ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 58

DOE Review Release 04/29/2011

Document Title:

Idaho National Laboratory and Argonne 

National Laboratory-West – Occupational 

External Dosimetry

Document Number: ORAUT-TKBS-0007-6 

Revision: 03

Effective Date: 04/19/2011

Type of Document: TBD

Supersedes: Revision 02 and ORAUT- 

 TKBS-0026-6 Revision 01

Subject Expert(s): Steven L. Bump

Approval: Signature on File Approval Date: 04/11/2011

JoAnn M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 04/11/2011

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 04/09/2011

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 04/08/2011

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/19/2011

Stuart L. Hinnefeld, Director, DCAS

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 2 of 58

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.

04/19/2011 03 Initiated to incorporate Argonne National Laboratory – West –

Occupational External Dose technical basis document (ORAUTTKBS-0026-6 Rev 01) into the Idaho National Laboratory 

Occupational External Dosimetry technical basis document. 

Incorporates formal internal and NIOSH review comments. 

Removed the Site Expert line from page 1 since the former Site 

Expert served no role in this revision. Updated Section 6.8 to reflect 

that in all cases where the former site expert’s information or prior 

studies or writings are included or relied on by the new Document 

Owner, those materials are fully attributed to the source in 

accordance with the “NIOSH Policy Statement: Management of 

Conflict or Bias in the Radiation Dose Reconstruction Program.” 

Constitutes a total rewrite of the document. Training required: As 

determined by the Objective Manager. Initiated by JoAnn M. 

Jenkins.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 3 of 58

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

6

6.1 Introduction ..............................................................................................................................9

6.1.1 Purpose......................................................................................................................10

6.1.2 Scope.........................................................................................................................10

6.2 Basis of Comparison..............................................................................................................10

6.3 Dose Reconstruction Parameters...........................................................................................11

6.3.1 Site Administrative Practices ...................................................................................... 11

6.3.2 Personnel Monitoring Systems ................................................................................... 18

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

6.3.2.2 Multiple-Filter NRTS Film Badge.................................................................. 19

6.3.2.3 Original Lithium Fluoride Teflon Thermoluminescent Dosimeter 

System......................................................................................................... 20

6.3.2.4 Automatic Thermoluminescent Analyzer System.......................................... 20

6.3.2.5 Harshaw Two-Chip Thermoluminescent Dosimeter System......................... 20

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

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

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

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

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

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

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

6.3.4.1 Penetrating Radiation................................................................................... 26

6.3.4.2 Non-Penetrating Radiation ........................................................................... 28

6.3.4.3 Neutron Radiation ........................................................................................ 30

6.3.4.3.1 Materials Test Reactor Neutron Radiation .................................. 33

6.3.4.3.2 Test Area North Fuel Storage Casks .......................................... 35

6.4 Adjustments to Recorded Dose..............................................................................................36

6.4.1 Photon Dose Adjustments .......................................................................................... 36

6.4.2 Electron Dose Adjustments ........................................................................................ 36

6.4.3 Neutron Dose Adjustments......................................................................................... 37

6.4.3.1 Neutron Energy Response Adjustments....................................................... 37

6.4.3.2 Radiation Weighting Factor Adjustments...................................................... 37

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

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

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

6.5.3 Missed Electron Dose................................................................................................. 40

6.5.4 Missed Neutron Dose ................................................................................................. 41

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
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 4 of 58

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

References ........................................................................................................................................48

Glossary ............................................................................................................................................55

LIST OF TABLES

TABLE TITLE

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

PAGE

14

6-2 Reasons codes ......................................................................................................................14

6-3 Irregularity codes ................................................................................................................... 14

6-4 Column 20 codes ................................................................................................................... 14

6-5 1981 FNCFs...........................................................................................................................22

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

parameters.............................................................................................................................23

6-7 Common sources of laboratory bias in the calibration parameters for 

neutron dosimeters ................................................................................................................24

6-8 Photon dose percentiles for monitored workers...................................................................... 26

6-9 Selection of IREP electron and photon energies for INL facilities ........................................... 28

6-10 Electron dosimeter thicknesses and associated under-reporting............................................ 30

6-11 Facilities and periods for neutron exposure............................................................................ 30

6-12 Electron dosimeter under-reporting correction factors............................................................ 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 Electron/photon dosimeter period of use, type, LOD, exchange frequency, 

and potential annual missed dose.......................................................................................... 39

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

minimum detectable limit, and maximum annual missed dose ............................................... 41
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 5 of 58

LIST OF FIGURES

FIGURE TITLE

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

PAGE

12

6-2 Film report form used in 1958 ................................................................................................ 13

6-3 Example of dosimeter reporting format used from March 1958 through 

November 1966 .....................................................................................................................13

6-4 Example of dosimeter reporting format used from December 1966–1985.............................. 15

6-5 Example of dosimeter reporting format used from 1986–1998 ............................................... 15

6-6 Example of dosimeter reporting format used from 1999–present ........................................... 15

6-7 Badge results from January 1961 for work in recovery from the SL-1 accident ...................... 16

6-8 Special badge report associated with a high beta reading in Figure 6-7 ................................. 17

6-9 Personnel exposure questionnaire partially completed for a hypothetical 

case .......................................................................................................................................17

6-10 Response of DuPont 508 film with various filters to 140 mrem from 

uranium beta and 100 mR of different energy photon irradiation ............................................ 19

6-11 Probability density of neutron spectrum from a 241Am-Be (α,n) source ................................... 25

6-12 Collective dose for INL personnel, 1951 to 1990.................................................................... 27

6-13 Distribution of electron ranges................................................................................................ 29

6-14 Distribution of reportable neutron dose for the first 9 months of 1995..................................... 31

6-15 Neutron radiation field characterization .................................................................................. 32

6-16 Neutron spectra that simulate INL facilities ............................................................................ 32

6-17 Sample MTR spectra from Hankins Bonner measurements ................................................... 33

6-18 MTR neutron field components .............................................................................................. 33

6-19 Correlation of fast neutron dose equivalent to gamma dose at MTR ...................................... 34

6-20 ZPPR Assembly-3 Phase-1A fast neutron spectrum measured midway 

between halves with both shields withdrawn .......................................................................... 35

6-21 ZPPR Assembly-3 Phase-1A fast neutron dose spectrum midway 

between halves with both shields withdrawn .......................................................................... 36
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 6 of 58

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

AX TAN Construction 

CF Central Facilities

CFA Central Facilities Area

CFR Code of Federal Regulations

Ci curie

cm centimeter

CPP Chemical Processing Plant

CX CPP construction

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 

EX EBR Construction

FNCF facility neutron correction factor

g gram

GSD geometric standard deviation

HP Health Physics

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

IRC INEL Research 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

LX Construction at LOFT

m meter
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 7 of 58

MeV megaelectron-volt, 1 million electron-volts

mg milligram

ML-1 Mobile Low Power Reactor

mm millimeter

mR milliroentgen

mrad millirad

mrem millirem

mrep millirep

mSv millisievert

MTR Materials Test Reactor

MX TRA Construction

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

OX Construction at OMRE

PBF Power Burst Facility

PEQ Personnel Exposure Questionnaire

PER Program Evaluation Report

PINS Portable Isotopic Neutron Source

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

SX SPERT Construction

TAN Test Area North

TBD technical basis document

TLD thermoluminescent dosimeter

TRA Test Reactor Area

TREAT Transient Reactor and Experiment Test

U.S.C. United States Code

WERF Waste Experimental Reduction Facility
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 8 of 58

wk week

WP Waste Processing Building

yr year

Z atomic number

ZPPR Zero Power Plutonium (later Physics) Reactor

α alpha particle

§ section
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 9 of 58

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

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

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

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

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

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

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

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

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

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

occupationally derived (NIOSH 2010):

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2010).

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 10 of 58

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, and the Idaho National Engineering and Environmental Laboratory 

(INEEL), and at the Argonne National Laboratory–West (ANL-W). For consistency, this document 

uses INL or ANL-W, except in special cases. 

