ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Idaho National Engineering and 

Environmental Laboratory (INEEL) – Occupational External 

Dosimetry

Document Number:

ORAUT-TKBS-0007-6 

Effective Date: 04/06/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 48

Subject Expert: Norman Rohrig

Document Owner

Approval: Signature on File Date: 04/06/2004

Norman Rohrig, TBD Team Leader

Approval: Signature on File Date: 04/06/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 04/06/2004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 04/06/2004 James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

3

Acronyms and Abbreviations ...............................................................................................................4

6.1 Introduction................................................................................................................................. 6

6.2 Basis of Comparison...................................................................................................................6

6.3 Dose Reconstruction Parameters ...............................................................................................8

6.3.1 Site Administrative Practices ..............................................................................................8

6.3.2 Personnel Monitoring Systems Used at INEEL................................................................. 11

6.3.2.1 Initial Film Badge.................................................................................................... 11

6.3.2.2 Multiple-Filter NRTS Film Badge ............................................................................ 14

6.3.2.3 Original Lithium Fluoride Teflon TLD System ......................................................... 16

6.3.2.4 INEEL ATLAS TLD System.................................................................................... 16

6.3.2.5 Harshaw Two-Chip TLD System ............................................................................ 16

6.3.2.6 Panasonic Four-Chip System................................................................................. 17

6.3.2.7 Nuclear Track Emulsion-Type A for Neutrons......................................................... 17

6.3.2.8 Neutron Albedo Dosimetry ..................................................................................... 18

6.3.3 Calibration........................................................................................................................19

6.3.3.1 Beta-Gamma Radiation.......................................................................................... 19

6.3.3.2 Neutron Calibration ................................................................................................ 19

6.3.4 Workplace Radiation Fields .............................................................................................. 21

6.3.4.1 Gamma Radiation .................................................................................................. 21

6.3.4.2 Beta Radiation........................................................................................................ 23

6.3.4.3 Neutron Radiation .................................................................................................. 25

6.3.4.3.1 MTR Neutron Radiation................................................................................ 27

6.3.4.3.2 TAN Fuel Storage Casks.............................................................................. 30
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6.3.4.3.3 Typical Workplace Neutron Dosimeter Hp(10) Performance ........................ 30

6.4 Adjustments to Recorded Dose................................................................................................. 31

6.4.1 Neutron Weighting Factor................................................................................................. 31

6.5 Missed Dose .............................................................................................................................32

6.5.1 Photon Missed Dose ........................................................................................................ 32

6.5.2 Neutron Missed Dose....................................................................................................... 32

6.5.2.1 Before October 1976.............................................................................................. 33

6.5.2.2 After October 1976................................................................................................. 33

6.6 Organ Dose ..............................................................................................................................34

6.7 Uncertainty................................................................................................................................34

References ........................................................................................................................................36

Glossary ............................................................................................................................................43

Attachment 6A: Interpretation of INEEL Dosimetry Codes ................................................................ 45

Attachment 6B: Occupational External Dose For Monitored Workers ............................................... 47

LIST OF TABLES

Table

6-1 INEEL facility fence direct gamma values (TLD – Background)................................................

Page

9

6-2 INEEL Facility Neutron Correction Factors (FNCF) 1981 ....................................................... 18

6-3 Laboratory sources of uncertainty for beta/photon dosimeter calibration 

parameters.............................................................................................................................20

6-4 Common sources of laboratory bias in the calibration parameters for neutron 

dosimeters .............................................................................................................................21

6-6 Selection of IREP beta and photon energies for INEEL Facilities ........................................... 23

6-5 Beta Dosimeter Thicknesses and Associated Underreporting ............................................... 25

6-7 Typical workplace neutron dosimeter Hp(10) performance..................................................... 30

6-8 Dose equivalent fractions and Q corrections, estimated and recommended........................... 32

6A-1 Area Codes............................................................................................................................45

6A-2 Reasons Codes .....................................................................................................................45

6A-3 Irregularity Codes................................................................................................................... 46

6A-3 Column 20 Codes ..................................................................................................................46

6B-1 INEEL beta/photon dosimeter period of use, type, minimum reporting level, 

exchange frequency, and potential annual missed dose. ....................................................... 47

6B-2 Dosimeter type, period of use, exchange frequency, laboratory minimum 

detectable limit, and maximum annual missed dose............................................................... 47

6B-3 Recommended IREP Neutron Energy Fractions and Correction Factors for 

INEEL Facilities......................................................................................................................48
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LIST OF FIGURES

Figure

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

Page

10

6-2 Film report form used in 1958 ............................................................................................... 11

6-3 Monthly badge reporting form from May 1959....................................................................... 12

6-4 Badge pull results from January 1961 for work in recovery from the SL-1 accident .............. 12

6-5 Special badge report associated with a high beta reading listed in Figure 6-4 ...................... 13

6-6 Personnel exposure questionnaire partially completed for a hypothetical case..................... 14

6-7 Response of Dupont 508 film with various filters to beta and photon irradiation.................... 15

6-8 Probability density of neutron spectrum from an 241AmBe (α,n) source................................. 21

6-9 Gamma and Beta Radiation Field Characterization .............................................................. 22

6-10 Distribution of Beta Ranges .................................................................................................. 24

6-11 Distribution of reportable neutron dose at the INEEL for the first 9 months of 1995 .............. 26

6-12 Neutron Radiation Field Characterization ............................................................................. 27

6-13 Neutron spectra simulating INEEL facilities .......................................................................... 27

6-14 Sample MTR Spectra from Hankins Bonner Measurements................................................. 28

6-15 MTR neutron field components............................................................................................. 28

6-16 Correlation of fast neutron dose equivalent to gamma dose at MTR..................................... 30

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 12/09/03 00-A New technical basis document for the Idaho National 

Engineering and Environmental Laboratory (INEEL) – 

Occupational External Dosimetry. Initiated by Norman 

Rohrig.

Draft 02/24/2004 00-B Incorporates ORAU and NIOSH review comments. 

Initiated by Norman Rohrig.

Draft 03/19/2004 00-C Incorporates additional NIOSH review comments. 

Initiated by Norman Rohrig.

04/06/2004 04/06/2004 00 First approved issue. Initiated by Norman Rohrig.
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 4 of 48

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANSI American National Standards Institute

AP anterior–posterior

Ci curie

cm centimeter

CPP Chemical Processing Plant

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

ERDA Energy Research and Development Administration 

FFTF Fast Flux Test Facility 

FNCF Facility Neutron Correction Factor

ICPP Idaho Chemical Processing Plant (formerly CPP and now INTEC)

ICRU International Commission on Radiation Units and Measurements

ICRP International Commission on Radiological Protection

ID Idaho Operations Office 

INEEL Idaho National Engineering and Environmental Laboratory

INEL Idaho National Engineering Laboratory

INTEC Idaho Nuclear Technology and Engineering Center (formerly ICCP and CPP)

IREP Interactive RadioEpidemiological Program 

in inch

KERMA kinetic energy released to matter

keV kilo electron volt, 1,000 electron volts

LET Linear Energy Transfer

MeV mega electron volt, 1 million electron volts

mg milligram

mm millimeter

mR milliroentgen

mrad millirad

mrem millirem

MRL minimum reporting level

MTR Materials Test Reactor

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurement

NRTS National Reactor Testing Station

NTA nuclear track emulsion-Type A

NVLAP National Voluntary Laboratory Accreditation Program

R roentgen

RBE relative biological effectiveness

rep roentgen-equivalent-physical
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RESL Radiological and Environmental Services Laboratory

RWMC Radioactive Waste Management Complex

Sv Sievert

TAN Test Area North

TLD thermoluminescent dosimeter

TRA Test Reactor Area
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6.1 INTRODUCTION

Technical Basis Documents and Site Profile Documents are general working documents that provide 

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

They will be revised in the event additional relevant information is obtained about the affected site(s). 

These documents may be used to assist NIOSH in the completion of the individual work required for 

each dose reconstruction.

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

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

employer facility” or a “Department of Energy facility” as defined in the Energy Employee 

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 7384l (5) and (12)). 

A branch of the Idaho Operations Office (ID, previously IDO) provided external dosimetry resources 

and services at the Idaho National Engineering and Environmental Laboratory (INEEL) from the start 

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

when the U.S. Department of Energy (DOE) transferred that responsibility to the prime operating 

contractor. Despite the fact that INEEL had several contractors at a time and that contractors 

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

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

records; thus providing a stable external dosimetry system. 

6.2 BASIS OF COMPARISON

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

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

the External Dose Reconstruction Implementation Guidelines (NIOSH 2002) provides conversions 

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

(10);

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

H*

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

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

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

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

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

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

dose corresponds to the shallow dose equivalent. 

