of NTA films for neutron dosimetry before 1976 is documented in INL 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 INL workers for two reasons: 

(1) An NTA film was developed and not read in accordance with the standard criteria, or (2) an NTA 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 42 of 59

film indicated a neutron dose equivalent that was less than the film’s 14-mrem MRL [42]. When the 

MRL for NTA film is used to estimate the missed neutron dose, it should be multiplied by 1.25 for most 

workers and by 2 for workers on the MTR experimental floor [43]. 

An estimate of the missed neutron dose in some facilities might appear to be attainable through use of 

neutron-to-photon dose ratios (NIOSH 2006). However, for the INL facilities there are several other 

sources of gamma exposure with no associated neutron exposure, so that approach would be 

erroneous.

6.5.4.2 October 1976 to Present

Since and including October 1976, the neutron dose has been measured using the Hankins albedo 

TLD. The characteristics of this dosimeter are well documented (Gesell et al. 1996), and the MRL to 

be used to estimate missed dose is 15 mrem. A location-specific FNCF has been applied to convert 

the reading to dose equivalent, so no additional adjustments should be required (Gesell et al. 1996). 

6.6 ORGAN DOSE

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

organ doses of interest using the external dose reconstruction guideline (NIOSH 2006). It is 

recommended that the AP geometry should be assumed for the irradiation geometry and for 

conversion to organ dose [44]. 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 [45]. For 1981 and after, the conversion factor from deep dose equivalent 

to organ dose should be used [46]. 

6.7 UNCERTAINTY

Measurement uncertainties arise from many sources. A lognormal distribution should generally be 

assumed unless stated otherwise. For gamma rays, the standards for exposure have existed with 

only minor changes since the 1930s as required for medical uses of radiation. INL used ionization 

chambers standardized by NBS for its calibrations. Use of a phantom for dosimeter irradiation began 

in the early 1980s, but backscatter only causes a minor change for high-energy photon dosimetry [47]. 

The over-response of the multielement film badge to deep dose in tissue is due to calibration to 

exposure, which is somewhat greater at low energies than for the deep dose [48]. The INL

environment did not have a significant low-energy photon field such as 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 [49]. This is roughly consistent with the 

results in Figure 6-10. For those measurements, the standardization instrument contributed some 

significant uncertainty.

For beta radiations in relation to skin cancer, etc., the reported nonpenetrating dose must be divided 

by the fraction of measured dose in Table 6-12. The uncertainty for beta radiation is somewhat larger 

at an estimated 50% at 1 sigma [50]. This is driven by uncertainties in the field geometry and 

because beta radiation is often stopped by air and thin materials such as clothing [51]. 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 [52]. The difference between point- and 

planar-source irradiation can confuse an algorithm [53]. Earlier techniques did not provide a thin 

detector with minimal covering, which is important for simulating the skin for beta dosimetry.
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 43 of 59

For neutron radiations, the situation is more complex. The NTA films in use before 1975 did not react 

to low-energy neutrons below 0.5 to 0.8 MeV (Cusimano 1963, Cipperley 1968). Corrections are 

described for handling this issue. The TLD albedo system provides a very indirect way of measuring 

dose equivalent to a person. Dose to workers is primarily due to hydrogen recoils rather than the 

6

Li(n,α) reactions [54]. The response of the 9-in. PNR-4 detector used to standardize the TLD 

measurements is also due to a different process than dose deposition in the human body. The total 

uncertainty for neutrons is probably a geometric standard deviation (GSD) of about 1.6 [55]. The 

uncertainty of the base curve for calculating the FNCF is a GSD of about 1.35 (Gesell et al. 1996) 

although this uncertainty does not contribute in the DOELAP process, which results in a total GSD of 

1.7 for albedo neutron measurements [56]. 

The cause of the greatest uncertainty for neutrons is the variation of dose caused by organ positions 

in the body. For 1-MeV neutrons, the dose facing the source is about a factor of 1,000 higher than the 

dose on the back side of a 30-cm-diameter sphere of tissue-equivalent material [57]. In a work 

environment, the primary 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, it is assumed in EEOICPA that the worker irradiation is in an AP geometry (from the front) [58]. 

Note that in Figure 6-12 the discrepancy in which the dosimeters report about one-half of the 

spectrometer result is because the spectrometer does not simulate the attenuation of the body, so it 

reads high by a factor of two.