Purpose

6.1.2

From the start of operations in 1951 until early 2005, ANL-W was operated by the University of 

Chicago under supervision of the Chicago Operations Office of the U.S. Atomic Energy Commission 

(AEC), the Energy Research and Development Administration (ERDA), and DOE. In 2005, the 

functions of ANL-W were combined with research activities at INEEL to form INL, and ANL-W was 

renamed the Materials Fuel Complex (MFC). A branch of the Idaho Operations Office provided 

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

1951 (when it was called the NRTS) 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 contractors 

changed often, the external dosimetry process has remained under the technical management of a 

single organization with responsibilities for dosimetry development, operational dosimetry, and 

radiological records, which has provided 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

Since the initiation of the Manhattan Engineer District project in the early 1940s, various radiation

dose concepts and quantities have been used to measure and record occupational dose. The basis

of comparison for reconstruction of dose is the personal dose equivalent, Hp(d), where d identifies the

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

radiation of significance to skin dose, d is 0.07 mm and the dose equivalent is noted as Hp(0.07). For

penetrating radiation of significance to whole-body dose, d is 10 mm and the dose equivalent is noted

as Hp(10). Both Hp(0.07) and Hp(10) are the radiation quantities recommended by the International

Commission on Radiation Units and Measurements (ICRU) for use as the operational quantities to be

recorded for radiological protection purposes (ICRU 1993). In addition, Hp(0.07) and Hp(10) are the

radiation quantities used since the 1980s in the U.S. Department of Energy (DOE) Laboratory

Accreditation Program (DOELAP) to accredit personnel dosimetry systems (DOE 1986a). The

International Agency for Research on Cancer selected Hp(10) as the quantity to assess error in 

historical recorded whole-body dose for workers in Agency nuclear worker epidemiologic studies

(Thierry-Chef et al. 2002). The basis for comparison for neutron radiation is more complicated

because the calibration of dosimeters to measure neutron dose was based historically on different

dose quantities such as first collision dose, multiple collision dose, dose equivalent index, and so

forth. The numerical difference in using these dose quantities compared to the Hp(10) dose used in

current DOELAP performance testing could be evaluated by using the relative values of the dose

conversion factors for the respective dose quantities in conjunction with characteristics of the

respective INL neutron dosimeter response characteristics and workplace radiation fields. 

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]. 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 11 of 58

6.3 DOSE RECONSTRUCTION PARAMETERS

6.3.1

It was INL and ANL-W policy 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. Control badges, which are used 

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

radiation levels have been monitored using film badges and then thermoluminescent dosimeters 

(TLDs). Table 4-13 in the environmental dose TBD (ORAUT 2010a) 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 site workers.

Site Administrative Practices

To provide consistency of radiation safety programs at the INL among a large variety of facilities and 

constantly changing contractors, the AEC established a Health and Safety (H&S) Laboratory at the 

INL to provide technical support for internal and external dosimetry programs. The name of this 

organization changed to Health Services Laboratory (HSL), then to the Health and Safety Division 

(HSD), then to the Idaho Center for Radiological and Environmental Sciences, and most recently to 

the Radiological and Environmental Sciences Laboratory (RESL). 

Some individuals who 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. Early on, 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, when it can be determined from the individual’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 redacted dosimetry information for former workers, which included a dose summary for 

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

location. See Figures 6-1 to 6-5 for an example. 

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). These dose cards were color coded to indicate at which facility the exposures 

were received; however, the copies being received from the DOE are only in black and white so work 

locations prior to1959 can only be determined from the “AEC-IDO-Locater File Cards” when provided 

or from the telephone interview information. 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.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 12 of 58

Figure 6-1. Individual dose reporting form in use until 1958 (Vivian and 

Rockhold 2003, p. 17). 

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. 

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.

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

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 shows a representative example of the records format used from March 1958 through 

November 1966. During that timeframe there were some periods with slightly different formats, but 

the general layout of the records was the same. 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. 

Figures 6-3 through 6-6 provide further examples of records formats used at INL.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 13 of 58

 

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

p. 43). 

Figure 6-3. Example of dosimeter reporting format used from March 1958 through November 

1966 (Vivian and Rockhold 2003, p. 46).
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 14 of 58

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

Area code Description Area code Description

1 AEC Headquarters Building 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 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 SL-1

12 EX at EBR 31 Shield test pool facility

13, 133,135 SPERT, PBF 65 ECF

14 OMRE 66 Nonsecurity

15 SX at SPERT 67 Division of Compliance

16 SL-1 68 STEP

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

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

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

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

Old Later years Old Later years

01 Regular pull 11 Lost pencil (or damaged) Visitor HP 

request

02 HP request Miscellaneous pull 12 HP 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 HP request 16 Correction

07 Recovered lost badge and 

withdrawn

17 Records withdrawn

08 Late pull or 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
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 15 of 58

Figure 6-4. Example of dosimeter reporting format used from December 1966–1985. 

Figure 6-5. Example of dosimeter reporting format used from 1986–1998. 

Figure 6-6. Example of dosimeter reporting format used from 1999–present.

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 

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

computer-generated report.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 16 of 58

For the early computer records (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 6-7 is a listing of some doses received during recovery from the Stationary Low-Power Reactor 

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

and it is notable that one result exceeds the dose limits and that there are few zeros. NIOSH has a 

significant dosimetry history for first-responder and recovery workers at the SL-1 facility. The 

dosimetric records provide enough data to accurately reconstruct doses. In some instances SL-1 

specific coworker dose is used. Figure 6-8 is a follow-up badge report for one result on Figure 6-7. 

When there was a question about an assigned dose value, a Personnel Exposure Questionnaire

(PEQ) was normally initiated as shown in Figure 6-9 (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.

Figure 6-7. 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. 03 Effective Date: 04/19/2011 Page 17 of 58

Figure 6-8. Special badge report associated with a high beta 

reading in Figure 6-7 (Vivian and Rockhold 2003, p. 57).

Figure 6-9. Personnel exposure questionnaire partially completed for 

a hypothetical case (Cipperley 1958). 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 18 of 58

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 open window. 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 the Materials Test Reactor and the Experimental Breeder Reactor

(Cipperley 1958). 

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

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

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

The 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 30 mR, and DuPont 558 film has a threshold level of 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 over-responds per 

unit exposure to low-energy photon radiation, as shown in Figure 6-10, by 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 high Z

severely attenuates the photons that get to the film, so the over-response is reduced to a factor of 2 at 

125 keV and becomes less than 1 at 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.

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.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 19 of 58

Figure 6-10. 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). 

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 6-10 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 

limit of detection (LOD) of 10 mrem was used for both beta and gamma radiation (Horan 1963).

The DOE Idaho Operations Office 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 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 20 of 58

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

uranium, the analysis had greater uncertainty.

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

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

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

efficiency of report and record generation. A 12-point calibration curve was generated for radium and 

for 137Cs gamma and uranium beta. Calibration did not use a phantom.

Experience 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 into SL-1 during recovery operations (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 TLD reader. LiF was chosen because the average Z is close to that of 

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 undated), 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 undated). 

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 LOD of 30 mR was used for gamma and beta (Walker 1974).

6.3.2.5 Harshaw Two-Chip Thermoluminescent Dosimeter System

Several unstable characteristics including over response, warping of the TLD material and 

reproducibility of results, 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 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 21 of 58

the prime contractor at Test Area North (TAN), Test Reactor Area (TRA), etc., and in May 1975 for 

ANL-W. These issues were evaluated by the HSL at the time and no modifications to worker doses 

were deemed necessary (Black 1974; Walker 1974). 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 

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 LOD was 15 mrem beta and gamma from January to July 1986 (Gesell 1986), 10 mrem gamma 

and 30 mrem beta from July 1986 to 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 Film for Neutrons

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

radiation protection staff requested it. NTA film 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 LOD 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 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 22 of 58

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

sheets support the LOD 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 LOD. 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 LOD 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).

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). A 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 LOD 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). 

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

Note: The facility neutron correction factors in this table were used by the site to generate the 

reported neutron doses.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 23 of 58

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 

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

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

backscatter.

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).

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 (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 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 to be too 

low because the dosimeter response is 

usually lower at non-AP angles. Effect is 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 24 of 58

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.

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 in 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 

conversion factor Cx value of 1.08 was used (DOE 1981). 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. 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).

Radiation fields have been characterized by comparing field measurements with a NaI(Tl) gamma 

spectrometer and TLDs mounted on a phantom (Reilly 1998). Most results lie between +27% and -

43%. The high gamma-bias results are for locations at RWMC looking at skyshine (backscattered 

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 100 

keV, so some low-energy photons were likely to have been missed. 

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). 

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. 

Depends on workplace neutron spectra. NTA 

recorded dose of record likely to be too low

because of high 500-keV threshold for detection 

of neutrons. 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 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. 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 25 of 58

Angular response INL dosimeters are calibrated using AP 

laboratory irradiation.