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

radiation protection requirements from the 1930s. In 1949, the newly formed National Committee 

(now Council) on Radiation Protection (NCRP) issued NCRP Report 7 [as National Bureau of 

Standards (NBS 1940, p 6) Handbook 42], which recommended a permissible dose of 0.3 R wk-1 (15 

R yr

-1) for occupational workers. The term dose was undefined. Roentgen (R) was defined as the 

quantity or dose of X-rays such that the associated ionization per 0.001293 gram of air (1 cm3 at 

standard temperature and pressure) produces 1 electrostatic unit of charge of either sign. A site 

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

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

quarterly or an annual limit.

In 1953, the International Commission on Radiation Units and Measurements (ICRU 1954) 

established a new quantity, absorbed dose, which is the energy deposited in material per unit mass 

by radiation, using the unit rad =100 erg/g. It specified the term exposure dose, later to become 

exposure, for the ionization capability in air for X- and gamma rays. In 1956, the ICRU defined the 
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term relative biological effectiveness (RBE) dose and the quantity rem (roentgen equivalent in man) 

and the concept of adding all types of external doses together (ICRU 1956). 

In 1957, the NCRP introduced an age prorating formula for the Maximum Allowable Dose of 5 rem × 

[age (yr) -18] (NBS 1958). This introduced 5 rem as an average annual dose but deemphasized it as 

a limit. The AEC issued AEC Manual Chapter 0524 “Permissible Levels of Radiation Exposure” on 

January 9, 1958 which adopted the prorating formula. It retained 15 rem as the maximum annual 

dose and superseded the 13 week whole-body limit of 3 rem with “the provision that not more than 

one-fourth of the 15 rem maximum permissible yearly dose shall be taken in one-fourth of a year”

(AEC 1958). 

The quarterly limit of 3 rem or 12 rem yr-1 replaced the 15 R yr-1 associated with the weekly limit (NBS 

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

AEC Manual Chapter 0524 was reissued in 1963 and reissued later as Energy Research and 

Development Administration (ERDA) Manual Chapter 0524, which provided requirements for radiation 

safety.

In 1957, NBS Handbook 63 (NBS 1957) specified a dependence of the relative biological 

effectiveness (RBE) on the linear energy transfer (LET) of the charged particles that actually deliver 

the dose. NBS used this in the Snyder calculations of maximum permissible neutron flux (NBS 1961) 

and it is still used in the radiation control regulations for DOE.

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

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

The ICRU recommended the quantity kerma in 1962.

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

annual dose limit of 5 rem, eliminating the quarterly limit. In April 1975, ERDA reissued Manual 

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

required adding internal and external dose equivalents if both are known. Monitoring was required 

“where the potential exists for the individual to receive a dose or dose commitment ... in excess of 

10% of the quarterly standard” of 3 rem. Personnel monitoring equipment for each individual was 

required for external radiation. “To achieve optimum accuracy, personnel dosimeters should comply 

with the performance parameters contained in American National Standards Institute standards 

N13.5, N13.7 (ANSI 1983a), and N13/42 WG1 Final draft 1974” (ERDA 1975, App. p 10). Quality 

factors from NCRP Report 38 are specified along with the neutron flux density for 100 mrem in 40 

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

interpolating in energy cannot be accomplished with an instrument. The dose equivalent conversion 

factors reported in ICRP 21 (1973) do not present that problem.

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

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

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

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

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

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

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

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

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

eliminate issues of field direction and nonuniform fields; it also introduced several dose equivalents: 
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ambient dose equivalent, directional dose equivalent, individual dose equivalent penetrating, and 

individual dose equivalent superficial (ICRU 1985).

ICRP Publications 26 and 30 (1977, 1979) introduced new dose limits and the associated quantity 

effective dose equivalent as averaged over the radiation-sensitive organs of the body. 

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

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

Protection for Occupational Workers,” superseded DOE Order 5480.1A, Chapter XI. This order 

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

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

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

neutron flux density.

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

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

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

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

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

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

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

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

mm (0.07 for surface and 10 for deep) from the surface for which the dose is measured. In 1987, 

DOE Order 5480.15, “Department of Energy Laboratory Accreditation Program for Personnel 

Dosimetry,” (DOE 1987) established the DOE Laboratory Accreditation Program (DOELAP) system 

for dosimetry accreditation. In 1986, Standard for the Performance Testing of Personnel Dosimetry 

Systems (DOE 1986a) specified the measurement of deep and shallow dose equivalents at depths of 

10 mm and 0.07 mm, respectively.

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

factor to radiation weighting factor, and generated new factors (ICRP 1990). These factors, invoked 

in NIOSH (2002), depend on neutron energy at the entrance to the body rather than on secondary 

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

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

referenced in the external dose implementation guide (NIOSH 2002).

Thus, the quantities to be measured and reported by the dosimetry systems have been evolving over 

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

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

6.3 DOSE RECONSTRUCTION PARAMETERS

6.3.1

It was INEEL policy that personnel expected to receive any radiation dose or personnel whose work 

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

stored at the respective operational area entrance security gate for INEEL facilities. Control badges, 

which are used to subtract background radiation, have also been located there. This practice may 

lead to subtracting environmental radiation from site activities reducing the reported doses. 

Environmental radiation levels have been monitored for most of the life of the INEEL, originally with 

film badges and later with TLDs. Table 6-1 presents results of this monitoring at facility fence-line 

Site Administrative Practices
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locations near the security gates which can be added to individual’s dose history or used for nonradiation workers working at the site.

Table 6-1. INEEL facility fence direct gamma values (TLD – Background) (mR). 

Year

ARA I & 

II SPERT TAN-TSF

TANLOFT

TANLPT CFA TRA ICPP RWMCEBRII TREAT Backgrnd

1952-72 226 12 42 10 12 52 438 446 32 36 18

1973 86 21 41 17 12 53 306 405 32 37 19 121

1974 162 48 8 7 0 59 320 627 370 35 17 123

1975 114 16 29 11 7 17 195 357 265 32 8 118

1976 66 27 22 15 12 20 140 311 155 56 50 113

1977 41 5 0 4 4 1 137 318 189 22 0 132

1978 52 12 2 9 4 7 143 251 106 56 2 129

1979 63 18 10 17 7 14 159 236 65 59 5 113

1980 65 17 8 19 13 18 251 203 57 51 12 119

1981 63 18 8 17 10 14 231 255 42 28 9 118

1982 50 26 6 12 6 10 163 124 42 20 12 117

1983 78 17 23 26 19 18 174 141 50 24 10 115

1984 80 19 11 19 12 15 205 181 48 31 13 124

1985 80 19 11 19 12 15 205 181 48 31 13 124

1986 80 19 11 19 12 15 205 181 48 31 13 124

1987 80 19 11 19 12 15 205 181 48 31 13 124

1988 80 19 11 19 12 15 205 181 48 31 13 124

1989 80 19 11 19 12 15 205 181 48 31 13 124

1990 80 10 10 11 9 11 28 39 27 19 13 124

1991 80 10 10 11 9 11 28 39 27 19 13 124

1992 80 10 10 11 9 11 28 39 27 19 13 124

1993 77 19 18 23 15 15 48 37 24 28 16 111

1994 69 0 0 0 0 0 24 28 25 15 3 130

1995 91 6 4 8 2 11 31 43 42 17 7 116

1996 52 4 14 0 0 13 28 49 40 22 21 129

1997 46 10 3 8 0 9 29 44 17 16 16 128

1998 62 8 0 5 0 12 25 31 20 0 11 131

1999 49 13 0 0 0 5 10 38 22 13 13 122

2000 28 16 7 16 8 19 40 55 61 25 26 129

2001 31 3 0 0 0 0 27 32 25 0 3 140

2002 41 11 0 0 0 9 34 54 33 18 39 120

Some individuals who might occasionally visit site facilities but did little work with radiation, had 

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

badges issued. Early on at INEEL, the badge change frequency was not the same for everyone. 

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

chance of exposure. Therefore, missed doses should be based on the actual change frequency for a 

person, and the frequency can be determined from the individual’s data package.

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

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

changed. 

DOE provided dosimetry information for a former INEEL worker, whose dose reconstruction is 

underway, should include a dose summary for the employment period and a copy of each weekly, 

monthly, quarterly, etc. form which will also show the work location, so the individual file could be 

several inches thick in hard copy. Each sheet is redacted so only the person of interest’s name and 

applicable information are visible. This file provides the recorded information as to the exchange 
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period for the person for that time period. Figures 6-1 through 6-5 are a partial example set of 

redacted dose reporting forms.

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

rezeroed pencil ionization chambers worn by workers, and entered the weekly badge result on the 

same card. On this sample, on October 28, November 16, and December 9, 1954, the badges were 

Figure 6-1. Individual dose reporting form in use until 1958. . 
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pulled and read in response to high pencil chamber readings. The personnel monitoring badges have 

always been considered more reliable than pencil dosimeters; so after the film badge results became 

available, the daily pencil readings were no longer considered doses of record. However, these 

values can be recovered from the earliest forms for a worst-case estimate of dose. In Figure 6-1, the 

pencil readings totaled 820 mR and the badges reported 0 mR for 18 badges.