6.8 ATTRIBUTIONS AND ANNOTATIONS 

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

information, conclusions, and recommendations to assist in the process of worker dose 

reconstruction. These callouts are listed in this section with information that identifies the source and 

justification for each item. Conventional references are provided in the next section that link data, 

quotations, and other information to documents available for review on the Oak Ridge Associated 

Universities (ORAU) Team servers.

Norm Rohrig served as the initial Document Owner of this document. Mr. Rohrig was previously 

employed at the INL site, where 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. 

Because gammas interact primarily with electrons and they are only weakly attenuated, 

different definitions of dose and different measurement techniques have little effect.

[2] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This statement generally applied to high-level supervisors such as DOE officials and company 

managers who occasionally came to the site for meetings.

[3] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This table is a requirement from ORAUT-PROC-0031, “DOE Technical Basis Document
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 44 of 59

Development, Review, and Approval Process” and uses values based on experience and from 

the Hanford Site Profile (ORAUT 2006b). 

[4] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This table is adapted to INL from the Hanford Site Profile (ORAUT 2006b) and uses values 

based on experience at the INL.

 [5] 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- to 250-keV band was 1.7% to 5.1% of the total dose. Monte Carlo 

calculations were performed using a code provided by S. Cohen & Associates, and the dose 

contribution from 0- to 250-keV photons through a concrete wall was 0.4% to 7.4%. 

Combining these effects shows that the 25%/75% split is appropriate.

[6] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

Photon dosimetry for higher energy photons is usually straightforward because of the minimal 

attenuation and because the primary interaction is by the Compton process with electrons.

[7] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

See the Site Description TBD (ORAUT 2005a). 

[8] Rohrig, Norman D. ORAU Team. Health Physicist. Date.

The contact radiation field on these devices is less than 0.5 mrem/hr (phone conversation 

Larry Burke, Radiological Engineer, ANL-W and INL, 11/06/2006, Rohrig 2006a).

[9] Rohrig, Norman D. ORAU Team. Health Physicist. November 2006. 

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 lowenergy photons) results.

[10] Rohrig, Norman. ORAU Team. Health Physicist. June-October 2003. 

Radiographer exposure is generally to the radiography sources identified and to radiation

surrounding the items being radiographed. 

[11] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This table is an EEOICPA requirement and uses values based on experience and cited in the 

text. Also see item [5] above.

[12] Rohrig, Norman. 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. 

[13] 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.

[14] 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 plot.

[15] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 45 of 59

This choice of angles and energies was used to generate a simplified depth-to-dose curve.

[16] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

 This correction factor is unlikely to have been applied during calibration. The net correction is 

the ratio of corrections for the calibration spectrum and the field spectrum, so the net effect is

smaller than the corrections in Table 6-9.

[17] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

Dosimeter thicknesses have been stated in previous paragraphs. The correction factors are 

derived based on the described theoretical approach. 

[18] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This table is based on information in the Site Description TBD (ORAUT 2005a) and work 

experience at the facility. 

[19] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This plot was generated from previously cited spectra (Ing and Makra 1978; Kluge and Weiss

1982).

[20] 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 ORAUT-OTIB-0009 for the different ratios (ORAUT 2006a). 

[21] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

The fractional error on gamma-to-fast-neutron ratio (83%) dominates that from the errors that 

are stated in the second paragraph of Section 6.3.4.4 of 1.8%, 4.1%, and 3.3%. 

[22] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

Based on the judgment 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.

[23] Rohrig, Norman D. ORAU Team. Health Physicist. February 20, 2007. 

Based on conversation on October 12, 2006 with a former health physicist who was in charge 

at TAN in 1986 and was in charge of INL dosimetry from about 1992 to 2006 (Rohrig 2006b). 

[24] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This subjective statement is 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.

[25] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This standard TBD table is based on information in the Hanford TBD (ORAUT 2005a) and, in 

the judgment of Mr. Rohrig, performance of INL dosimeters for similar parameters of 

performance. 

[26] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This is based on guidance in OTIB-0055 (ORAUT 2006c). 

[27] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 46 of 59

For neutrons, shielding by a human is significant (factor of 1,000 for 1-MeV neutrons), and 

dosimeters thus need to face the radiation source to read reliably.

[28] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This is based on guidance in OTIB-0055 (ORAUT 2006c). 

[29] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This table was derived with spectral data in a spreadsheet as shown in Figures 6-13 and 6-14 

with the associated energy-dependent quality and dose effectiveness factors. The last four 

rows are averages of the associated calculated values higher in the table.

[30] 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.

[31] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This table is derived from the values in Table 6-12.

[32] 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. 