Recorded dose of record is likely to be too low 

because the dosimeter response is lower at nonAP angles. The effect is highly dependent on 

neutron energy.

Environmental 

stability

INL NTA film dosimeters and TLDs are 

subject to signal fade with time, heat, 

humidity, light, etc.

Recorded dose of record is likely to be too low; 

however, this depends strongly on 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).

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 11 

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

AmBe source should be only 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-cm2

/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 11% high. Monte Carlo calculations for 5-MeV 

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

/n averaged over the 0- to 2-cm shell on a 30-

cm-diameter cylindrical phantom (NCRP 1971). 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).

Figure 6-11. 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 and ANL-W as discussed in the Site Description TBD 

(ORAUT 2010b). 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. 03 Effective Date: 04/19/2011 Page 26 of 58

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 

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 6-12 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.

6.3.4.1 Penetrating Radiation

Penetrating gamma rays and X-rays are considered to be ≥30 keV photons. The radiation fields at 

the site, 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].

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
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 27 of 58

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

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

There are a few 250-keV X-ray generators and a 320-keV and a 160-keV X-ray generator that were 

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 

Figure 6-12. Collective dose for INL personnel, 1951 to 1990 

(Horan and Braun 1993). 

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].
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 28 of 58

All 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 Non-Penetrating Radiation

Non-penetrating dose is important for certain cancers (e.g. skin, breast, testes, etc...). Nonpenetrating radiations are considered to be >15 keV electrons and <30 keV photons and X-rays. 

Because the site had essentially no separated plutonium, all non-penetrating doses should be 

assigned as >15 keV electron doses. Electron 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 electron fields were not unusual at the Idaho Nuclear Technology and Engineering Center 

(previously the ICPP) where there are large quantities of fission products. Pure high-energy electron

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 electron fields would therefore be measured as gamma fields (Black 1974; Jenson 1976; 

INEL ca. 1978).

The high-bias electron results in Figure 6-10 from comparison of TLDs to a Phoswich electron 

spectrometer are for electron sources at contact or at 1 cm, which results in hard-to-reproduce 

geometry. The low-bias electron results are for large area sources for which even the spectrometry 

results have large variations. The electron occupational radiation fields (only three) have a bias of 

more than 15% (Reilly 1998).

Electron field dosimetry became reasonably accurate with the definition of DOELAP requirements in 

the early 1980s. Before then, electron monitoring systems had various flaws, primarily in a detector 

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

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

Process/

buildings

Description

Radiation 

type

Energy 

group 

Operations Begin End (keV) Percentage

Reactors Highly dispersed fields of higher energy photon radiation 

fields from fission process, activation, and fission product 

nuclides. Potential for significant airborne nuclides, and 

there might be significant higher energy electron

radiation

Electron >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 electron radiation during sampling and 

maintenance work resulting from fission products. 

Electron >15 100

Photon 30–250 25

>250 75

ICPP, ANL-W 1952 2003

Calibrations Calibration of instruments and dosimeters. Electron >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.

Electron >15 100

Photon 30–250 25

RWMC, WERF 1953 2003 >250 75

Uranium 

handling

Produced special armor from depleted uranium. 

Primarily electron radiation from U-238 daughters. Some 

Electron >15 100

Photon 30–250 90
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 29 of 58

gamma from contaminants and Cs-137 sources used in 

process.

>250 10

SMC 1985 2003

high-energy electrons from either uranium or strontium (see Section 6.3.3.1). The dose from lowenergy electrons can be missed altogether if the electron energy is not sufficient to penetrate the 

detector cover, and the dose can be under-reported if the electron energy is not sufficient to penetrate 

the entire detector. The mean electron energy for the spectrum from a particular nuclide is about onethird of the maximum electron energy for that nuclide.

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

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

greater than the abscissa is plotted on the ordinate in Figure 6-13. Electron-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 electron

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 electron 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 electron 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 electron dose correctly at a depth of 7 mg/cm2

, which was chosen in the 

early 1980s. These curves demonstrate why the electron 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.

 

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 6-13. Distribution of electron ranges [14].

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 electron

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 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 30 of 58

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.

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

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

the average as described above. To determine the corrected electron 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. Electron dosimeter thicknesses and associated under-reporting [17]. 

Dosimeter 

system Period

Covers

(mg/cm2)

Detector thickness

(mg/cm2)

Electron

correction factor

Self-service system Aug 1951–Feb1958 ~50 ~50 2.0

Multifilter film Mar 1958–Sept 1973 100 ~50 2.8

LiF Teflon TLD Nov 1966–1974 100 75 3.0

ATLAS Feb 1974–May1975 100 ~100 3.3

Harshaw 2 chip TLD Dec 1974–1975 104 240 4.8

Modified Harshaw TLD 1976–1985 4 240 2.4

Panasonic TLD 1986–2006 16 15 1.0

a. Note: The site began phasing out the use of the multifilter film dosimeter in November 1966 but didn't completely phase 

out its use until September 1973.

6.3.4.3 Neutron Radiation

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

Neutron fields are specific to a few facilities. The high-dose locations where most of the gamma and 

electron 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]. 

Location Period Comments

Programmatic

ANL-W, TREAT 1958–1994 Experiment floor

ANL-W, ZPPR 1969–1992 Between reactor halves during maintenance

MTR 1952–1970 Experiment floor

TAN Warm Shop 1986–1988 See EPRI 1986, 1987

TAN 1986–2006 Spent fuel storage pad

ANL-W 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

ANL-W, ZPPR 1969–1992 For instrument calibration
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 31 of 58

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 6-14 (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 has been used by the site 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 6-15. 

These results show a greater dispersion than the gamma results. The two lowest values (-52% and 

-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 

Figure 6-14. 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). 

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). 

Total number of dosimeters with reportable dose = 54 Total number of dosimeters with reportable dose = 54
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 32 of 58

 

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 6-15. Neutron radiation field characterization (Reilly 1998). 

Figure 6-16 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 6-16. Neutron spectra that simulate INL facilities [19]. 

Most of the reactors built at INL and ANL-W had no beam ports. The neutrons were 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
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 33 of 58

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.

6.3.4.3.1 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 an experiment floor. The Zero 

Power Plutonium (later Physics) Reactor (ZPPR) and the Transient Reactor and Experiment Test 

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

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

provide all components of the radiation field. Hankins 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 6-17 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 6-18 shows the correlations of the thermal and intermediate neutron dose equivalents to the 

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

The trend line for the reanalyzed intermediate-energy neutron dose equivalent has 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 

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, 

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 6-17. 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 6-18. MTR neutron field 

components (ORAUT 2006a).

the fraction of neutron dose equivalent above 0.8 MeV has an average value of 0.52 ± 0.08 and varies 

from 35% to 66% depending on the location (ORAUT 2006a). To correct for the NTA film dosimeter’s 

energy response limitations, neutron doses for the MTR workers measured with NTA film should be 

adjusted by a factor of 2 ± 0.3 (1/0.52 ± 0.08/0.522

) (ORAUT 2006a). 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 34 of 58

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 6-19 shows the correlation of fast neutron dose 

equivalent to the gamma dose for these measurements. 

 

 

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 6-19. Correlation of fast neutron dose 

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

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

Figure 6-19 are shown with the triangles at one-half of the smallest measured value. Using the 

Shapiro-Wilks Normality Test (Gilbert 1987) and including the insignificant fast neutron values at onehalf 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 experiment floor [20]: 

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 in 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 

experiment floor while the reactor was operating. For example, health physics technicians would 

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. Further extrapolating experience at 

the MTR at TRA to TREAT or ZPPR at ANL-W is not recommended [22]. While similarities of the 

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

neutron radiation might not be. In addition, dosimetry data sheets demonstrate that small neutron 
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 35 of 58

doses were as likely or more likely to be determined as small photon doses, so this sort of correction 

is not necessary [23].

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

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

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

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

Figure 6-20) and neutron dose assignments based on time and neutron dose rate. Neutron doses 

were estimated based on the measured neutron:gamma ratio and gamma doses measured with 

pencil dosimeters. These results were entered in the dosimetry records (Rohrig 2006a). Shields 

were also used to minimize the reactor surface open to workers. Figure 6-21 shows the integral 

distribution of neutron dose with energy for this spectrum. The fraction of dose below the NTA cutoff 

is quite consistent with the MTR result.