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

sensitive open window (OW) result was zero, so the shielded film was not read. On the other two, the 

open window and shielded values were at the minimum recorded density of 0.02 which corresponded 

to a 30 mR penetrating dose. 

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

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

also ionization chambers), the INEEL operational health physics staff would rezero the dosimeters. 

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

reports badge reading for May 1959 when badges were exchanged every two weeks. The column 

under the P of Personnel is an area designator with the code listed under Location at the bottom of 

the page. The next column was unused and dropped somewhat later. The next column is a reason 

code. They are listed in Attachment 6A. Figure 6-4 is a listing of some doses received during

recovery from the SL-1 accident. Workers from several areas were pulled into the accident recovery 

process, and it is notable that one result exceeds the dose limits and that there are few zeros. Figure 

6-5 is a follow-up badge report for one result on Figure 6-4.

When there has been a question about a dose value being assigned to an INEEL worker, a Personnel 

Exposure Questionnaire was normally initiated as shown in Figure 6-6 (shows a hypothetical case).

6.3.2

6.3.2.1 Initial Film Badge

Personnel Monitoring Systems Used at INEEL

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 type 552 film was used 
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Figure 6-3. Monthly badge reporting form from May 1959. . 

Figure 6-4. Badge pull results from January 1961 for work in recovery from the SL-1 accident. 
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Figure 6-5. Special badge report associated with a high beta reading listed in Figure 6-44.

for beta-gamma dosimetry for most locations; DuPont type 558 film (type 508 sensitive and type 1290 

insensitive) was used at two reactor areas. 

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. A calibration curve was used to 

convert the cadmium-shielded portion to penetrating gamma exposure in roentgen. The open window 

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

density and the remainder was converted to beta dose in rep.

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

10 mR. The open window will respond to beta radiation as well as X-rays and low-energy gamma 

rays. Because of the high atomic number (Z) of film in relation to air or tissue, the open window will 

over respond per unit exposure to low-energy photon radiation, as shown in Figure 6-7, by about a 

factor of 30 at 40 keV. Using a cadmium filter with its high Z severely attenuates the photons that get 

to the film, so the over response is reduced to about a factor of 2 at 125 keV. The beta particle range 

is independent of the atomic number Z depending only on the density, so the 1-mm cadmium filter 

(~900 mg/cm2

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

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

film with a silver or cadmium filter. Pencil ionization chambers were used to monitor daily doses and 

control operational activities. The dosimetry group read and recorded these pencil readings on cards. 

Film badge readings were written on the same cards to indicate the dose of record. In 1958, the 

Victoreen 352 pencil ionization chambers being read by the dosimetry group were replaced with 

self-reading dosimeters that were read and rezeroed by the field health physics technicians (AEC 

1959, p 11). Film readings remained the dose of record.
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Figure 6-6. Personnel exposure questionnaire partially completed for a hypothetical case. 

6.3.2.2 Multiple-Filter NRTS Film Badge

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

1 mm cadmium, 0.013 mm. silver, and 0.5 mm aluminum with thicknesses of 950 mg cm-2, 203 mg 

cm-2, and 175 mg cm-2, respectively, including the plastic in which they are mounted (Cipperley 1968). 

This NRTS badge was also a security badge, resulting in an absorber thickness of 100 mg cm-2 in the 

open window, which filtered out beta radiation below 360 keV.
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Figure 6-7. Response of Dupont 508 film with various 

filters to 140 mrem uranium beta and 100 mR of different 

energy photon irradiation provided by NBS. The original 

badge used the open window and cadmium shielded 

films. The multiple filter badge used all three filters plus 

the open window (Cipperley and Gammill 1959). 

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

somewhat determine photon energy. Figure 6-7 shows photon energy dependence of the darkening 

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

(Cipperley and Gammill 1959). With Dupont type 508 film, mixed exposures of radium gamma and 

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

minimum reporting level of 10 mrem was used for both beta and gamma radiation (AEC 1962).

The ID 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 open window, aluminum, and silver filters were 1, 0.2, and 0.1, respectively. By assuming a 

beta signal and subtracting it, the remaining signal could be attributed to low-energy photons and the 

energy could be estimated. For beta other than uranium, the analysis had greater uncertainty.

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

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

weekly schedule continued (AEC 1959). The introduction of punch cards increased the efficiency of 
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 16 of 48

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

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

Experience following the SL-1 accident showed a wide variation of beta-to-gamma ratios and 

necessitated controlling both radiations rather than just the gamma. A set of as many as 18 badges 

could and in many cases was fastened on a belt around the worker to determine a beta:gamma ratio 

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

6.3.2.3 Original Lithium Fluoride Teflon TLD System

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

0.5 rem yr-1 were placed on a lithium fluoride (LiF) disk thermoluminescent dosimeter (TLD) badge, 

which was exchanged quarterly (Cusimano and Cipperley 1968). Two 13-mm-diameter Teflon disks, 

0.4 mm thick (100 mg (75 mg/cm2

) impregnated with 28 mg LiF), were mounted in a badge behind an 

open window and a 1-mm cadmium filter (Watkins). The disks, manufactured by Teledyne Isotopes, 

were read with the Teledyne Model 7300 TLA reader. LiF was chosen because the average Z is 

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

badge could read 30 mR on a quarterly basis, so more small doses were reported. The angular 

dependence of the gamma response (within 10% to 70o) is superior to film because the material acts 

like an ionization chamber. For normal monitoring, only the open window TLD was read and 

considered penetrating dose unless it read more than 125 mrem, in which case the shielded TLD was 

also read. 

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

particularly for individuals who would receive low doses, with longer exchange cycles, typically 3 or 6 

months. In July 1968, the monitoring period was increased from 3 to 6 months (AEC 1969). In 

December 1972, annual processing was used for 1,190 low-dose individual TLDs and 960 were 

processed quarterly (Cusimano 1972). Employees on a monthly badge change were moved to this 

system as late as September 1973.

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

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

earlier calibrations probably used radium.

6.3.2.4 INEEL ATLAS TLD System

Development began in 1969 on a patented Automatic Thermo Luminescent Analyzer System. It used 

LiF in a homogeneous mixture with Teflon and replaced the film in the multi-element 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). 

6.3.2.5 Harshaw Two-Chip TLD System

Several unstable characteristics with the Automatic Thermo Luminescent Analyzer System led to 

rapid implementation of a two-chip TLD system beginning in December 1974 for ICPP, in February 

1975 for the prime contractor at TAN, TRA, etc., and in May 1975 for Argonne. This commercial 

Harshaw system used two LiF TLDs 240 mg cm-2 thick. In 1976, holes were punched in the security 

badges to restore the open window. One chip was covered by 540 mg cm-2 of aluminum and the 

other was under 4 mg cm-2 of Mylar. The aluminum-covered chip provided penetrating dose at a 

tissue depth of nominally 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 
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 17 of 48

beta energy used in calibration. Field calibrations were used to reduce the problem with beta energy 

dependence. The thin aluminum filter (density thickness 350 mg cm-2) allowed higher-energy beta 

radiation to expose the chip used for measuring the penetrating (1,000 mg cm-2) dose. 

The practice was to read only the open window chip to determine if the nonpenetrating dose was 

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

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

6.3.2.6 Panasonic Four-Chip System

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

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

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

filter, an improved measurement of the shallow dose equivalent. A calcium sulfate (CaSO4) TLD 

provides a strong low-energy photon response. The elements are 15 mg cm-2 thick. Element 1 has 

filtration of 16 mg cm-2, element 2 has filtration of 58 mg cm-2 plastic, and elements 3 (CaSO4) and 4 

have filtration of 550 mg cm-2 of plastic and 50 mg cm-2 of aluminum. Although none of the elements 

are at a depth of 7 mg cm-2, the specified depth for the shallow dose equivalent, an acceptable 

response can be obtained by using elements at 16 and 58 mg cm-2. This system is qualified in 

DOELAP beta and photon, and mixture, performance categories.

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

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

gamma and 30 mrem for beta until 1993 (Perry, Andersen, and Ruhter 1993), when it returned to 10 

mrem gamma.

6.3.2.7 Nuclear Track Emulsion-Type A for Neutrons

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

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

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

grains of emulsion.

The minimum dose assigned was 14 mrem. 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 (MRL = 30 mrem). A blank is recorded for 17 people on 

the data sheet, probably because they were not monitored for neutrons or the film was not read. Only 

the two individuals received measurable gamma doses. A person on weekly exchange from January 

to March 1958 received neutron dose equivalents of 14, 28, 42, and 14 mrem and gamma doses of 

130, 70, 30, 30, 50, 30, and 20 mrem with both neutron and gamma doses received twice. Thus data 

sheets support the MRL of 14 mrem.