[33] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This table uses values in columns 3 to 5 that are previously cited in this report.

[34] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This statement is based on an understanding of the type and variety of work at INL.

[35] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003.

This statement is based on guidance provided in ORAUT-OTIB-0023, “Assignment of Missed 

Neutron Doses Based on Dosimeter Records.” 

[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. 

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.

 [39] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This reflects the period discussed in Section 6.3.2.7.

[40] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This reflects the period discussed in Section 6.3.2.8.

[41] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 47 of 59

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 would not 

contribute a major uncertainty. 

[42] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

By definition of the MRL, values less than the MRL are reported as zero (practice at INL) or a 

blank. 

[43] 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.

 [44] 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 

exclusive use of the AP geometry.

[45] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This recommendation is made because the measurement quantity of dose in that period was 

exposure.

[46] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This recommendation is made because the measurement quantity of dose in that period was 

deep dose equivalent.

[47] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

For high energy, the angular distribution is primarily forward. See for example Figure 2-19 of 

Knoll (1989).

[48] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

At low energies, air that has very little hydrogen in comparison to tissue has a higher crosssection than tissue because of the Z dependence of the photoelectric effect.

[49] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This estimate is partially based on Figure 6-9 and partially based on uncertainties of many 

error processes that contribute uncertainty.

[50] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This estimate is partially based on Figure 6-9 and partially based on uncertainties of many 

error processes that contribute uncertainty.

[51] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

The range of beta particles is shown in Figure 6-10.

[52] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

These estimated corrections are provided in Table 6-9, and they are not small.

[53] Rohrig, Norman. ORAU Team. Health Physicist. June-October 2003. 

For this reason, DOELAP has had different categories for point and slab beta geometries. 

[54] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 48 of 59

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.

[55] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

This reflects the spread in Figure 6-13 and the factors in Table 6-12. 

[56] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

Most of this uncertainty is from the FNCF for the field of the neutron exposure.

[57] Rohrig, Norman D. ORAU Team. Health Physicist. June-October 2003. 

See Figure 16 of NCRP Publication 38 (NCRP 1971b).

[58] 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 exclusive use of the AP geometry.
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 49 of 59

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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

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

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

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 unit is sievert, and 1 sievert equals 

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) 

This program accredits DOE site dosimetry programs based on performance testing and 

onsite reviews performed on a 2-year cycle. 

dosimeter

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

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

of absorbed dose or dose equivalent received by an individual. 

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

In the technical sense, the ionization produced by gamma or X-rays in air (roentgen); in the 

generic sense, ionizing radiation applied to matter. 

exposure-to-dose-equivalent conversion factor for photons (Cx) 

The ratio of exposure in air to the dose equivalent at a specified depth in a material of 

specified geometry and composition. The 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 ignoring trace elements, or soft tissue including trace elements).
Document No. ORAUT-TKBS-0007-6 Revision No. 02 Effective Date: 06/18/2007 Page 59 of 59

linear energy transfer (LET)

The lineal rate of local energy deposition by a charged particle.

minimum reporting level (MRL)

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

nonpenetrating dose

Dose from beta and lower energy photon radiation. Determined from the open window minus 

the shielded.

pencil dosimeters

A type of ionization chamber used by personnel to measure radiation dose. Other names: 

pencil, pocket dosimeter, pocket pencil, pocket ionization chamber.

penetrating dose equivalent

Photon dose measured by shielded INL film or elements plus neutron dose equivalent. 

Essentially, personal dose equivalent Hp(10). 

personal dose equivalent, Hp(d)

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

radiological protection purposes (ICRU 1993). The personal dose equivalent is represented 

by Hp(d), where d identifies the depth (in millimeters) from the point of reference for dose in 

tissue. For weakly penetrating radiation of significance to skin dose, d = 0.07 millimeters and 

is noted as Hp(0.07). For penetrating radiation of significance to whole-body dose, d = 

10 millimeters and is noted as Hp(10). 

polymethyl methacrylate

Scientific name for plastic commonly known as Lucite or Plexiglas.

range

The distance an energetic charged particle will go through a material before it stops. The 

range is an increasing function of energy and depends on the elemental makeup of the 

material and the density.

redacted

To select item(s) to be visible for viewing or for publication by obscuring others.

shallow absorbed dose (Ds)

The absorbed dose at a depth of 0.07 mm in a material of specified geometry and 

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 mm in tissue (sum of penetrating and nonpenetrating dose 

equivalent).

tissue rad

Absorbed dose in tissue.