6.3.4.3.2 Test Area North Fuel Storage Casks

As noted in the Site Description TBD (ORAUT 2010b, 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 six people 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 to 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 

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 

Figure 6-20. ZPPR Assembly-3 Phase-1A fast neutron spectrum measured midway between halves 

with both shields withdrawn (Simons, Young, and Thalgott 1972).
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 36 of 58

Figure 6-21. ZPPR Assembly-3 Phase-1A fast neutron dose spectrum 

midway between halves with both shields withdrawn (Simons, Young, 

and Thalgott 1972).

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 [24].

6.4 ADJUSTMENTS TO RECORDED DOSE 

6.4.1

No adjustments to photon doses are necessary.

Photon Dose Adjustments

6.4.2

Electron field dosimetry became reasonably accurate with the definition of DOELAP requirements in 

the early 1980s. Before then, electron monitoring systems had various flaws, primarily in a detector 

too thick to give a good surface result or one that was covered with extra material. The dose from 

low-energy electrons can be missed altogether if the electron energy is not sufficient to penetrate the 

detector cover, and the dose can be under-reported if the electron energy is not sufficient to penetrate 

the entire detector. To determine the corrected electron dose, the measured and missed electron

results from the dosimetry system should be multiplied by the values in Table 6-12. 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. For the Panasonic system that was 

used after 1985, such a correction has already been made (Gesell 1986), so the recommended 

correction is 1.

Electron Dose Adjustments
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 37 of 58

Table 6-12. Electron dosimeter under-reporting 

correction factors. 

Period Electron correction factor

1951–1957 2.0

1958–1965 2.8

1966–1973 3.0

1974–1975 4.8

1976–1985 2.4

1986–2006 1.0

6.4.3

6.4.3.1 Neutron Energy Response Adjustments

Neutron Dose Adjustments

Because of the NTA film dosimeter’s energy response limitations, the following adjustment factors 

should be applied to the measured and missed neutron doses reported prior to October 1976. In 

October 1976 the site began using TLDs to measure neutron doses, which did not have any 

significant energy response limitations.

As previously recommended in section 6.3.4.3, the neutron doses for the MTR workers measured with 

NTA film should be adjusted by a factor of 2 ± 0.3. Section 6.3.4.3 also indicated that the neutron 

spectra for the TREAT and ZPPR facilities were comparable to the MTR’s neutron spectra. 

Therefore, the adjustment factor for the MTR neutron doses is applicable to the TREAT and ZPPR 

facilities. To simplify the dose reconstruction process, the measured and missed neutron doses from 

the MTR, TREAT, and ZPPR should be multiplied by 2.3, which is the upper-bound for this correction 

factor.

As indicated in Table 6-15, the fraction of lower energy neutrons at the other site facilities is ≤ 20%, 

which means the NTA film dosimeters were capable of measuring at least 80% of the neutron doses 

at those facilities. Therefore, the neutron doses for workers at facilities other than the MTR, TREAT, 

or ZPPR should be adjusted by a factor of 1.25 (1/0.80).

6.4.3.2 Radiation Weighting Factor Adjustments

All measured and reported dose equivalents from INL in this TBD 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 [27]. 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). 

The reported dosimetry data require correction to change from dose equivalent (pre-ICRP Publication 

60) to the newer dose quantity (ICRP 1991; NIOSH 2007). 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 (2007), 

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.

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)
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 38 of 58

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 [29].

These factors are incorporated in 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 1971). 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 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 [28].

Table 6-13 summarizes the locations where neutron dose is credible. Table 6-14 lists the calculated 

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

equivalent to equivalent dose for site 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 [31]. The lower part of the table lists the recommended 

default values for the dose equivalent fractions and quality factor corrections. These values are 

combined in Table 6-13. This correction should be applied to the measured and missed neutron 

doses.

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

[30]. 

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 (1996) H*10 /NCRP (1971b) 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. 03 Effective Date: 04/19/2011 Page 39 of 58

Table 6-14. Recommended IREP neutron energy fractions and correction factors [32].

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 sourcebased research2003 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

Missed dose for site workers would occur when a zero dose was recorded for the dosimeter systems 

for any response less than the site dose recording threshold (the LOD). In the 1975–1985 dosimetry 

records, when the electron, penetrating, and neutron doses are all reported as “000001” mrem, the 

worker was not in the area and no missed doses should be assigned for those dosimeters.

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 2004). The latest revision of Use of Coworker Dosimetry 

Data for External Dose Assignment (ORAUT 2008a) 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 workers would occur when a zero dose was recorded for the dosimeter 

systems for any response less than the site dose recording threshold (the LOD). The missed dose for 

dosimeter results less than the LOD is particularly important for earlier years when LODs were higher 

and dosimeter exchange was more frequent [33]. The missed dose is calculated as described in 

NIOSH (2007) using LOD/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 LODs 

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.

Missed Photon Dose

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

annual missed dose [34].

Period of usea Dosimeter

Exchange 

frequency

LODb (mrem)

Annual missed 

dose (mrem)c

Photon Electron Photon Electron

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 Weekly (n=52) 10 30 260 780
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 40 of 58

film

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. LODs 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 × LOD/2 from NIOSH (2007).

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 LOD was 15 mrem from January 1, 1986, to July 7 1986; 10 mrem from July 7, 1986, to September 1989; and 15 

mrem until 1993, when it returned to 10 mrem.

f. The LOD was 15 mrem from January 1, 1986, to July 7 1986, and 30 mrem after that.

6.5.3

Non-penetrating dose is important for certain cancers (e.g. skin, breast, testes, etc...). Because the 

non-penetrating and penetrating doses were measured by the same dosimeter, because the 

dosimeters’ LODs for electron doses are sometimes higher than the LODs for photon doses, and 

because the dosimeter correction factors that only get applied to electron doses are sometimes 

significant, special instructions are needed for assigning the missed doses for cancer locations 

affected by non-penetrating radiation. The following are the special instructions for each situation that 

may be encountered for the affected cancer locations.

Missed Electron Dose

1. When the reported non-penetrating result for a dosimeter is less than its electron LOD/2 value 

and the corresponding penetrating result is greater than its photon LOD/2 value, a missed 

dose is assigned as a >15 keV electron dose using the applicable parameters for electron 

doses. 

2. For instances when the non-penetrating and penetrating doses for a dosimeter are both below 

their respective electron and photon LOD/2 values, the missed dose is calculated as an 

electron dose using the applicable parameters for electron doses, but is assigned as a more 

favorable to claimant 30-250 keV photon dose.

3. When the reported non-penetrating result for a dosimeter is greater than its electron LOD/2 

value and the corresponding penetrating result is less than its photon LOD/2 value, a missed 

dose is assigned as a 30-250 keV photon dose using the applicable parameters for photon 

doses.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 41 of 58

6.5.4

As previously indicated, only workers that had a potential to receive more than an incidental neutron 

dose were the workers that were monitored for neutron dose and had their neutron dosimeters read. 

The only exception to this might have been the workers that worked around the fuel storage casks on 

a storage pad at TAN during the period of 1986-2006. However, those workers were unmonitored 

and should only have unmonitored neutron doses assessed. Therefore, missed neutron doses should 

only be assessed for the workers with reported neutron dosimeter results using the Table 6-16 

information and any applicable dosimeter adjustment factors.

Missed Neutron Dose

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

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

Dosimeter Period

Exchange 

frequency

Laboratory 

LOD (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

Because of the use of placeholder zeros in the 1975–1985 dosimetry records, determining a worker’s 

neutron exposure potential can be difficult; the following guidance should be used for this period. 

1. If a worker had a reported neutron dose greater than zero during the period of 1975–1985 for 

a given dosimeter location code, measured and missed neutron doses should be assessed for 

all dosimeter results during the 1975–1985 period with that location code. 

2. If all of a worker’s reported neutron doses for the 1975–1985 period were zero, but neutron 

monitoring can be confirmed for a given dosimeter location code during the years immediately 

prior to 1975 or immediately after 1985, missed neutron doses should be assessed for all 

dosimeter results during the 1975–1985 period with that location code.

3. For any situations outside of those described in 1) and 2), a worker’s potential to receive 

neutron dose was low and no missed neutron doses should be assessed.

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 

LODs. 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 

film indicated a neutron dose equivalent that was less than the film’s 14-mrem LOD [45]. When the 

LOD 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 experiment floor and on the TREAT or ZPPR experiment 

floor [46]. 

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

neutron-to-photon dose ratios (NIOSH 2007). However, for other site facilities there are several other 

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

erroneous.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 42 of 58

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 LOD 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 2007). For photons 

before 1981, the conversion factor from exposure to organ dose should be used [47]. For 1981 and 

after, the conversion factor from deep dose equivalent to organ dose should be used [48].