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

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

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

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

January 1962 a data sheet shows a 20 mrem dose. A data sheet from April 1959 shows a neutron 

dose equivalent of 20, 20, and 40 mrem. These values suggest an MRL of 20 mrem.
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 18 of 48

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

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

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

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

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

to read NTA film occurred if the hard spectra neutron exposure was likely to exceed 10 mrem. 

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 result were several designs using the 

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

Li in a TLD. 

In the Hankins dosimeter used at the INEEL (Hankins 1973), TLDs (6

Li to capture thermal neutrons) 

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

thermal neutrons from outside) and 7

Li TLDs are used to subtract the gamma dose. 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). From the ratio of the dose equivalent measured with a 9-in.-diameter Eberline PNR-4 

and the corresponding signal (reads in mrem, but not dose equivalent) with the detector in the 

3-in.-diameter PNR-4 insert, one can generate an FNCF that converts the TLD gamma equivalent 

signal to neutron dose equivalent (Hankins 1976). Values of the FNCF as shown in Table 6-2 

(Cusimano 1981) were measured for different fields at INEEL, were tabulated for assigning the dose 

equivalent from the badge results, and were routinely updated. 

Table 6-2: INEEL Facility Neutron Correction Factors (FNCF) 1981. 

Organization FNCF Organization FNCF

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

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

ANL-TREAT 1.05 EG&G-LOFT 3.5

ANL-ZPPR 0.94 EG&G-ARA III 2.0

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

EG&G-TRA (SA) 2.7

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

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

record suggests it occurred near the end of 1974 or early 1975 (Ruhter and Perry 2002; Gesell 1996), 

but an informal list of “Dosimetry Branch Changes” from 1978 (Dosimetry 1978) states “initial testing 

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

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

should make the claimant-favorable assumption that this transition occurred in October 1976.
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6.3.3

6.3.3.1 Beta-Gamma Radiation

Calibration

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

used to measure radiation fields (AEC 1959, p. 132). Uranium metal bars 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 (AEC 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 area, 2.5 cm2

, was constructed by the Amersham Searle Corporation in 

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

following the Three Mile Island TMI-2 reactor accident in 1978. TLD badges were calibrated to 500 

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

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

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

Separation of penetrating dose from non-penetrating dose was an issue in 1957 (Bennett 1957) and 

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

unusual. 

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

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

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

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

a phantom backing. To convert from exposure in roentgen to dose equivalent index in rem, a 

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

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

1984).

In 1989, the INEEL dosimetry service transferred from RESL to EG&G, Inc., the prime contractor. 

Calibrations continued to use DOE RESL sources and no changes were made to the dosimetry 

system. The 1991 Tiger Team Review of the INEEL site indicated that the INEEL contractor and the 

Idaho Operations Office using the same sources for calibration led to a conflict of interest or an 

advantage in DOELAP tests. As a result, EG&G purchased a Shepherd panoramic irradiator with a 

137Cs source for badge irradiations. This irradiator does not use a phantom, but was cross-referenced 

using many TLD irradiations to the DOE source using a phantom (Anderson 1994). In addition, the 

contractor developed and characterized their own uranium slab for beta irradiations (Bean 1995).

Table 6-3 lists common sources of laboratory bias for personnel beta/photon dosimeter calibration 

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

6.3.3.2 Neutron Calibration

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

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

polonium interact in the reaction 9

Be(α,n)12C, and generate a broad spectrum of neutrons up to about 
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 20 of 48

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

Parameter Historical description Uncertaintya Comment

In-air 

calibration

In 1981, INEEL began exposing 

calibration dosimeters on 

phantoms (used to simulate 

worker body). Previous 

calibrations do not include 

response from radiation 

backscatter response.

+10% Recorded dose of record too high. 

Backscatter radiation from worker body 

is highly dependent on dosimeter 

design. 

Radiation 

quantity

Before 1981, INEEL dosimeter 

systems were typically calibrated 

to a photon beam measured as 

exposure. 

-5% For higher energy 226Ra and 137Cs 

gamma radiation used to calibrate 

dosimeters, this caused a slight (about 

3%) under response in recorded dose. 

Tissue depth 

of dose

Historically, INEEL used an 

unspecified depth to estimate the 

deep dose.

±5% The numerical effect of this for photon 

radiation is comparatively low. INEEL 

dosimeter designs had filtration density 

thickness of about 1,000 mg/cm2 that 

would relate closely to the 1-cm depth in 

tissue.

Angular 

response

INEEL dosimeter system is 

calibrated using 

anterior-posterior (AP) laboratory 

irradiations.

> 300 keV, 

~20%

Recorded dose of record likely too low 

since the dosimeter response is usually 

lower at non-AP angles. Effect is highly 

dependent on dosimeter type, radiation 

type, energy, and angle. 

Environmental 

stability

INEEL film dosimeter and TLD 

systems are subject to signal 

fade with time, heat, humidity, 

light, etc.

±10% Recorded dose of record depends 

strongly on dosimetry parameters such 

as when calibration dosimeters were 

irradiated and processed. Mid-cycle 

calibration minimizes effects. 

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

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

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

Kluge and Weiss (1982) calculate conversion factors of 3.51 to 3.76 × 10-8 rem cm2 n-1 depending on 

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

AmBe of 3.8 × 10-8 rem cm2

 n-1 for the maximum average dose equivalent. A dose equivalent of 

1.5 rem required 3.6 × 107 n cm-2 (Cusimano 1963), corresponding to a dose conversion factor of 

4.17 × 10-8 rem cm2 n-1 so the recorded dose will be about 11% high. Monte Carlo calculations for 

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

 n-1 averaged over the 0- to 2-cm 

shell on a 30-cm-diameter cylindrical phantom (NCRP 1971b). Use of the 50-Ci AmBe source 

continued until 1993.

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

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

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

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

Several common sources of expected laboratory bias are discussed in Table 6-4 for personnel 

neutron dosimeters based on comparison of the recorded dose with Hp(10).
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 21 of 48

Figure 6-8. Probability density of neutron spectrum from an 241AmBe (α,n) source (Kluge and Weiss 

1982). 

Table 6-4. Common sources of laboratory bias in the calibration parameters for neutron dosimeters.

a

Parameter Historical description Anticipated laboratory biasb

Source energy 

spectrum

In 1976, INEEL began using dosimeters that were 

calibrated on a phantom to simulate a worker’s 

body. The previous calibrations did not include 

response from backscattered radiation. 

NTA film tends to be insensitive to albedo 

neutrons, so probably had minimal effect.

Radiation quantity Neutron dose quantities that were used to 

calibrate INEEL neutron dosimeters have varied 

historically. The first collision dose for fast 

neutrons and a quality factor of 10 was used for 

many years. 

As noted above, NTA calibration would result

in the reported dose being about 11% high. 

The effects of the respective neutron dose 

quantities used to calibrate INEEL dosimeters 

is uncertain and could be evaluated in 

comparison to the Hp(10) dose used in 

DOELAP performance testing. 

Angular response INEEL dosimeters are calibrated using 

anterior-posterior (AP) laboratory irradiation.

Recorded dose of record is likely too low 

because the dosimeter response is lower at 

non-AP angles. The effect is highly dependent 

on neutron energy.

Environmental 

stability

INEEL NTA film and TLD dosimeters are subject 

of signal fade with time, heat, humidity, light, etc.

Recorded dose of record is likely too low; 

however, this depends strongly upon when the 

calibration dosimeters are irradiated during the 

dosimeter exchange cycle. Mid-cycle 

calibration minimizes the effects. 

a. Judgment based on INEEL dosimeter response characteristics.

b. Recorded dose compared to Hp(10).

6.3.4

6.3.4.1 Gamma Radiation

Workplace Radiation Fields

In response to a Tiger Team finding, radiation fields at the INEEL have been characterized by making 

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

comparing the results (Reilly 1998). The percentage bias is shown in Figure 6-9 for the beta and 

gamma measurements. Most results lie within +27% to -43 %. 
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The high gamma bias results are for locations at RWMC looking at skyshine (back scattered and thus 

low energy photons) from low level waste in the Subsurface Disposal Area (SDA). The doses 

measured with NaI(Tl) were low (6 and 11 mrem), and the threshold energy on the NaI(Tl) detector 

was about 100 keV so some low energy photons are likely to have been missed. 

The radiation fields at INEEL, with a few exceptions, are generated primarily by mixed fission and 

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

than 250 keV. The INEEL dosimeters are judged to measure these fields well. 

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 would predominantly contain 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.

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

easily shielded so the fields are usually low.

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

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

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

roof shielding so dose rates of <10 mrem hr-1 are possible, usually for short periods. The scattered 

photons will 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 

worksite and some from 300-keV 192Ir or 1.25-MeV 60Co.

Figure 6-9: Gamma and Beta Radiation Field Characterization

0

1

2

3

4

5

6

7

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

-45 to-35%

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

-15 to -5%

-5 to 5%

5 to 15%

1'5 to 25%

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

95% to 105%

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

Relative Bias [(Dosimeter-Spectrometer)/Spectrometer]

Number

Gamma

Beta
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Essentially all INEEL radiological work areas involved beta/photon radiation covering a wide range of 

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

Table 6-6. Selection of IREP beta and photon energies for INEEL Facilities. 