6.7 UNCERTAINTY

Measurement uncertainties arise from many sources. For gamma rays, the standards for exposure 

have existed with only minor changes since the 1930s as required for medical uses of radiation. INL 

used ionization chambers standardized by NBS for its calibrations. Use of a phantom for dosimeter 

irradiation began in the early 1980s, but backscatter causes only a minor change for high-energy 

photon dosimetry [49]. 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 [50]. 

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 electron radiation. A realistic estimate of 

total uncertainty for photon dosimetry is ± 35% at 1 sigma [51]. This is roughly consistent with the 

results in a study that was performed on the dosimeters in 1998 (Reilly 1998). In that study, the 

standardization instrument contributed some significant uncertainty.

The uncertainty for beta radiation is somewhat larger at an estimated ± 50% at 1 sigma [52]. This is 

driven by uncertainties in the field geometry and because beta radiation is often stopped by air and 

thin materials such as clothing. 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 [53]. The difference between point- and planar-source irradiation can confuse an 

algorithm [54]. Earlier techniques did not provide a thin detector with minimal covering, which is 

important for simulating the skin for beta dosimetry.

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 [55]. 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 larger at 60% at one sigma. 

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 a factor of 1,000 higher than the dose 

on the back side of a 30-cm-diameter sphere of tissue-equivalent material [56]. 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) [57]. 

In Figure 6-18 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.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 43 of 58

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 that was published in 2004 and 

revised in 2006. Because 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, the 

document was reviewed and revised by a new Document Owner in 2007. The 2007 revision was 

made in full accordance with the “NIOSH Policy Statement: Management of Conflict or Bias in the 

Radiation Dose Reconstruction Program.” The new Document Owner was fully responsible for the 

content of the revised document, including all findings and conclusions. In all cases where such 

information or prior studies or writings are included or relied on by the new Document Owner, those 

materials are fully attributed to the source. Mr. Rohrig served no role in this latest revision. 

[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, Site Profile and TBD Development

(ORAUT 2007), and uses values based on experience and from the Hanford Site Profile 

(ORAUT 2010c).

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

This table is adapted to INL from the Hanford Site Profile (ORAUT 2010c) and uses values 

based on experience at 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 2010b).

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

The contact radiation field on these devices is less than 0.5 mrem/hr (Rohrig 2006a).
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 44 of 58

[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. 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.

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 6-10. 

[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 2010b) and work 

experience at the facility.

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

This plot was generated from 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%.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 45 of 58

[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 might not be directly related to each other; they often 

have different sources. Because the gamma and neutron fields do not have a common 

source, there is no reason to believe the ratios from one facility would apply at another facility. 

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

Using photon-to-neutron ratios to derive neutron doses assumes that valid photon data are 

considerably more likely to exist than valid neutron data. This does not seem to be the case 

on examining dose records.

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

Based on conversation on October 12, 2006, with a former health physicist with knowledge of 

TAN and INL (Rohrig 2006b).

[25] 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 on the similarity between the calibration and workplace radiation field.

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

This standard TBD table is based on information in the Hanford TBD (ORAUT 2010b) and, in 

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

performance. 

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

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

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

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

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

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

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

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

This table was derived with spectral data in a spreadsheet as shown in Figures 6-15 and 6-16 

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.

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

The values in Table 6-13 are different, namely 1.08 and 1.46 or 1.47.

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

This table is derived from the values in Table 6-13. 

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

The calculated missed dose, based on OCAS-IG-001 guidance (NIOSH 2007) 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.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 46 of 58

[34] 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 TBD. 

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

This statement is made based on studying dose record sheets and conferring with INL 

dosimetry staff.

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

This judgment is based on reviews of historical INL dosimetry information and Watson et al. 

(1994). 

[37] 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.

[38] 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 (ORAUT 2008b). 

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

This judgment is based on reviews of historical INL dosimetry information. 

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

These values are based on weighting neutron spectra with dose conversion factors to 

determine the fraction of the dose below 0.8 MeV.

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

The values in Table 6-15 are required to calculate the missed dose per ORAUT-PROC-0031 

(ORAUT 2007) and are based on the judgment of Mr. Rohrig.

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

This reflects the period discussed in Section 6.3.2.7.

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

This reflects the period discussed in Section 6.3.2.8.

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

A perfect calibration source would have all its radiation detected by the detector being 

calibrated. If most of the radiation is detected, the adjustment would be small and would not 

contribute a major uncertainty. 

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

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

blank. 

[46] 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.

[47] 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.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 47 of 58

[48] 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.

[49] 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).

[50] 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.

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

This estimate is based partially on Figure 6-12 and partially on uncertainties of many error 

processes that contribute uncertainty.

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

This estimate is based partially on Figure 6-12 and partially on uncertainties of many error 

processes that contribute uncertainty.

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

These estimated corrections are provided in Table 6-12, and they are not small.

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

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

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

Lithium is only a trace constituent of the human body. Dose from neutron radiation to people 

is primarily due to hydrogen atom recoils in tissue rather than the 6

Li(n,α) neutron absorption 

reactions of the neutron dosimeter.

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

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

[57] 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. 03 Effective Date: 04/19/2011 Page 48 of 58

REFERENCES

AEC (U.S. Atomic Energy Commission), 1960, Annual Report of Health and Safety Division, 1959, 

Idaho Operations Office, IDO-12014, Idaho Operations Office, Idaho Falls, Idaho, October. 

[SRDB Ref ID: 8080]

Andersen, B. L., 1995, Panoramic Irradiator – Program Development to Acquire NIST Equivalent 

Irradiations for the INEL Dosimetry Programs, EDF ODU-94-5.0, EG&G Idaho, Idaho Falls, 

Idaho, January 6. [SRDB Ref ID: 8102]

Anderson, M. E., and M. R. Hertz, 1971, “Thick Target Yield for the 9

Be(α,n) Reaction,” Nuclear 

Science and Engineering, volume 44, pp. 437–439. [SRDB Ref ID: 3208]

ANSI (American National Standards Institute), 1972, Performance Specifications for Direct Reading 

and Indirect Reading Pocket Dosimeters for X- and Gamma Radiation, ANSI N13.5-1972, New 

York, New York.

Aoki, I., 1979, "General Information for Evaluating Penetrating Exposure Summary at INEL for Years 

1950 through 1979," memorandum to Dosimetry File, May 31. [SRDB Ref ID: 8529]

Bean, L. C., 1995, Establishing a Beta Source for QA/QC Irradiations, EDF RDRU 95-1.0, EG&G 

Idaho, Idaho Falls, Idaho, February 24. [SRDB Ref ID: 8465]

Bennett, W. P., 1957, Correlation of Instrument and Film Radiation Readings – CPP, PTR-257, 

Phillips Petroleum Company, Idaho Falls, Idaho, November 15. [SRDB Ref ID: 8469]

Black, D. E., 1974, "ATLAS System Recommendations," BLA-155-74, letter to D. I. Walker (U.S. 

Atomic Energy Commission), Allied Chemical Corporation, Idaho Falls, Idaho, June 21. 

[SRDB Ref ID: 8128, p. 3]

BRH (Bureau of Radiological Health), 1970, Radiological Health Handbook, Revised Edition, U.S. 

Department of Health, Education and Welfare, Rockville, Maryland.

Cipperley, F. V., 1958, Personnel Metering Operating Techniques, Self-Service System, Effective 

August 20, 1951 through March 29, 1958, IDO-12007, U.S. Atomic Energy Commission, Idaho 

Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 8112]

Cipperley, F. V., 1966, A General Summary of Personnel Monitoring Systems Utilized at the National 

Reactor Testing Station, IDO-12056, U.S. Atomic Energy Commission, Idaho Operations 

Office, Idaho Falls, Idaho, September. [SRDB Ref ID: 8130]

Cipperley, F. V., 1968, Personnel Dosimetry Standard Operating Procedure Effective March 30, 1958, 

IDO-12070, U.S. Atomic Energy Commission, Idaho Operations Office, Idaho Falls, Idaho, 

December. [SRDB Ref ID: 8086]

Cipperley, F. V., and W. P. Gammill, 1959, Improvements in Personnel Metering Procedures at 

NRTS, U.S. Atomic Energy Commission, Idaho Operations Office, Idaho Falls, Idaho. [SRDB 

Ref ID: 8113]

Cipperley, F. V., R. W. Henry, and J. P. Cusimano, 1965, “Personnel Monitoring Practices at the 

National Reactor Testing Station,” Symposium on Personnel Dosimetry for Accidental HighLevel Exposure to External and Internal Radiation, International Atomic Energy Agency, March 

8-12, 1965, Vienna, Austria. [SRDB Ref ID: 8105]
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 49 of 58

Cusimano, J. P., 1963, Latent Image Fading in Personal Neutron Monitoring Film at the NRTS, IDO12031, U.S. Atomic Energy Commission, Idaho Operations Office, Idaho Falls, Idaho. [SRDB 

Ref ID: 8135]

Cusimano, J. P., 1972, "Suggestions and/or Alternatives Concerning the ATLAS Dosimetry Program," 

memorandum to F. V. Cipperley, Health Services Laboratory, Idaho Falls, Idaho, December 7. 

[SRDB Ref ID: 8111]

Cusimano, J. P., 1981, "Update of Facility Neutron Calibration Factor (FNCF)," memorandum to 

Dosimetry File, U.S. Department of Energy, Idaho Operations Office, Idaho Falls, Idaho, 

January 21. [SRDB Ref ID: 8145]

Cusimano, J. P., 1982, “Changeout and Storage of Hurst NADs until Disposition,” memorandum to B. 

Burkhart, April 22. [SRDB Ref ID: 8109]

Cusimano, J. P., and F. V. Cipperley, 1968, “Personnel Dosimetry Using Thermoluminescent Lithium 

Fluoride-Teflon Dosimeters,” Health Physics, volume 14, pp. 339–344. [SRDB Ref ID: 8089]

DOE (U.S. Department of Energy), 1981, "Personnel Dosimetry Performance," memorandum to file 

from Dosimetry Branch, Idaho Operations Office, Idaho Falls, Idaho, January 8. [SRDB Ref 

ID: 8125]

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

Testing of Personnel Dosimetry Systems, DOE Laboratory Accreditation Program for 

Personnel Dosimetry Systems, DOE/EH-0027, Assistant Secretary for Environment, Safety, 

and Health, Washington, D.C., December. [SRDB Ref ID: 12294]

DOE (U.S. Department of Energy), 1986b, Handbook for the Department of Energy Laboratory 

Accreditation Program for Personnel Dosimetry Systems, DOE/EH-0026, Assistant Secretary 

for Environment, Safety, and Health, Washington, D.C., December. [SRDB Ref ID: 12293]

EPRI (Electric Power Research Institute), 1986, The Castor-V/21 PWR Spent-Fuel Storage Cask: 

Testing and Analysis, EPRI NP-4887, Palo Alto, California, November. [SRDB Ref ID: 79247]

EPRI (Electric Power Research Institute), 1987, The MC-10 PWR Spent-Fuel Storage Cask: Testing 

and Analysis, EPRI NP-5268, Palo Alto, California, July. [SRDB Ref ID: 78941]

Gesell, T. F., 1982a, “Calibration of the INEL Personnel Dosimeter for Non-Penetrating Radiation,” 

letter to S. Marchetti (West Valley Nuclear Services), U.S. Department of Energy, Idaho 

Operations Office, Idaho Falls, Idaho, October 26. [SRDB Ref ID: 8127]

Gesell, T. F., 1982b, “Estimation of External Dose Equivalents from TL Dosimeters,” memorandum to 

Dosimetry File, U.S. Department of Energy, Idaho Operations Office, Idaho Falls, Idaho, 

February 3. [SRDB Ref ID: 8167]

Gesell, T. F., 1986, “Minimum Reporting Levels for the Panasonic Dosimetry Systems,” letter to 

Dosimetry Customers, Radiological and Environmental Sciences Laboratory, May 15. [SRDB 

Ref ID: 8123]

Gesell, T. F., R. Boston, B. L. Anderson, and R. Perry, 1996, Technical Basis of the INEL Personal 

Neutron Dosimeter, 960112, Idaho State University, Department of Physics, January 12. 

[SRDB Ref ID: 8157]
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 50 of 58

Gesell, T. F., M. A. Hall, and B. L. Anderson, 1992, Technical Basis of the INEL Personal Photon-Beta 

Dosimeter, Idaho State University, Department of Physics, December 1. [SRDB Ref ID: 8163]

Gilbert, R. O., 1987, Statistical Methods for Environmental Pollution Monitoring, Van Nostrand 

Reinhold, New York, New York, January 1.

Gupta, V. P., 1981, Standardization and Use of 2.5 Ci 90Sr/90Y Source for Calibration of Personnel 

Dosimeters and Portable Instruments, IDO-12093, U.S. Department of Energy, Idaho 

Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 8155]

Hankins, D. E., 1961, A Method of Determining the Intermediate Energy Neutron Dose, IDO-16655, 

Phillips Petroleum Company, Atomic Energy Division, Idaho Falls, Idaho, March 10. [SRDB 

Ref ID: 8084]

Hankins, D. E., 1973, A Small, Inexpensive Albedo-Neutron Dosimeter, LA-5261, University of 

California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, July. [SRDB Ref ID: 

8081]

Hankins, D. E., 1976, The Effect of Energy Dependence on the Evaluation of Albedo Neutron 

Dosimeters, UCRL-77198, University of California, Lawrence Livermore Laboratory, 

Livermore, California. [SRDB Ref ID: 13720]

Hertel, N. E., and J. W. Davidson, 1985, “The Response of Bonner Spheres to Neutrons from Thermal 

Energies to 17.3 MeV,” Nuclear Instruments and Methods in Physics Research, A238(1985), 

pp. 509–516. [SRDB Ref ID: 8440]

Hill-Ellis, M., 2003, Fax of tables, October 1. [SRDB Ref ID: 29240]

Horan, J. R., editor, 1959, Annual Report of the Health and Safety Division, 1958, Idaho Operations 

Office, IDO-12012, U.S. Atomic Energy Commission, Idaho Operations Office, Idaho Falls, 

Idaho. [SRDB Ref ID: 8079]

Horan, J. R., 1962, Health and Safety Division Annual Report 1961, IDO-12021, U.S. Atomic Energy 

Commission, Idaho Operations Office, Idaho Falls, Idaho, May 21. [SRDB Ref ID: 8076]

Horan, J. R., 1963, Health and Safety Division Annual Report 1962, IDO-12033, U.S. Atomic Energy 

Commission, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 8075]

Horan, J. R., and J. B. Braun, 1993, Occupational Radiation Exposure History of Idaho Field Office 

Operations at the INEL, EGG-CS-11143, EG&G Idaho, Idaho National Engineering 

Laboratory, Idaho Falls, Idaho, October. [SRDB Ref ID: 1023]

IAEA (International Atomic Energy Agency), 1988, Guidelines on Calibration of Neutron Measuring 

Devices, Technical Reports Series No. 285, Vienna, Austria, August 31.

ICRP (International Commission on Radiological Protection), 1973, Data for Protection Against 

Ionizing Radiation from External Sources: Supplement to ICRP Publication 15, Publication 21, 

Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1977, Recommendations of the 

International Commission on Radiological Protection, Publication 26, Pergamon Press, Oxford, 

England.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 51 of 58

ICRP (International Commission on Radiological Protection), 1979, Limits for the Intake of 

Radionuclides by Workers, Part 1, Publication 30, Part 1, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1987, Data for Use in Protection Against 

External Radiation, ICRP Publication 51, Pergamon Press.

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

ICRP (International Commission on Radiological Protection), 1996, Conversion Coefficients for Use in 

Radiological Protection Against External Radiation, Report 74, Pergamon Press, Oxford, 

England.

ICRU (International Commission on Radiation Units and Measurements), 1985, Determination of 

Dose Equivalents Resulting from External Radiation Sources, Report 39, Bethesda, Maryland.

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

in Radiation Protection Dosimetry, Report No. 51, Bethesda, Maryland.

INEEL (Idaho National Engineering and Environmental Laboratory), 2001, Technical Basis of the 

INEEL Personal Photon-Beta Dosimeter, INEEL/EXT/01-00636, Idaho Falls, Idaho. [SRDB 

Ref ID: 8085]

INEL (Idaho National Engineering Laboratory), ca. 1978, “Dosimetry Branch Changes." [SRDB Ref 

ID: 8436]

Ing, H., and S. Makra, 1978, Compendium of Neutron Spectra in Criticality Accident Dosimetry, 

Technical Reports Series No. 180, International Atomic Energy Agency, Vienna, Austria.

Jenson, D. K., 1976, “Personnel Dosimetry Problem Scope,” memorandum to B. L. Rich, July 29. 