Process/

buildings Description

Operations Radiation 

type

Energy 

Begin End group (keV) Percentage

Reactors Highly dispersed fields of higher energy photon radiation fields 

from fission process, activation and fission product nuclides. 

Potential for significant airborne nuclides, and there might be 

significant higher energy beta radiation

Beta >15 100

Photon 30–250 25

> 250 75

PBF, TRA, ARA, TAN, EBR, ANL, SPERT

1952 2003

Processing 

plants 

Highly dispersed fields of higher energy photon radiation fields 

from activation and fission product nuclides dominant to most 

exposure profiles. Potential for higher-energy beta radiation 

during sampling and maintenance work resulting from fission 

products. 

Beta > 15 100

Photon 30–250 25

> 250 75

ICPP 1952 2003

Calibrations Calibration of instruments and dosimeters Beta > 15 100

CF 633, 636 1952 2003 Photon 30 – 250 25

> 250 75

Waste 

handling

Radiation characteristics are highly dependent on source of 

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

dominant. For TRU waste from Rocky Flats contains 241Am with 

59 keV photon.

Beta > 15 100

Photon 30–250 25

> 250 75

RWMC, WERF 1953 2003

Uranium 

handling

Produced special armor from depleted uranium. Primarily beta 

radiation from U-238 daughters. Some gamma from 

contaminants and Cs-137 sources used in process.

Beta > 15 100

Photon 30 – 250 90

> 250 10

SMC 1985 2003

6.3.4.2 Beta Radiation

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

outside of a container such as a spill or only lightly shielded or in hot cells. High beta fields were not

unusual at the INTEC where large quantities of fission products exist. Pure high energy beta fields in 

some locations, particularly in the exhaust stream, have caused dosimetry problems because the 

badge shielding or instrument packages did not provide a full 10 mm tissue equivalent coverage and 

thus beta fields would be measured as gamma fields.

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

for sources at contact or at 1 cm resulting in hard to reproduce geometry. The low bias beta results 

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

occupational radiation fields (only 3) have a bias better than 15%.

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

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

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

betas from either uranium or strontium. The dose from low energy betas will be missed altogether if 

the beta energy is not sufficient to penetrate the detector cover and will be underreported if the beta 

energy is not sufficient to penetrate the entire detector. The mean beta energy for the spectrum from 

a particular nuclide is about 1/3 of the maximum beta energy for that nuclide. 
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Based on the range energy curve for beta particles and the beta energy distribution of beta emitters 

(HEW 1970, p 90, 91,123) the fraction of beta radionuclides with ranges greater than the abscissa is 

plotted on the ordinate in Figure 6-10. The beta nuclides varied from location to location and time to 

time at the INEEL so a correction factor common for all facilities was estimated. We use the entire 

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

variety of radionuclides used at the INEEL. To reflect that the beta spectrum is not monoenergetic 

because of the energy carried off by the neutrino, a curve is presented for the mean energy or 1/3 of 

the maximum energy. To reflect that some beta particles enter the detector at an angle, a curve is 

provided for 45o incidence at the maximum energy and the mean energy. These curves of the fraction 

of nuclides with a larger range essentially show the depth dependence of beta dose, because the 

energy loss of electrons does not have much energy dependence. These curves also demonstrate 

why early dosimeters with thick sensitive elements failed to correctly report the beta dose at a depth 

of 7 mg cm-2, a depth that was chosen in the early 1980s. These curves also demonstrate why the 

beta dose 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 cm-2, and for most of us at depths greater than 1 cm.

Figure 6-10: Distribution of Beta Ranges

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

 

To calculate the fraction of dose missed by a dosimeter, you simply need 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 cm-2. The appropriate curve should be the curve of the mean range for the beta spectrum and the 

angular distribution of the radiation exposure. To estimate this we have chosen to add the mean 

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

the mean energy and to add in 1⁄2 that value for 45o incidence for the maximum energy. The curves 

are the result of a polynomial trendline to the data, so averaging the fraction of radionuclides is 

relatively easy.
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Table 6-5 provides the cover and detector thicknesses for the beta badges used at the INEEL. 

Thicknesses with a “~” are estimated. The fraction of beta dose measured shown in Table 6-5 is the 

average as described above. To determine the corrected beta dose, divide the nonpenetrating result 

from the dosimetry system by the values in the last column of Table 6-5. 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.

Table 6-5: Beta Dosimeter Thicknesses and Associated Underreporting. 

Dosimeter system Time period Covers

(mg cm-2)

Detector thickness 

(mg cm-2)

Estimated Dose 

Recorded

Two filter film 1951-1958 ~50 ~50 0.49

Multi-filter film 1958-1974 100 ~20 0.35

Low dose TLD 1969-1974 100 75 0.33

ATLAS 1974-1975 100 ~100 0.30

Harshaw TLD 1975-1976 104 344 0.21

Harshaw TLD 1976-1985 4 240 0.41

Panasonic TLD 1986-present 16 15 0.86

6.3.4.3 Neutron Radiation

Most INEEL workers are not exposed to neutrons and so are not badged to measure neutrons. 

Neutron fields at the INEEL are specific to a few facilities. The high dose locations where most of the 

gamma and beta dose is received, such as the ICPP and SL-1, do not have associated neutron dose.

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

be evaluated out of 2900 film-badged employees and more than 3000 personnel monitored with TLD 

badges (AEC 1969).

For calendar year 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) compared to about 

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

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

use since 1975 see all neutron fields. A FNCF determined from the 9- to 3-in. ratio in the worker 

location is used to adjust the measurement result to dose equivalent.

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

nearly the same time a ROSPEC neutron spectrometer (Reilly 1998). The relative biases [(DosimeterSpectrometer)/Spectrometer] for the neutron fields are shown in Figure 6-12. These results show a 

greater dispersion than the gamma results. 

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

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

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

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

values (0.94 and 0.71) are for a waste drum which was reanalyzed and a new 9”/3” ratio determined 
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 26 of 48

because of the unsatisfactory initial result. The report suggests that other waste barrels may have 

had neutron sources causing interference. The remaining bias values lie between -0.16 and 0.44.

Figure 6-11. Distribution of reportable neutron dose at the 

INEEL for the first 9 months of 1995. Of 1461 dosimeters, 

1407 were below the 15 mrem reporting level. 

Fig. 6-12: Neutron Radiation Field Characterization

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

Total number of dosimeters with reportable dose = 54 Total number of dosimeters with reportable dose = 54
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 27 of 48

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 works 

reasonably well. Use of small 252Cf sources for research began after albedo badge use began. 

Figure 6-13 provides spectra for the AmBe (Kluge and Weiss 1982) and 252Cf (fission) neutron 

sources and the 14-MeV neutron generator as seen through 10 cm of polyethylene shielding (Ing and 

Makra 1978) typical of the INEEL facilities.

 

0.0001

0.001

0.01

0.1

1

0.000001 0.00001 0.0001 0.001 0.01 0.1 1 10 100

Neutron Energy (MeV)

Number neutrons

Fission

14 MeV

AmBe

Figure 6-13. Neutron spectra simulating INEEL facilities. 

Most of the reactors built at the INEEL had no beam ports. Thus 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 will be attenuated much more quickly in concrete or water than will the gamma fields. This is 

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

Thus neutron fields are generally not a problem at an enclosed reactor.

6.3.4.3.1 MTR Neutron Radiation

The exception to the above discussion is the MTR, which operated from 1952 to 1970 and had beam 

ports and neutron beams extending onto a research floor. Also in this category are the Zero Power 

Physics Reactor (ZPPR) and Transient Reactor Test (TREAT), both at Argonne West. Some neutron 

surveys of the MTR experimental floor have been recovered (Sommers 1959, 1962; Hankins 1961), 

but these individually do not provide all components of the radiation field. Hankins (1961) used 2-, 3-, 

and 8-in. polyethylene Bonner balls in a cadmium shield to characterize the intermediate and fast 

neutrons at 21 locations around the MTR floor. He made these measurements and also measured the 

thermal neutron component at six other locations. The Hankins data have been reanalyzed (Rohrig 

2004) using more recent Bonner response curves (Hertel and Davidson 1985). Figure 6-14 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 6-15 shows the 

correlations of the thermal and intermediate neutron dose equivalents to the fast neutron dose 

equivalent for the Rohrig reanalysis of the Hankins data. The trend line for the reanalyzed 

intermediate energy neutron dose equivalent has a R2 value of 0.86 and 0.92 as compared to a R2 

value of 0.50 for the original analysis demonstrating a better fit to the data. The average ratio of 

thermal to fast neutron is 0.071 ±0.025, for the low energy intermediate to the fast is 0.177±0.057, and 
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 28 of 48

for the higher energy intermediate to the fast is 0.149 ±0.046 where fast neutrons are taken as those 

above 0.2 MeV. 