[SRDB Ref ID: 8133]

Jones, D., 1980, “DOE Headquarters Request for Comments on Proposed Changes to Neutron 

Quality Factor,” letter to J. Barry (Operational Safety Division), May 28, Radiological and 

Environmental Sciences Laboratory, Idaho Falls, Idaho. [SRDB Ref ID: 8455]

Kalbeitzer, F. L., 1983, “Reportable Threshold Reading from the Unshielded Dosimeter of the INEL 

Personnel Badge,” memorandum to Dosimetry Branch, Radiological and Environmental 

Sciences Laboratory, Idaho Falls, Idaho, March 25. [SRDB Ref ID: 8154]

Kalbeitzer, F. L., 1984, “Computerized Determination of Penetrating/Non-Penetrating Calibration 

Factors,” memorandum to Dosimetry Branch Staff, U.S. Department of Energy, Idaho 

Operations Office, Idaho Falls, Idaho, September 27. [SRDB Ref ID: 8151]

Kluge, H., and K. Weise, 1982, “The Neutron Energy Spectrum of a 241Am-Be(α,n) Source and 

Resulting Mean Fluence to Dose Equivalent Conversion Factors,” Radiation Protection 

Dosimetry, volume 2, number 2, pp. 85–89. [SRDB Ref ID: 8087]

Knoll, G. F., 1989, Radiation Detection and Measurement, Second Edition, John Wiley and Sons, 

New York, NY.
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 52 of 58

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

Neutron Radiation, NCRP Report 38, Bethesda, Maryland, January 4.

NIOSH (National Institute for Occupational Safety and Health), 2007, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 3, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 21.

NIOSH (National Institute for Occupational Safety and Health), 2010, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October. 

ORAUT (Oak Ridge Associated Universities Team), 2004, Meeting with Paper, Allied-Industrial, 

Chemical and Energy Workers Local Union 8-0652; Amalgamated Transit Union 1517; 

Security Police and Fire Professionals of America, Local 3; International Brotherhood of 

Teamsters Local 983, April 28.

ORAUT (Oak Ridge Associated Universities Team), 2006a Technical Information Bulletin in Support 

of INEEL Technical Basis Document Section 6: Reanalysis of Hankins MTR Bonner Sphere 

Surveys, ORAUT-OTIB-0009, Rev. 01, Oak Ridge, Tennessee, December 13.

ORAUT (Oak Ridge Associated Universities Team), 2006b, Technical Basis for Conversion from 

NCRP Report 38 Neutron Quality Factors to ICRP Publication 60 Radiation Weighting Factors 

for Respective IREP Input Neutron Energy Ranges, ORAUT-OTIB-0055, Rev. 00, Oak Ridge, 

Tennessee, June 5. 

ORAUT (Oak Ridge Associated Universities Team), 2006c, Occupational On-Site Ambient Dose 

Reconstruction for DOE Sites, ORAUT-PROC-0060, Rev 01, June 28. 

ORAUT (Oak Ridge Associated Universities Team), 2007, Site Profile and TBD Development, 

ORAUT-PROC-0031, Rev. 02, Oak Ridge, Tennessee, August 17.

ORAUT (Oak Ridge Associated Universities Team), 2008a, Use of Coworker Dosimetry Data for 

External Dose Assignment, ORAUT-OTIB-0020, Rev. 02, Oak Ridge, Tennessee, 

December 4.

ORAUT (Oak Ridge Associated Universities Team), 2008b, Assignment of Missed Neutron Doses 

Based on Dosimeter Records, ORAUT-OTIB-0023, Rev. 01, Oak Ridge, Tennessee, May 14.

ORAUT (Oak Ridge Associated Universities Team), 2010a, Idaho National Laboratory and Argonne 

National Laboratory- West – Occupational Environmental Dose, ORAUT-TKBS-0007-4, Rev. 

02, Oak Ridge, Tennessee, January 8.

ORAUT (Oak Ridge Associated Universities Team), 2010b, Idaho National Laboratory and Argonne 

National Laboratory-West – Site Description, ORAUT-TKBS-0007-2, Rev. 04, Oak Ridge, 

Tennessee, August 2.

ORAUT (Oak Ridge Associated Universities Team), 2010c, Hanford Site – Occupational External 

Dosimetry, ORAUT-TKBS-0006-6, Rev. 04, Oak Ridge, Tennessee, January 7.

Perry, O. R., B. L. Andersen, and P. E. Ruhter, 1993, Minimum Reporting Level for the Idaho National 

Engineering Laboratory (INEL) Thermoluminescent Dosimetry (TLD) Systems, EGG-ESQ10917, Idaho National Engineering Laboratory, Idaho Falls, Idaho. [SRDB Ref ID: 8161]
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 53 of 58

PNL (Pacific Northwest Laboratory), VPC (Virginia Power Company), and EG&G Idaho, 1987, The 

TN-24P Spent-Fuel Storage Cask: Testing and Analyses, EPRI NP-5128, PNL-6150, Electric 

Power Research Institute, Palo Alto, California, April. [SRDB Ref ID: 12609]

Purcell, R. B., and H. McGary, 1963, Automatic Film Density Reading System, IDO-12029, U.S. 

Atomic Energy Commission, Idaho Operations Office, Idaho Falls, Idaho, April 1. [SRDB Ref 

ID: 8096]

Reilly, P. S., 1998, Characterization of Radiation Fields at the Idaho National Engineering and 

Environmental Laboratory, Engineering Design File RDR-98005, EG&G Idaho, Idaho National 

Engineering and Environmental Laboratory, Idaho Falls, Idaho, October 6. [SRDB Ref ID: 

8162]

Roberson, P. L., and K. L. Holbrook, 1984, Guidelines for the Calibration of Personnel Dosimeters, 

PNL-4515, Battelle Memorial Institute, Pacific Northwest Laboratory, Richland, Washington, 

January. [SRDB Ref ID: 8100]

Rohrig, N. 2006a, “N. Rohrig and L. Burke conversation on Neutron Dosimetry at ZPR 3 and 

Analytical X-ray Machines at ANL-W,” documented communication, Idaho Falls, Idaho, 

November 6. [SRDB Ref ID: 31190]

Rohrig, N. 2006b, “N. Rohrig and P. E. Ruhter conversation on Neutron Exposure from Spent Fuel 

Cask at TAN,” documented communication, Idaho Falls, Idaho, October 12. [SRDB Ref ID: 

31191]

Ruhter, P. E., 2003, “Re: Need search for TLD results,” e-mail to L. E. Rockhold, Idaho National 

Laboratory, and November 10. [SRDB Ref ID: 29237]

Ruhter, P. E., and O. R. Perry, 2002, Idaho National Engineering and Environmental Laboratory 

(INEEL) Dosimetry Systems History and Minimum Dose Reporting Level (MRL) Information 

(External Dose Summary Profile – 1951 to March 2002), Bechtel BWXT Idaho, Idaho National 

Engineering and Environmental Laboratory, Idaho Falls, Idaho, April 8. [SRDB Ref ID: 1774, 

p. 2]

Schubauer-Berigan, M. K., G. V. Macievic, D. F. Utterback, C-Y. Tseng, and J. T. Flora, 2005, An 

Epidemiologic Study of Mortality and Radiation-Related Risk of Cancer Among Workers at the 

Idaho National Engineering and Environmental Laboratory, a U. S. Department of Energy 

Facility, HHS (NIOSH) Publication No. 2005-131, National Institute for Occupational Safety 

and Health, Division of Surveillance, Hazard Evaluations, and Field Studies, Cincinnati, Ohio, 

January.

Simons, G. G., J. C. Young, and F. W. Thalgott, 1972, Personnel Neutron-dose-equivalent Studies at 

the Zero Power Plutonium Reactor, in Applied Physics Division Annual Report, ANL-8010, 

Argonne, Illinois, p. 365-372. [SRDB Ref ID: 18900]

Sommers, J. F., 1959, "Neutron Activity Levels at the MTR due to ANP-1," memorandum to R. L. 

Dean, Som-74-59A, Idaho Falls, Idaho, August 27. [SRDB Ref ID: 8451]

Sommers, J. F., 1962, "Thermal Column Measurements," memorandum to W. H. Burgus, Som-126-

62A, Idaho Falls, Idaho, December 18. [SRDB Ref ID: 8442]
Document No. ORAUT-TKBS-0007-6 Revision No. 03 Effective Date: 04/19/2011 Page 54 of 58

Sommers, J. F., 1967, “Change in TRA Neutron Monitoring Methods,” memorandum to Personnel, 

SOM-59-67, Idaho Nuclear Corporation, Idaho Falls, Idaho, November 3. [SRDB Ref ID: 

8110]

Sommers, J. F., 1969, “Neutron Dosimetry,” memorandum to All Reactor HPs, SOM-30-69, Idaho 

Nuclear Corporation, Idaho Falls, Idaho. [SRDB Ref ID: 8144]

Sommers, J. F., 1975, “Personnel Neutron Dosimetry,” memorandum to T. H. Stickley, Som 21-75, 

Aerojet Nuclear Company, March 10. [SRDB Ref ID: 8091]

Thierry-Chef, I., F. Pernicka, M. Marshall, E. Cardis, and P. Andreo, 2002, “Study of a Selection of 10

Historical Types of Dosimeters: Variation of the Response of Hp(10) with Photon Energy and

Geometry of Exposure,” Radiation Protection Dosimetry, volume 102, number 2, pp. 101–113.

Vallario, E. J, D. E. Hankins, and C. M. Unruh, 1969, AEC Workshop on Personnel Neutron 

Dosimetry, September 23 to 24, 1969, BNWL-1340, Battelle Memorial Institute, Pacific 

Northwest Laboratory, Richland, Washington. [SRDB Ref ID: 11096]

Vivian, K. A., DOE-ID and L. E. Rockhold, Bechtel BWXT Idaho, 2003, Transmittal of Radiological 

Control Forms Used at the INEEL, letter to N. Rohrig, Intrepid Technology & Resources, CCN 

44746, September 4. [SRDB Ref ID: 26769] 

Vivian, K. A. (DOE-ID), and L. E. Rockhold (Bechtel BWXT Idaho), 2004, “Transmittal of Sample 

Dosimetry Forms,” letter to N. Rohrig, Intrepid Technology & Resources, CCN 47174, January 

6. [SRDB Ref ID: 26767]

Voelz, G. L., 1969, 1968 Annual Report of the Health Services Laboratory, IDO-12071, U.S. Atomic 

Energy Commission, Idaho Operations Office, Idaho Falls, Idaho, April. [SRDB Ref ID: 1359]

Voelz, G. L., 1970, 1969 Annual Report of the Health Services Laboratory, IDO-12073, U.S. Atomic 

Energy Commission, Idaho Operations Office, Idaho Falls, Idaho, July. [SRDB Ref ID: 1362]

Wages, R., S. Markowitz, S. Kieding, M. Griffon, and M. Ellenbecker, 1998, Former Worker Medical 

Surveillance Program at Idaho National Engineering and Environmental Laboratory (INEEL), 

Phase I: Needs Assessment, October 9. [SRDB Ref ID: 12214]

Walker, D. I., 1974, untitled letter to D. E. Black (Allied Chemical Corporation), U.S. Atomic Energy 

Commission, Idaho Operations Office, Idaho Falls, Idaho, July 23. [SRDB Ref ID: 8128, p. 8]

Warren, S., J. S. Mitchell, A. C. Chamberlain, G. J. Neary, E. F. Edson, G. O. Failla, J. G. Hamilton, 

L. Hempelman, H. M. Parker, K. Z. Morgan, B. S. Wolf, A. Brues, L. S. Taylor, W. Langham, 

D. Hoffman, W. B. Lewis, A. J. Cipriani, G. C. Laurence, H. Carmichael, G. H. Guest, 

E. Renton, G. E. McMurtrie, and A. O. Braaten, 1949, Minutes of the Permissible Doses 

Conference Held at Chalk River, Canada, September 29th-30th, 1949, R.M.-10, Tri-Partite 

Conference, Chalk River, Canada. [SRDB Ref ID: 16640] 

Watkins, V., no date (c. 1966–1968?), NRTS Routine Thermoluminescent Dosimetry System, U.S. 

Atomic Energy Commission, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 

8149]

Watson, J. E., Jr., J. L. Wood, W. G. Tankersley, and C. M. West, 1994, “Estimation of Radiation 

Doses for Workers without Monitoring data for Retrospective Epidemiologic Studies,” Health 

Phys. 67(4):402-405.
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GLOSSARY

1/E spectrum

Typical neutron spectrum for neutrons emitted during prompt fissioning. Spectrum defined by 

the number of neutrons in an energy interval being proportional to the width of the energy 

interval divided by the average energy of the neutrons in that interval.

albedo dosimeter 

Thermoluminescent dosimeter that measures the thermal, intermediate, and fast neutrons 

scattered and moderated by the body or a phantom from an incident fast neutron flux.

alpha particle (α) 

See alpha radiation. 

alpha radiation 

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2. 

anterior–posterior (AP) 

Physical orientation of the body relative to a penetrating directional radiation such that the 

radiation passes through the body from the front to the back. 

backscatter 

Reflection or refraction of radiation at angles over 90 degrees from its original direction.

beta particle (β) 

See beta radiation. 

beta radiation 

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

deep dose equivalent index

Dose equivalent in units of rem or sievert for a 1-centimeter depth in tissue (1,000 milligrams 

per square centimeter). See dose. 

dose 

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rads, reps, or grays. 

dose equivalent (H) 

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose. 

dose equivalent index

Historical measure for neutron source calibration defined by the International Commission on 

Radiation Units and Measurements as the sum of the maximum dose equivalents delivered 

within a sphere at any depth for the respective neutron energies even though the maximum 
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dose occurred at different depths and discounting the outer 0.07-millimeter-thick shell. Also 

called unrestricted dose equivalent index.

DOE Laboratory Accreditation Program (DOELAP) 

Program for accreditation by DOE of DOE site personnel dosimetry and radiobioassay 

programs based on performance testing and the evaluation of associated quality assurance, 

records, and calibration programs. 

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See albedo dosimeter, film dosimeter, neutron film dosimeter, 

pocket ionization chamber, and thermoluminescent dosimeter. 

effective dose equivalent (HE)

Average of the dose equivalents weighted for the susceptibility of harm to different tissues or 

organs in units of rem or sievert. See dose. 

exposure

In general, the act of being exposed to ionizing radiation.

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

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

geometry and composition. Cx factors are a function of photon energy, material geometry 

(e.g., sphere, slab, or torso), and material composition (e.g., tissue-equivalent plastic, soft 

tissue without trace elements, or soft tissue with trace elements). 

film dosimeter 

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

gamma radiation 

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus. 

gamma ray, particle, or photon (γ) 

See gamma radiation. 

linear energy transfer (LET)

Rate of energy absorption by media from particulate or electromagnetic radiation.

neutron (n) 

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron film dosimeter 

Film dosimeter with a nuclear track emulsion, type A, film packet.
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neutron radiation 

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron. 

nonpenetrating dose

Dose from beta and lower energy photon (X-ray and gamma) radiation which does not 

penetrate the skin. It is often determined from the open window dose minus the shielded 

window dose. See dose. 

nuclear track emulsion, type A (NTA) 

Film sensitive to fast neutrons made by Eastman Kodak. The developed image has tracks 

caused by neutrons that become visible under oil immersion with about 1,000-power 

magnification.

occupational dose 

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

open window (OW) 

Area of a film dosimeter that has little to no radiation shielding (e.g., only a holder and visible 

light protection). See film dosimeter. 

pencil dosimeters

See 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)

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeters 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. These are 

noted as Hp(0.07) and Hp(10), respectively. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as dose quantity for radiological 

protection. 

photon radiation 

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays. 

pocket ionization chamber (PIC) 

Cylindrical monitoring device commonly clipped to the outer clothing of an individual to 

measure ionizing radiation. A PIC may be self-reading or require the use of a outside device 

to be able to read the dosimeter. Also called pencil, pocket pencil, pencil dosimeter, and 

pocket dosimeter.
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polymethyl methacrylate

Scientific name for plastic commonly known as Lucite or Plexiglas.

range

Distance an energetic charged particle will go through a material before it stops. Range is an 

increasing function of energy and depends on the elemental makeup of the material and the 

density. 

redact 

To put into suitable form for publication by such processes as editing and revision.

shallow absorbed dose (Ds)

Absorbed dose at a depth of 0.07 millimeters (7 milligrams per square centimeter) in a material 

of specified geometry and composition.

shallow dose equivalent (Hs)

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

thermoluminescent dosimeter (TLD) 

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated by radiation, release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose. 

tissue rad

Absorbed dose in tissue. 

X-ray 

See X-ray radiation. 

X-ray radiation 

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.