10

100

1000

10000

100000

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

Energy (eV)

Number Neutrons

Position 23

Position 3

Figure 6-14. Sample MTR Spectra from Hankins 

Bonner Measurements. 

 

y = 4.633x

R2

 = 0.501

y = 0.148x

R2

 = 0.914

y = 0.049x

R2

 = 0.701

y = 0.115x

R2

 = 0.852

0.01

0.1

1

10

100

0.01 0.1 1 10

Fast Neutron Dose Equivalent Rate (mrem/hr)

Dose Equivalent Rate (mrem/hr)

InterHankins

Inter .5 eV-10 keV

Inter 10-200 keV

ThermalNeut

Figure 6-15. MTR neutron field components.

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

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

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

to 66% depending on the location. The dosimetry record location code for the TRA was 4 (later 40 to 

45). To correct for missed dose on the MTR experiment floor, the NTA results from MTR should be 
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 29 of 48

multiplied by 2 ± 0.2 (1/0.52, 0.08/0.52) for a Monte Carlo dose reconstruction or 3 (1/0.35) for the 

less accurate worst case reconstruction.

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

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

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

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

insignificant for several of the measurements. These values are shown with the triangles at one-half 

of the smallest measured value. Using the Shapiro-Wilks Normality Test (Gilbert 1987) and including 

the insignificant fast neutron values at 1⁄2 of the minimum reported value suggests that the normal 

distribution is a slightly better description of the data than a log-normal distribution. The fast neutron 

dose equivalent is 0.42 ± 0.35 of the gamma dose rate for this data set. Combining these results, one 

can conclude that the total neutron dose equivalent is 0.58 ± 0.48 of the gamma dose equivalent on 

the MTR Experimental Floor. 

48.058.0)149.0177.0071.01)(35.042.0(

1( )

±=+++±=

= + ++

Fast

HiInter

Fast

LoInter

Fast

Thermal

Gamma

FastN

GammaDose

TotalNeutronDose

 (Eq. 1)

The variation within the different components of neutron dose rate are so much smaller than the 

variation between the fast neutron and gamma dose equivalent rate as to be unimportant. 

Figure 6-16: Correlation of Fast Neutron Dose Equivalent 

to Gamma Dose at MTR

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

Many of the people wearing NTA film would receive gamma dose at locations other than on the MTR 

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

often be covering jobs with only beta gamma fields. A craftsperson may service pumps carrying 
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 30 of 48

radioactive water and not receive any neutron dose. Therefore, simply multiplying the gamma dose 

by 1.6 (i.e. 1+0.58) or 2.1 (i.e. 1+1.06) although claimant favorable is probably inappropriate

6.3.4.3.2 TAN Fuel Storage Casks 

As noted in the INEEL Site Description TBD (section 2.2.1.3), fuel storage casks are located on a 

storage pad at TAN. The dose rates are 25 to 30 mrem hr-1 gamma and 40 mrem hr-1 neutron. The 

metal cask attenuates the gamma radiation, but doesn’t appreciable affect the neutron field. While

the loaded casks were being temperature tested in 1985 and 1986, they were located in the TAN 

Warm Shop which has offices nearby at one end on the second floor. Neutron radiation levels were 

discovered in the offices, and the people (about 6) in these offices were not wearing albedo neutron 

badges, but were wearing beta/gamma dosimetry. The casks were each in the area about 2 weeks

while temperature measurements were made before they were moved outside onto the storage pad. 

The casks contain irradiated oxide fuel so the 18O(α,n) reaction generates neutrons in the few MeV 

energy range. 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 EPRI 

documents (EPRI 1986, EPRI 1987a, EPRI 1987b) the temperatures were measured from 9/11 to 

9/23/85, 1/14 to 2/6/86, and 6/2 to 6/27/86. The office most affected was for an operations person. 

Also affected were safety- and radiological- engineers. The estimated dose equivalent for full time 

occupancy is less than 50 mrem for each cask evolution.

6.3.4.3.3 Typical Workplace Neutron Dosimeter Hp(10) Performance

Table 6-7 summarizes typical neutron personnel dosimeter parameters important to Hp(10) 

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

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

Table 6B-2 summarizes the locations at the INEEL where neutron dose is credible.

Table 6-7. Typical workplace neutron dosimeter Hp(10) performance.

a

Parameter Description Potential workplace biasb

Workplace neutron 

energy spectra

NTA dosimeter response decreases and 

TLD response increases with decreasing 

neutron energy 

Depends upon workplace neutron 

spectra. NTA recorded dose of record 

likely too low because of high 500-keV 

threshold for detection of neutrons. 

Exposure geometry NTA dosimeter response increases with 

increasing exposure angle and TLD 

response decreases with increasing 

exposure angle.

NTA recorded dose likely too high since 

dosimeter response is higher at angles 

other than AP. TLD recorded dose is 

lower at angles other than AP. Effect is 

highly dependent on neutron energy. 

Missed dose Doses less than Minimum Reporting

Limit (MRL) recorded as zero dose.

Recorded dose of record is likely too low. 

The impact of missed dose is greatest in 

earlier years because of the higher MRLs 

and shorter exchange cycle of the 

neutron dosimeters. 

Environmental effects Workplace environment (heat, humidity, 

etc.) fades the dosimeter signal.

Recorded dose of record is likely too low.

a. Judgment based on INEEL dosimeter response characteristics.

b. Recorded dose compared to Hp(10).
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6.4 ADJUSTMENTS TO RECORDED DOSE

6.4.1

All dose equivalents measured at the INEEL and reported in this document used the quality factors 

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

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

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

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

Neutron Weighting Factor

A correction needs to be made to the INEEL reported data to change from dose equivalent (pre-ICRP 

60) to a newer dose quantity (ICRP 1990; NIOSH 2002). ICRP 74 (1996) tabulates the ambient dose 

equivalent (dose equivalent at 10 mm depth in a 30 cm diameter sphere) for neutrons. The ratios of 

organ dose equivalents to ambient dose equivalent are tabulated in the NIOSH External Dose 

Reconstruction Implementation Guide (NIOSH 2002), so this quantity will be used for the conversion. 

Ambient dose equivalent is an ICRU quantity so it uses a revised Q(L) rather than a wR so the 

correction factors are not as large as in other TBDs.

The dose equivalent for a spectrum of particle energies is the result of an integral of the fluence 

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

over the range of energies considered:

.)()(

2

1

H DECF EE dE

E

E

∫ = φ (Eq. 2) 

Error! Bookmark not defined.

For ambient dose equivalent, the same equation applies except that a tabulation of the ambient dose 

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

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

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

the correction factors. 

These factors are incorporated into statements of dose equivalent 

values and calibrations following generally accepted principles. The conventional dose conversion 

factors are most clearly and correctly stated in ICRP Publication 21 (ICRP 1973). NCRP Report 38 

tabulates a neutron flux density associated with the annual dose limit that is proportional to the 

reciprocal of the dose conversion factor (NCRP 1971b). Conventionally, the primary geometry 

considered is from one direction with the maximum dose in the body tabulated. More recent 

references (ICRU 1985; ICRP 1987, 1996) consider the dose to individual organs for different 

irradiation geometries, so the more recent tabulations give results lower by factors up to about 10 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 work best for an anterior-posterior (AP) irradiation geometry 

with the dosimeter on the front of the body.

Table 6-8 lists the calculated fractions of dose equivalent in the IREP energy groups and the 

conversion factors from dose equivalent to equivalent dose for INEEL spectra. The ratios of average

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

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

for the fission and 14 MeV source than for the reactor spectrum. The lower part of the table lists the 

recommended default values for the dose equivalent fractions and quality factor corrections.
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Table 6-8. Dose equivalent fractions and Q corrections, estimated and recommended. 

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

Spectrum Calculated Values

Dose equivalent fractions

Bare fission 4.4E-05 0.20 0.80

AmBe 0.15 0.85

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

MTR exp floor ave 0.18 0.06 0.49 0.28

MTR exp floor max 0.24 0.08 0.52 0.35

MTR exp floor min 0.13 0.03 0.46 0.19

ICRP 74 H*10 /NCRP 38 H

Bare fission 1.46 1.32 1.09

AmBe 1.41 1.05

14 MeV 10 cm poly 0.69 1.47 1.36 0.93

MTR exp floor ave 0.86 1.08 1.33 1.12

MTR exp floor max 0.80 1.08 1.37 1.12

MTR exp floor min 0.92 1.08 1.30 1.12

Recommended defaults

Dose equivalent fractions

14 MeV 10 cm poly 0.05 0.95

Source calibrations 0.20 0.80

MTR exp floor 0.2 0.05 0.50 0.25

H*(10)/H 1 1.1 1.4 1.1

6.5 Missed Dose

6.5.1

Missed photon dose for INEEL workers would occur where (1) there is no recorded dose because 

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

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

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

MRLs were higher and dosimeter exchange was more frequent. One option to calculate the missed 

dose described in NIOSH (2002) is to estimate a claimant-favorable maximum potential missed dose 

where MRL/2 is multiplied by the number of zero dose results. Table 6B-1 lists the potential missed 

photon dose according to year, dosimeter type, and badge exchange frequency. The minimum 

reporting levels (MRL) shown are based on Cipperley (1958, 1968) and Cusimano (1963) for film; 

Kalbeitzer (1983), Gesell (1986, 1992), and Perry et al. (1993) for TLDs; and Ruhter and Perry (2002) 

for film and TLD. The exchange frequency will need to be determined from the dose submittal 

package for each individual and year because it was shorter for highly exposed individuals and longer 

for those with lower doses. 

Photon Missed Dose

6.5.2

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

short time at TRA and TAN, small sources used for research at TRA and IRC, calibration laboratories 

(CF 633 and 636), and for calibration of criticality alarms (TRA and ICPP).

Neutron Missed Dose

For other locations, there is likely limited missed neutron dose because of the very low potential for 

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

worked near neutrons and which category of neutrons. This can best be done by looking for the work 
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 33 of 48

location and whether a worker or others in the badge reporting group were assigned any neutron dose 

equivalent. The work location code for TRA where the MTR operated is 4 (also 40 to 45). If no 

neutron dose was assigned to him/her or to coworkers for several months, the dose reconstructor 

should assume that the person was not exposed to neutrons so no neutron dose would be missed. 

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

equivalent using Table 6B-3 for the times when workers did not have reported neutron dose. For the 

period when NTA film was used, the dose should be multiplied by 1.25 for all facilities except the MTR 

experimental floor and by 2 for the MTR experimental floor when the MTR was operating between 

1953 and 1970. Then the dose equivalent is apportioned into the IREP groups using Table 6B-3. 

For example, if in 1955 a person was an experimenter at the MTR, and 7 of the weekly badges 

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

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

about one-half the MTR neutron dose equivalent (from Section 6.3.4.3.1), the total dose equivalent is 

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

dose, multiply the total dose equivalent by the last column of Table 6B-3 to get 200 mrem to the 

<10-keV group, 60 mrem to the 10- to 100-keV group, 700 mrem to the 0.1- to 2-MeV group, and 

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

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

before 1975 when only NTA film dosimeters were used with supplemental recording of thermal 

neutron doses from B-10 pencil dosimeters. The second period is after 1975 when only Hankins 

albedo dosimeters were used. The estimated MRLs for these neutron dosimeters are summarized in 

Table 6A-2. It is possible to estimate the missed neutron dose using the MRLs because the neutron 

dosimeters were calibrated with neutron sources that had energies similar to those encountered in the 

workplace and most of the neutrons to which workers were normally exposed had energies greater 

than the 500- to 800-keV threshold of the NTA film dosimeters. There was, of course, no threshold 

energy for the measurements using neutron albedo TLD badges.

6.5.2.1 Before October 1976

The use of NTA films for neutron dosimetry prior to 1976 is documented in various INEEL reports 

(Cusimano 1963; Cipperley 1958, 1968). As noted above, it is possible to estimate the missed dose 

using the MRLs. There are a lot of recorded zeros in the neutron dose data for INEEL workers for two 

reasons: (1) an NTA film was developed and not read per standard criteria, or (2) an NTA film 

indicated a neutron dose equivalent that was less than the film’s MRL, 14 mrem. When the MRL for 

NTA film is used in estimating the missed neutron dose, it should be multiplied by 1.25 for most 

workers and by 2 for workers on the MTR Experimental Floor. 

It is also possible to estimate the missed neutron dose in some facilities through neutron-to-photon 

dose ratios (NIOSH 2002). However, for the INEEL facilities there are several other sources of 

gamma exposure with no associated neutron exposure, so this would be erroneous.

6.5.2.2 After October 1976

Since October 1976, the neutron dose has been measured using the Hankins albedo-type TLD. The 

characteristics of this dosimeter are well documented (Gesell et al. 1996), and the MRL to be used in 

estimating missed dose is 15 mrem. A location-specific FNCF has been applied to convert the 

reading to dose equivalent, so no additional adjustments should be required.
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For a couple months in the mid 1980s a few office workers without neutron dosimetry were exposed 

to neutrons at TAN (see 6.3.4.3). For these individuals, 50 mrem of neutron dose should be assigned 

for each period they were so exposed.

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 implementation guidelines (NIOSH 

2002). Consistent with OCAS/ORAUT agreements, it is recommended that the AP (front to back) 

geometry should be assumed for the irradiation geometry and for conversion to organ dose. The 

calculated neutron doses in each energy group should be multiplied by the conversion factors from 

ambient dose equivalent to organ dose for AP irradiation from Appendix B of NIOSH 2002. For 

photons prior to 1981 the conversion factor from exposure to organ dose should be used. For 1981 

and after, the conversion factor from deep dose equivalent to organ dose should be used.

6.7 Uncertainty

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

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

INEEL used ionization chambers standardized by NBS (now NIST) for their calibrations. Use of a 

phantom for dosimeter irradiation began in the early 1980s, but backscatter only causes a minor 

change for high energy photon dosimetry. The over response of the multi-element film badge to deep 

dose in tissue is because they were calibrated to exposure which is somewhat more at low energies 

than the deep dose. The INEEL environment did not have a significant low energy photon field like a 

plutonium finishing plant so the nonpenetrating component was attributed to beta radiation. A realistic 

estimate of total uncertainty for photon dosimetry is about 35% at one sigma. This is roughly 

consistent with the results presented in Figure 6-9. For those measurements, the standardization 

instrument contributed some significant uncertainty.

For beta radiations when considering skin cancer, one must divide the reported nonpenetrating dose 

by the fraction of dose measured as reported in Table 6-5. The uncertainty for beta radiation is 

somewhat larger, an estimated 50% at one sigma. This is driven by uncertainties in the field 

geometry and in the fact that beta radiation is often stopped by thin materials such as clothing and air. 

Algorithms are used to estimate the dose at a depth of 7 mg cm-2 from dosimeters at depths of 15 to 

250 mg cm-2 and such depth differences can change the signal significantly. The difference between 

a point source irradiation and a planar source can confuse an algorithm. 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 used before 1975 did not see 

low energy neutrons below 0.5 to 0.8 MeV. 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 

people is primarily due to hydrogen recoils and not due to 6

Li(n,α) reactions. The response of the 9” 

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 about 60% 

at one sigma. The cause of the greatest uncertainty for neutrons is the variation of dose caused by 

an organs position in the body. For 1 MeV neutrons, the dose facing the source is about a factor of 

1000 higher than the dose on the back side of a 30 cm diameter sphere of tissue like material. In a 

work environment, the direction of the neutrons may 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). Note that in 
Effective Date: 04/06/2004 Revision No.00 Procedure No. ORAUT-TKBS-0007-6 Page 35 of 48

Figure 6-11 the discrepancy where the dosimeters are reporting about 1⁄2 of the spectrometer result is 

because the spectrometer does not simulate the attenuation of the body so it read high by a factor of 

two.
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Testing and Analysis, Palo Alto, CA, EPRI NP-4887, November.

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and Environmental Sciences Laboratory.

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

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

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Hankins, D. E., 1973, A small, inexpensive albedo-neutron dosimeter, LA-5261, Los Alamos Scientific 

Laboratory, Los Alamos, New Mexico.

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Bureau of Radiological Health, Rockville, MD, January.

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

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

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

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by Workers, Publication 30, Pergamon Press, Oxford, England

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and Units, Report 19, Washington, D.C.
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the Determination of Dose Equivalent, Report 25, Washington, D.C. 

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and Units, Report 33, Washington, D.C.

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

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in Radiation Protection Dosimetry, Report 51, Bethesda, Maryland.

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

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July 29, Idaho Chemical Processing Plant.

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Quality Factor,” letter to J. Barry (ID Operational Safety Director), May 22, Radiological and 

Environmental Sciences Laboratory.

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

Personnel Badge,” letter to Dosimetry Branch, RSL/ID, March 25, Idaho Falls, Idaho.

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

Factors,” memorandum to Dosimetry Branch Staff, September 17.

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Resulting Mean Fluence to Dose Equivalent Conversion Factors,” Radiat. Prot. Dosim., 

volume 2, pp. 85–93.

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Atomic Energy Agency.

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NCRP Report No. 7, Washington, D.C.

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Electron Volts, Handbook 63, Washington, D.C., Nov 22.

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1951(subsequent printing)}.
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Mixtures of Neutrons and Gamma Rays, Recommendations of the National Council on 

Radiation Protection and Measurements. Handbook 75, NCRP Report No. 25, Washington, 

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Implementation Guidelines, OCAS-IG-001, Revision 1, Office of Compensation Analysis and 

Support, Cincinnati, Ohio.

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December 18. 

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Walker, D. I., 1974, letter to D. E. Black (Allied Chemical, ICPP), July 23, Idaho Operations Office, 

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GLOSSARY

1/E spectrum

For neutrons, the number of neutrons in an energy interval scales as the width of the energy 

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

beta particle

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

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

dose equivalent (H)

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

The special unit is the rem, the SI unit is Sievert (Sv) (1 Sv = 100 rem).

dose equivalent index

Maximum dose equivalent within the ICRU sphere centered at the point in space to which the 

quantity is assigned, HI. The outer 0.07 mm thick shell is ignored. It is also called the 

unrestricted dose equivalent index.

deep dose equivalent index

Maximum dose equivalent in the ICRU sphere within a core radius of 14 cm. The sphere is 

centered at the point in space to which the quantity is assigned. This quantity is one of the two 

restricted dose equivalent indices.

DOELAP

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

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

dosimeter

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

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

of absorbed dose or dose equivalent received by an individual 

effective dose equivalent, HE 

The weighted average of the dose equivalents in certain organs or tissues of the body, HT, 

each weighted by an organ weighting factor, WT. The organ weighting factors were chosen by 

the ICRP to reflect the relative risk of death from cancer or occurrence of severe hereditary 

effects in the first generations after uniform whole body exposure.

exposure

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

generic sense, ionizing radiation applied to matter. 

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

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

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

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

plastic, soft tissue ignoring trace elements, or soft tissue including trace elements).

Linear Energy Transfer (LET)

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

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

non-penetrating dose

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

the shielded.

pencil dosimeters

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

pencil, pocket dosimeter, pocket pencil, pocket ionization chamber (PIC).

penetrating dose equivalent

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

Essentially, personal dose equivalent HP (10).

personal dose equivalent, Hp(d)

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

radiological protection purposes (ICRU 1993). The Personal Dose Equivalent is represented 

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

For weakly penetrating radiation of significance to skin dose, d = 0.07 mm and is noted as 

Hp(0.07). For penetrating radiation of significance to “whole-body” dose, d = 10 mm and is 

noted as Hp(10). 

polymethyl methacrolate

Scientific name for plastic commonly known as Lucite or Plexiglas.

redacted

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

shallow absorbed dose (Ds)

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

composition.

shallow dose equivalent (Hs)

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

equivalent).

tissue rad

Absorbed dose in tissue. 
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ATTACHMENT 6A: INTERPRETATION OF INEEL DOSIMETRY CODES

LIST OF TABLES

Table

6A-1 Area Codes............................................................................................................................

Page

45

6A-2 Reasons Codes .....................................................................................................................45

6A-3 Irregularity Codes................................................................................................................... 46

6A-3 Column 20 Codes ..................................................................................................................46

6B-1 INEEL beta/photon dosimeter period of use, type, minimum reporting level, 

exchange frequency, and potential annual missed dose. ....................................................... 47

6B-2 Dosimeter type, period of use, exchange frequency, laboratory minimum 

detectable limit, and maximum annual missed dose............................................................... 47

6B-3 Recommended IREP Neutron Energy Fractions and Correction Factors for 

INEEL Facilities......................................................................................................................48

Table 6A-1: Area Codes: 

01 Headquarters (AEC 

Bldg.)

18 WP

02 EBR #1 19 TAN (Phillips & AEC)

03 CFA 20 TREAT

04 MTR 21 LX

42 TRA Monthly 22 GCRE

45 TRA Quarterly or Yearly 23 OX

05 CPP 24 ARHG

53 CPP Monthly 25

55 CPP Quarterly or Yearly 26 EBR #2

06 NRF 27 ML-1

07 TAN 28 On-Site Site Survey

08 Service 29 Off-Site Site Survey

09 NX (X is construction) 30 ANP at SL-1

10 AX 31 STPF

11 CX 65 ECF

12 EX 66 Non-Security

13 SPERT 67 Div. of Compliance

14 MORE 68 STEP

15 SX 69 LPTF (GE)

16 SL-1 70 LPTF (Phillips & AEC)

17 MX 71 CADRE

Table 6A-2: Reasons Codes (Column 68-69) . 

01 Regular Pull 11 Lost Pencil (or damaged)

02 H.P. Request 12 H.P. Check

03 High Dosimeter Reading 13 Late Pull

04 Recover Lost Badge 14 Withdrawn Badge

05 Ring Reading 15 Termination

06 Wrist Badge Reading 16 Correction

07 Recovered Lost Badge & Withdrawn 17 Records Withdrawn

08 18 Lost Film Reading

09 Miscellaneous Pull 19 X-Ray Exposure

10 Temporary Film 20 Exi mat Exposure
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Table 6A-3: Irregularity Codes (Columns 70-71) . 

01 Defective Film 12 Dropped in Canal or Reactor

02 Impossible to Read 13

03 Light Leak 14 Not in Area

04 Water Soaked 15

05 16

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

07 Lost in Processing 18 Stuck Film

08 Heat Exposure 19 Not Available

09 Recovered Lost Badge 20 Lost Badge

10 Contaminated Badge 21 No Film

11 Wore Two Badges at one Time

Table 6A-4: Column 20 Codes.

“X” Master Card 6 Fast Neutron

1 Summary Card 7 Urinalysis

3 Sens. Beta-Gamma 8 Summary Card

4 Insen. Gamma 9 Summary Card

5 Slow Neutron 0 Total Body Results Card
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ATTACHMENT 6B: OCCUPATIONAL EXTERNAL DOSE FOR MONITORED WORKERS

Table 6B-1. INEEL beta/photon dosimeter period of use, type, minimum reporting level, exchange 

frequency, and potential annual missed dose 

Period of usea Dosimeter

MRLb

(mrem)

Exchange 

frequency

Max. annual missed 

dose (mrem)c

August 1951–March 

1958

INEEL Initial Film 552 Dupont Film 30 Weekly (n=52) 780

Monthly (n=12) 180

Reactor Areas INEEL 558 Dupont Film 10 Weekly (n=52) 260

March 1958–

December 1966

INEEL Multi-Element Dupont 508 Film 10 Weekly (n=52) 260

Biweekly (n=26) 130

Monthly (n=12) 60

December 1966–

February 1974

INEEL Multi-Element Dupont 508 Film 10 Weekly (n=52) 260

Biweekly (n=26) 130

Monthly (n=12) 60

INEEL LiF TLD 15 Quarterly (n=4) 30

Semi-ann (n=2) 15

Annual (n=1) 7.5

February 1974–May 

1975d INEEL Atlas TLD LiF in Teflon 30 Monthly (n=12) 180

Quarterly (n=4) 60

Semi-ann(n=2) 30

Annual (n=1) 15

December 1974 

through December 

1985d 

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

Quarterly (n=4) 30

Annual (n=1) 7.5

January 1986 to 

present

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

Quarterly (n=4) 30

10e Monthly (n=12) 60

Quarterly (n=4) 20

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

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

b. Minimum reporting levels (MRL) are based on Cipperley (1958), Cipperley (1968), Cusimano (1963), Kalbeitzer (1983), 

Gesell (1986), Gesell (1992), Perry et al. (1993), and Ruhter (2002).

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

d. ICPP began using the Harshaw in December 1974, the prime contractor in February 1975, and Argonne in May 1975.

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

15 mrem until 1993 when it returned to 10 mrem.

Table 6B-2. Neutron dosimeter type, period of use, exchange frequency, laboratory minimum 

detectable limit, and maximum annual missed dose. 

Dosimeter Period

Exchange 

frequency

Laboratory MRL 

(mrem)

Maximum annual 

missed dose 

(mrem)

NTA film 1951-1958 Weekly 14 364

NTA film 1959–

September 1976

Weekly 20 520

Biweekly 20 260

Monthly 20 120

TLD October 1976–

present

Biweekly 15 195

Monthly 15 90

Quarterly 15 30
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Table 6B-3. Recommended IREP neutron energy fractions and correction factors for INEEL facilities.

Process Description Operations

Neutron 

energy

Default 

dose (%)

Ambient Dose 

Equiv / dose 

equiv

Net correction 

factor

Instrument 

calibration

Alpha Be source calibrations 1951 1993 0.1-2 MeV 20% 1.4 0.28

Cf-252 source calibrations 1993 2003 2-20 MeV 80% 1.1 0.88

Waste 

characterization

RWMC SWEPP 14 MeV 

neutron generator

~1980 2003 0.1-2 MeV 5% 1.4 0.07

2-20 MeV 95% 1.1 1.05

Neutron source 

based research

2003 0.1-2 MeV 20% 1.4 0.28

2-20 MeV 80% 1.1 0.88

MTR, ZPPR, and 

TREAT reactors

Experiment floor and adjacent 

rooms during operation

1953 1970 < 10 keV 20% 1 0.20

10 -100 keV 5% 1.1 0.06

0.1-2 MeV 50% 1.4 0.7

2-20 MeV 25% 1.1 0.28